United States        Office of          EPA 520/1 -85-023
          Environmental Protection     Radiation Programs       August 1985
          Agency          Washington, D.C. 20460


          Radiation
<&EPA    High-level and Transuranic
          Radioactive Wastes

          Background Information
          Document for Final Rule

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40 CFR Part 191                                             EPA 520/1-85-023
Environmental Standards for the
Management and Disposal of Spent
Nuclear Fuel, High-Level and
Transuranic Radioactive Wastes
                        BACKGROUND INFORMATION DOCUMENT
          FINAL RULE  FOR HIGH-LEVEL AND  TRANSURANIC  RADIOACTIVE WASTES
                                 August 1985
                    U.S.  Environmental Protection Agency
                        Office of Radiation Programs
                           Washington, D.C.  20460

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                                  CONTENTS
1.   Introduction                                                      1-1

          1.1  EPA authorities for the rulemaking                      1-1
          1.2  History of the high-level radioactive  waste  program
                 and the EPA rulemaking                                1-2
          1.3  Purpose and scope of BID                                1-5
          1.4  Computer codes utilized                                 1-5
          1.5  Program technical support documents                     1-6

2.   Current Regulatory Programs and Strategies                        2-1

          2.1  Introduction                                            2-1
          2.2  The International Commission on Radiological
                 Protection and the National Council  on Radiation
                 Protection and Measurements                           2-2
          2.3  Federal guidance                                        2-8
          2.4  The Environmental Protection Agency                     2-10
          2.5  Nuclear Regulatory Commission                           2-12
          2.6  Department of Energy                                    2-13
          2.7  Department of Transportation                            2-14
          2.8  State agencies                                          2-14
          2.9  Indian tribes                                           2-15

3.   Quantities, Sources, and, Characteristics of Spent Nuclear
       Fuel and High-Level and Transuranic Wastes                      3-1

          3.1  Introduction                                            3-1
          3.2  Spent nuclear fuel                                      3-1
          3.3  High-level adioactive wastes                            3-4
          3.4  Transuranic wastes                                      3-10

4.   Planned Disposal Programs                                         4-1

          4.1  Introduction                                            4-1
          4.2  Civilian Radioactive Waste Management  Program           4-2
          4.3  Geological media                                        4-5
          4.4  Waste Isolation Pilot Plant                             4-7
          4.5  Disposal of DOE defense high-level wastes               4-8
          4.6  Alternative disposal methods                            4-8
                                     iii

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                            CONTENTS  (continued)


                                                                      Page

 5.   Radiation Dosimetry

           5.1  Introduction                                            5-1
           5.2  Definitions                                             5-1
           5.3  Dosimetric models                                       5-3
           5.4  EPA dose calculation                                    5-9
           5.5  Uncertainty analysis                                    5-11
           5.6  Distribution of doses in the general population         5-26
           5.7  Summary                                                 5-29

 6.   Estimating the Risk of Health Effects Resulting From
       Radionuclides                                                   6-1

           6.1  Introduction                                            6-1
           6.2  Cancer risk estimates for low-LET radiations            6-2
           6.3  Fatal cancer risk resulting from high-LET
                 radiation                                             6-19
           6.4  Uncertainties in risk estimates for radiogenic
                 cancer                                                6-22
           6.5  Other radiation-induced health effects                  6-30
           6.6  Radiation Risk - a perspective                          6-51

 7.   Movement and Health Risks of Radionuclide Releases to the
       Accessible Environment                                          7-1

           7.1  Introduction                                            7-1
           7.2  Methodology                                             7-2
           7.3  Releases to surface water                               7-3
           7.4  Releases to an ocean                                    7-11
           7.5  Releases directly to land surface                       7-11
           7.6  Releases due to a volcanic eruption or meteorite
                 impact                                                7-11
           7.7  Special considerations for Garbon-14 environmental
                 risk commitment                                       7-12
           7.8  Fatal cancers per curie released to the accessible
                 environment                                           7-13
           7.9  Uncertainty analysis                                    7-17


8.   Risk Assessment of Disposal of High-Level Radioactive Waste
       in Mined Geologic Repositories                                  8-1

          8.1  Introduction                                            8-1
          8.2  Waste disposal system model                             8-2
          8.3  Time frame                                              8-4
          8.4  Measures of risk                                        8-4
                                     iv

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                            CONTENTS (continued)
          8.5  Computer codes utilized                                 8-7
          8.6  Site parameters                                         8-7
          8.7  Repository parameters                                   8-11
          8.8  Waste package parameters                                8-14
          8.9  Release mechanisms                                      8-16
          8.10 Risk assessments for models of geologic repositories    8-22
          8.11 Uncertainties in the risk assessment                    8-29
          8.12 Radiation risks from other sources                      8-46

Appendix A     A description of the RADRISK and CAIRD computer
                 codes used by EPA to assess doses and risks
                 from radiation exposure                               A-l

Appendix B     Glossary and acronyms                                   B-l

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                                   FIGURES
Number                                                                Page

4.2-1     Regions Identified by DOE as Under Consideration for
            Geological Disposal of High-level Nuclear Waste            4-3

4.2-2     Sites Identified by DOE as Potentially Acceptable
            for the First Repository                                   4-4

5.3-1     Typical Pattern of Decline of Activity of a Radio-
            nuclide in an Organ, Assuming an Initial Activity
            in the Organ and No Additional Uptake of Radio-
            nuclide by the Organ                                       5-5

5.3-2     The ICRP Task Group Lung Model for Particulates              5-6

5.3-3     Schematic Representation of Radionuclide Movement
            Among Respiratory Tract, Gastrointestinal Tract,
            and Blood                                                  5-7

5.5-1     Dose Rate Front Chronic Ingestion of Iodine-131 in
            Water at a Concentration of 1 uCi/fc                        5-15

5.5-2     Dose Rate From Chronic Inhalation of Iodine-131 in
            Air at a Concentration of 1 yCi/m3                         5-16

5.5-3     Compartments and Pathways in Model for Strontium in
            Skeleton                                                   5-18

5.5-4     Dose Rate From Chronic Ingestion of Strontium-90
            in Water at a Concentration of 1 uCi/i                     5-19

5.5-5     Dose Rate From Chronic Inhalation of Strontium-90
            in Air at a Concentration of I uCi/m9                      5-20

5.5-6     Dose Rate From Chronic Inhalation of Plutonium-239
            in Air at a Concentration of 1 yCi/ta3                      5-22

5.5-7     Compartments and Pathways in Model for Plutonium
            in Skeleton                                                5-23

5.5-8     Dose Rate From Chronic Ingestion of Plutonium-239
            in Water at a Concentration of 1 uCi/i                     5-2A
                                     vi

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                             FIGURES (continued)

Number                                                                 Page

5.5-9     Dose Rate From Chronic Inhalation of Plutonium-239
            in Air at a Concentration of 1 wCi/m3                      5-25

7.9-1     Probability Distribution of Population Risks Per Curie
            of AM-243 Released to Surface Water                        7-19

8.2-1     Components Included in the Risk Assessment of Radioactive
            Waste Releases                                             8-3

8.4-1     Hypothetical Complementary Cumulative Distribution Function
            (CCDF) of Human Health Effects                             8-6

8.6-1     Cross Section of the Rock Formation at the Generic
            Repository Site                                            8-10

8.10-1    Population Risks From Disposal in Geologic Repositories
           (Reference Cases)                                           8-24

8.10-2    Population Risks From Disposal in Geologic Repositories
            (Logarithmic Scale, Reference Cases)                       8-25

8.10-3    Complementary Cumulative Distribution Functions of the
            Population Risks for Disposal in Basalt and Tuff           8-27

8.10-4    Complementary Cumulative Distribution Functions of the
            Population Risks for Disposal in Bedded Salt               8-28

8.10-5    Radiation Exposures From Drinking Ground Water at a
            2-Kilometer Distance From a Repository                     8-30

8.11-1    The Effect of Canister Life and Waste Form Leach Rate on
            Population Risks for three Potentially Suitable Reposi-
            tory Media                                                 8-33

8.11-2    The Effect of Canister Life and Waste Form Leach Rate on
            Estimated Population Risks for Repositories in Granitic
            Formations                                                 8-34

8.11-3    Effect of Canister Life and Waste Form Leach Rate on Radi-
            ation Exposures From Drinking Ground Water at 2 Kil-
            ometers From Repository                                    8-35

8.11-4    Effects of Geochemical Parameters on Population Risks for
            Different Geologic Media                                   8-37

8.11-5    Effect of Solubility Limits on Population Risks for Amer-
            icium in Different Geologic Media                          8-39
                                     vii

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                                   FIGURES

Number                                                                 Page

8.11-6    Sensitivity of Population Risks to Repository Distance to
            the Accessible Environment                                 8-40

8.11-7    Sensitivity of Population Risks to Event Probabilities       8-42

8.11-8    Sensitivity of Population Risks to Ground Water Flow in
            Tuff Formations                                            8-43
                                    viii

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                                   TABLES
Number                                                                Page

3.2-1     Historical and Projected Mass and Radioactivity of
            Commercial Spent Fuel                                      3-3

3.2-2     Historical and Projected Installed Nuclear Electric
            Power Capacity                                             3-3

3.3-1     Current Volume of HLW in Storage by Site Through
            1983                                                       3-6

3.3-2     Current Radioactivity of HLW in Storage by Site
            Through 1983                                               3-7

3.3-3     Historical and Projected Volume and Associated
            Radioactivity of HLW in Storage by Site Through
            2000                                                       3-8

3.4-1     Inventories and Characteristics of DOE/Defense TRU
            Wastes Buried Through 1983                                 3-13

3.4-2     Inventories and Characteristics of DOE/Defense
            Waste in TRU Retrievable Storage Through 1983              3-14

3.4-3     Estimated Inventories of Items That Might Require
            Special Handling and/or Treatment as TRU Waste             3-15

3.4-4     Physical Composition of TRU Wastes at DOE/Defense Sites      3-16

3.4-5     Estimated Isotopic Composition of Burled, Retrievably
            Stored, and Future TRU Waste                               3-17

3.4-6     Current Inventories and Projections of DOE Buried and
            Stored TRU Waste From Defense Activities                   3-19

5.3-1     Organs for Which Dose Rates are Calculated                   5-9

5.5-1     Age-Dependent Parameters for Iodine Metabolism in
            the Thyroid                                                5-17
                                     ix

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                              TABLES  (continued)

 Number

 6.2-1     Range  of  Cancer  Fatalities Induced by  10 Rads  of Low-LET
            Radiation  (Average  Value Per  Rad Per Million Persons
            Exposed)                                                    6-10

 6.2-2     A Comparison of  Estimates  of  the Risk  of Fatal Cancer
            From a  Lifetime Exposure at 1 Rad/Year (Low-LET
            Radiation)                                                  6-11

 6.2-3     Proportion of the Total Risk of Fatal  Radiogenic Cancer
            Resulting  From Cancer at a Particular Time                  6-15

 6.2-4     UNSCEAR77 Estimates of Cancer Risks at Specified Sites        6-16

 6.2-5     Comparison of Proportion of the Total  Risk of  Radiogenic
            Cancer  Fatalities by Body Organ                             6-16

 6.3-1     Estimated Number of Cancer Fatalities  From a Lifetime Expo-
            sure to Internally  Deposited Alpha Particle  Emitters        6-22

 6.4-1     A Ranking of Causes of Uncertainty in  Estimates of the Risk
            of Cancer                                                   6-29

 6.5-1     ICRP Task Group Estimate of Number of  Cases of Serious
            Genetic 111 Health  in Liveborn From  Parents  Irradiated
            With 106 Man-Rem in a Population of  Constant Sizea
            (Assumed Doubling Dose -' 100 Rad)                           6-35

 6.5-2     BEIR-3 Estimates of Genetic Effects of an Average Popu-
            lation Exposure of  1 Rem Per 30-Year Generation             6-36

 6.5-3     UNSCEAR 1982 Estimated Effect of 1 Rad Per Generation of
            Low Dose or Low Dose Rate, Low-LET Radiation on a
            Population of 106 Liveborn According to the Doubling
            Dose Method (Assumed Doubling Dose - 100 Rad)               6-37

6.5-4     Summary of Genetic Risk Estimates Per  106 Liveborn for
            an Average Population Exposure of 1 Rad of Low Doae or
            Low Dose Rate, Low-LET Radiation in a 30-Year Generation    6-38

6.5-5     Estimated Frequency of Genetic Disorders in a Birth Cohort
            Due to Exposure of  the Parents to 1 Rad Per Generation      6-43

7.2-1     Release Modes and Environmental Pathways                      7-3

7.2-2     Fatal Cancer Conversion Factors                               7-4

7.3-1     Bioaccumulation Factors for Freshwater Fish                   7-8

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                             TABLES (continued)

Number                                                                 Page

7.3-2     Radionuelide Intake Factors for Farm Products Raised in
            Areas Using Contaminated Irrigation Water                  7-9

7.3-3     Values for Persons Fed Per Unit Area of Land                 7-10

7.8-1     Fatal Cancers Per Curie Released to the Accessible
            Environment for Different Release Modes                    7-14

7.8-2     Fatal Cancers Per Curie Released to the Accessible
            Environment for Releases to Surface Water                  7-15

7.8-3     Development of Release Limits Presented in Table 1 of
            40 CFR Part 191                                            7-16

8.6-1     Site Parameters Considered in Risk Assessment                8-9

8.6-2     Geochemical Parameters Used in Risk Assessment               8-12

8.7-1     Repository Parameters Considered in Risk Assessment          8-13

8.6-1     Radionuclide Inventory in Repository (Spent Nuclear Fuel)    8-15

8.9-1     Release Mechanism Parameters Considered in Risk Assess-
            ments                                                      8-17

8.10-1    Fatal Cancers Over 10,000 Years by Release Mechanism
            and Radionuclide                                           8-26

8.11-1    Alternative Geochemical Parameters Considered in Risk
            Assessment                                                 8-36

8.12-1    Distribution of Natural Radiation Annual Dose Equivalents
            (Terrestrial, Cosmic* and Internal)                        8-48
                                     xi

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                        Chapter 1:  INTRODUCTION
     The U.S. Environmental Protection Agency (EPA) is responsible for
developing and Issuing environmental standards, guidelines, and criteria
to ensure that the public and the environment are adequately protected
from potential radiation impacts.

     Toward this end, EPA is promulgating generally applicable environ-
mental standards for the management and disposal of spent nuclear fuel
and high-level and transuranic radioactive wastes.  These standards
provide the basic framework for long-term control through management and
disposal of three types of waste:

     1)   Spent nuclear reactor fuel if disposed of without reprocessing.

     2)   High-level radioactive liquid or solid wastes from the re-
          processing of spent nuclear fuel.

     3)   Transuranic wastes containing long-lived radionuclides of
          elements heavier than uranium and defined as containing 100
          nanocuries or more of alpha-emitting transuranic nuclides, with
          half-lives greater than 20 years, per gram of waste.

1.1  EPA Authorities for the Rulemaking

     These standards have been developed pursuant to the Agency's author-
ities under the Atomic Energy Act of 1954, as amended, and Reorganization
Plan No. 3 of 1970.

     The basic authority under the Atomic Energy Act of 1954, as amended
and transferred from the Atomic Energy Commission to EPA through the
Reorganization Plan No. 3 of 1970, includes the function 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 quan-
tities of radioactive material, in the general environment outside the
boundaries of locations under the control of persons possessing or using
radioactive material"  (N170).

     The Nuclear Waste Policy Act (NWPA) of 1982 established formal
procedures regarding the evaluation and selection of sites for geologic
repositories, including procedures for the interaction of State and
                                     1-1

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 Federal governments; reiterated the existing responsibilities of the
 Federal agencies involved in the national program; and provided a time-
 table for several key milestones to be met by the Federal agencies in
 carrying out the program.  As part of this national program, the EPA,
 pursuant to its authorities under other provisions of law, "shall, by
 rule, promulgate generally applicable standards for protection of the
 general environment from offsite releases from radioactive material in
 repositories" (NWPA82).

 1.2  History of the High-level Radioactive Waste Program and the EPA
      Rulemaking

      Since the inception of the nuclear age in the 1940's, the Federal
 government has assumed ultimate responsibility for the care and disposal
 of high-level radioactive wastes regardless of whether they are produced
 by commercial or national defense activities.   In 1949 the Atomic Energy
 Commission (ABC) initiated research and development vork aimed at develop-
 ing systems for the conversion of high-level liquid wastes to chemically
 stable solids.  Then,  in 1955, at the request  of the AEC, a National
 Academy of Sciences - National Research Council (NAS-NRC) advisory commit-
 tee was established to consider the disposal of high-level radioactive
 wastes within the United States.   Their report, issued in 1957, contained
 two general recommendations (NAS57):   1)  that  the AEC continue efforts to
 develop processes for the solidification  of high-level radioactive
 liquid wastes, and 2)  that naturally  occurring salt formations are the
 most promising medium for the  long-term isolation of these solidified
 wastes.  Project Salt  Vault,  conducted from 1965-1967 by  the AEC in an
 abandoned salt mine near Lyons,  Kansas,  demonstrated the  safety and
 feasibility of handling and storing solid wastes in salt  formations
 (Mc70).

      In 1968,  the AEC  again requested the NAS-NRC to establish a Committee
 on Radioactive Waste Management  (CRWM)  to advise the AEC  concerning its
 long-range radioactive waste management  plane  and to evaluate the feasibi-
 lity of disposing of solidified  radioactive wastes in bedded salt.  The
 CRWM convened  a panel  to discuss  the  disposal  of radioactive waste In
 salt mines.   Based on  the recommendations of the panel,  the CRWM concluded
 that the use  of  bedded salt is satisfactory for the disposal of radioac-
 tive waste (NAS70).

      In 1970,  the AEC  announced  the tentative  selection of a site at
 Lyons,  Kansas,  for the establishment  of  a national radioactive waste
 repository (AEC70).  During the next  two  years,  however,  in-depth site
 studies  raised  several questions  concerning the safe plugging of old
 exploratory wells and  proposed expanded  salt mining activities.   These
 questions  and  the growing public  opposition to the Lyons  site prompted
 the  AEC  in late  1971 to pursue alternatives to the salt site at Lyons
 (Do72).

      In  1976,  the Federal government  intensified its program to develop
 and  demonstrate  a permanent disposal  method for high-level radioactive
wastes.  The Office  of  Management and Budget (OMB)  established an inter-
 agency  task force  on commercial nuclear wastes in March 1976.   The  OMB

                                      1-2

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 interagency task force defined the scope and responsibility of each
 Federal agency's activities on high-level vaste management, including the
 preparation of environmental standards for high-level waste by EPA (Ly76,
 En77a,b).

     A status report on the management of commercial radioactive nuclear
 wastes, published in May 1976 by the President's Federal Energy Resources
 Council (ERC), emphasized the need for coordination of administration
 policies and programs relating to energy.  The ERC established a nuclear
 subcommittee to coordinate Federal nuclear policy and programs to assure
 an integrated government effort.  This report called for an accelerated
 comprehensive government radioactive waste program plan calling for an
 Interagency task force to coordinate activities among the responsible
 Federal agencies.  The EPA was given the responsibility of establishing
 general environmental standards governing waste activities, including
 high-level radioactive wastes that must be delivered to Federal reposi-
 tories for long-term management (FERC76).

     In October 1976, President Ford issued a major statement on nuclear
 policy.  As part of hie comprehensive statement, he announced new steps
 to assure that the United States has the facilities for long-term manage-
 ment of nuclear wastes from commercial powerplants.  The President's ac-
 tions were based on the findings of the OMB interagency task force formed
 in March 1976.  He announced that the experts had concluded that the most
 practical method for disposing of high-level waste is geologic storage in
 repositories in stable formations deep underground.  Among the many steps
 to be taken was EPA's issuance of general environmental standards govern-
 ing nuclear facility releases to the biosphere above the natural back-
 ground radiation level (Fo76).  These standards were to place a numerical
 limit on long-term radiation releases outside the boundary of the reposi-
 tory.

     In December 1976, EPA announced its intent to develop environmental
 radiation protection standards for high-level radioactive waste to assure
 protection of the public health and the general environment (EPA76).
 These efforts have included frequent interaction with the public, which
 began with a series of public workshops on radioactive waste disposal in
 1977 and 1978 (EPA77a,b, EPA78a,b).

     In 1978, President Carter established the Interagency Review Group
 (IRG) to develop recommendations for the establishment of an administra-
 tive policy with respect to long-term management of nuclear wastes and
 supporting programs to implement the policy.  The IRG report reemphasized
 EPA's role in developing generally applicable standards for the disposal
 of high-level waste, spent nuclear fuel, and transuranic wastes (DOE79).
 In a Message to Congress on February 12, 1980, the President outlined a
 comprehensive national radioactive waste management program based on the
 IRG report.  The message called for an interim strategy for disposal of
high-level and transuranic wastes that would rely on mined-out geologic
 repositories.  The message repeated that the EPA was responsible for
 creating general criteria and numerical standards applicable to nuclear
waste management activities (CaBO).
                                     1-3

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      In November 1978, the EPA published, proposed "Criteria for Radioac-
 tive Wastes," which were intended as Federal guidance for storage and
 disposal of all forms of radioactive wastes (EPA78c).  In March 1981,
 however, EPA withdrew the proposed criteria because the many different
 types of radioactive wastes made the issuance of generic disposal guid-
 ance too difficult (EPA81).

      Development efforts continued, and on December 29, 1982, EPA pub-
 lished a proposed rule on "Environmental Radiation Protection Standards
 for the Management and Disposal of Spent Nuclear Fuel,  High-Level and
 Transuranic Radioactive Wastes" (EPA82).

      Shortly after the EPA proposed rule was published. Congress passed
 the^Nuclear Waste Policy Act of 1982 (P.L. 97-425), wherein the EPA was
 to "...promulgate generally applicable standards for protection of the
 general environment from offsite releases from radioactive material in
 repositories" not later than January 1984 (NWPA83).

      After the first comment period on the proposed rule ended on May 2,
 1983, EPA held two public-4iearings on the proposed  standards—one in
 Washington,  D.C., May 12-14, 1983, and one in Denver, Colorado,
 May 19-21, 1983—and requested  post-hearing comments during a second
 comment period (EPA83a,b).   More than 200 comment letters were received
 during these two comment periods,  and 13 oral statements were made at the
 public hearings.   Responses to  comments received from the public are pre-
 sented in "Response to Comments, Volume I - Public  Comments" (a companion
 document).

      In parallel  with this  public  review and comment, the Agency conduct-
 ed an independent scientific review of the technical basis for the pro-
 posed 40 CFR 191  standards  through a special Subcommittee of the Agency's
 Science Advisory  Board (SAB).   This Subcommittee held nine public meet-
 ings  from January 18,  1983,  through September 21,  1983  (EPA83c).  The SAB
 then  prepared a final report that  was transmitted on February 17,  1984
 (SAB84).   Although the SAB  review found that the Agency's analyses in
 support  of the proposed standards  were comprehensive and scientifically
 competent, the report contained several findings and recommendations for
 improvement.   The public was notified of the availability of this  report
 on May  8,  1984, and encouraged  to  comment on the findings and recommenda-
 tions  (EPA84).  Responses to the SAB report  are  presented in another
 companion  document  entitled "Response to Comments,  Volume II - Science
 Advisory Board Comments."

     On February  8,  1985, the Natural Resources  Defense Council, Inc.,
 the Environmental Defense Fund,  the Environmental Policy Institute,  the
 Sierra Club,  and  the  Snake  River Alliance brought suit  against the Agency
 and the Administrator  because of noncompliance with the January 7,  1984,
 deadline mandated by  the NWPA for  promulgation of standards.   A consent
 order was  negotiated with the plaintiffs that required  the standards to
be promulgated on or before  August 15,  1985.
                                     1-4

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1.3  Purpose and Scope of BID

     The purpose of this document is to provide background information
that, when considered together with the promulgated generally applicable
standards, supports the final actions taken by the EPA vith regard to the
management and disposal of spent nuclear fuel and high-level and trans-
uranic wastes.  It also contains an integrated risk assessment that
provides a scientific basis for these actions.

     The scope encompasses the conceptual framework for assessing radia-
tion risks, including identification of the sources of possible radionu-
clide releases, analysis of the movement of the radionuclides from the
source through environmental pathways, estimates of doses received by
human individuals and populations, and calculations of the probability of
genetic and somatic health effects.

1.4  Computer Codes Utilized

     A number of computer codes have been used as tools itv the Agency's
risk analyses.  The central tool has been the program REPRISK, which has
been under development at the Agency since 1978.  This code, described in
Chapter 8, makes use of conversion factors that relate the amount of
radioactive material released to the accessible environment to population
health effects.  These conversion factors are obtained by using another
EPA computer code called WESPDOSE (Sm85).  UESPDOSE considers a number of
pathways for the environmental transport of radionuclides.  For calcula-
tions involving individual doses and time frames longer than ten thousand
years, the computer code NWFT/ DVM, developed by Sandia under contract to
NRC, has been used (Ca81).  This code models the transport of decay
chains whose elements have different retardation factors in the ground-
water system.  A more complex groundwater code, SWIFT, has also been used
to support the EPA risk analyses, primarily to validate some of the
hydrologic calculations carried out using simpler models  (Re81).

     Four kinds of "release mechanisms" are addressed by REPRISK:

     1)   Direct impact on a waste package with associated releases to
          the air and/or the land surface  (e.g., volcano, meteorite,
          drilling/direct hit).

     2)   Direct impact on a waste package with associated releases to an
          aquifer (e.g., faulting, breccia pipes).

     3)   Disruption of the repository with associated releases to the
          land surface  (e.g., drilling/no hit).

     A)   Disruption of the repository with associated releases to an
          aquifer (e.g., normal groundwater flow,  faulting, breccia
          pipes, drilling/no hit).
                                      1-5

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     Each release mechanism leads to several pathways to human exposure.
The consequences of a radioactivity release to the accessible environment
ate expressed in terms of 1) number of somatic health effects (fatal
cancers), 2) number of genetic health effects* 3) ratio of released
amount to the release limit in 40 CFR Part 191, and/or 4) curies released
of each radionuclide.

     Two time frames are used by the model.  One, called a dose commitment
period, is for modeling the occurrence of release mechanisms at the site.
The other, a dose integration period, is used for measuring the conse-
quences of the releases.  This way, consequences may be measured beyond
the time when a particular perturbation may be active at a site.

1.5  Program Technical Support Documents

     A number of technical support documents have been prepared and
published during the history of the rulemaking and standards program to
help establish the technical basis for the standards.  These documents
should also be considered as part of the technical background for the
present rulemaking process,   The following is a listing of these docu-
ments and a short abstract of each.

     (1)  Technical Support of Standards for High-Level Radioactive Waste
          Management - Volume A, Source Term Characterization, EPA 520/4-
          79-007A,  March-July 1977.

          This report provides a characterization of commercial spent
          nuclear fuel and high-level waste,  including comparisons of
          source terms from various  fuel cycles and fuel mixes;  a char-
          acterization of government high-level and transuranic wastes; a
          comparison with commercial waste;  and an estimation of existing
          and projected quantities of spent nuclear fuel and high-level
          and transuranic wastes.  The data  are presented in several
          formats and on a specific  basis (per unit of fuel used or
          energy generated),  as well as on a  total basis for a given
          number of nuclear  powerplants.

     (2)   Technical Support  of Standards for  High-Level Radioactive
          Wastes Management  - Volume B,  Engineering Controls,  EPA 520/4-
          79-007B,  March-August 1977.

          This  report  reviews the  technology  for  engineering control of
          spent  fuel and  high-level  and TRU wastes and projected costs of
          the various  disposal technologies.   Analysis includes  process-
          ing and packaging  technology,  alternative geologic disposal
          techniques,  effectiveness  of  engineering controls,  and the cost
          considerations.

    (3)  Technical  Support of Standards  for  High-Level Radioactive Waste
         Management - Volume  C, Migration Pathways,  EPA 520/4-79-007C,
         March-July  1977.
                                    1-6

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     This report  assesses geologic  site  selection  factors; quanti-
     fies the potential  for the  migration of nuclides  through the
     geosphere to the biosphere,  and  considers  dose  implications of
     a repository for wastes containing  large quantities of radio-
     nuclides in  high concentrations  that might become dispersed
     into the biosphere  over geologic times.

(4)   Technical Support of Standards for  High-Level Radioactive Waste
     Management - Volume D, Release Mechanisms, EPA  520/4-79-007D,
     March 1980.

     This report  analyzes the potential  for the release of radionu-
     clides from  a generic deep-mined repository for radioactive
     wastes.   Five different geologic media are considered:  bedded
     salt, dome salt, granite, basalt, and shale.  A. range of poten-
     tial containment failure mechanisms were evaluated and  com-
     pared.  Results are combined with radionuclide  transport and
     dose calculations for assessment of the potential effects  of a
     repository on human health.

(5)   Technical Support of Standards for  High-Level Radioactive  Waste
     Management - Addendum to Volumes C  and D,  EPA 520/4-79-007E,
     March 1982.

     This report  is an update of information and  issues relevant  to
     the conclusions of  Volumes  C and D.

(6)   Assessment of Waste Management of Volatile Radionuclides,  EPA
     ORP/CSD-79-2, May 1979.

     This report  reviews waste management technologies in terms of
     immobilization, containment, and disposal  of  the radionuclides
     1-129, Kr-85, H-3,  and C-14.  Included are alternative  disposal
     options that may be applied to isolate these  wastes  from the
     human environment.

(7)   Radiation Exposures From Solidification Processes for High-
     Level Radioactive Liquid Wastes, EPA 520/3-80-007, May  1980.

     This report  is an assessment of a generic  high-level liquid
     waste solidification plant  and the potential environmental
     impact of atmospheric discharges during normal  operations
     involving four different solidification processes.

(8)   A Review of  Radiation Exposure Estimates  From Normal Operations
     in the Management and Disposal of High-Level Radioactive Wastes
     and Spent Nuclear Fuel, EPA 520/3-80-008,  August 1980.

     This report  provides an analysis of the estimated radioactive
     releases during normal waste management operations (i.e..
     preparation  for storage or disposal, storage, and emplacement)
     and the resulting radiation doses.
                                1-7

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 (9)  Alternative Disposal Concepts for High-Level and Transuranic
      Radioactive Waste Disposal, EPA ORP/CSD-79-l, May 1979.

      This report examines several technologies that have been pro-
      posed as alternative concepts to geologic disposal, including
      transmutation, extraterrestrial disposal, seabed disposal,
      ice-sheet disposal, and other continental geologic disposal.

(10)  Economic Impacts of 40 CFR 191:  Environmental Standards and
      federal Guidance for Management and Disposal of Spent Nuclear
      Fuel, High-Level and Transuranic Radioactive Wastes, EPA
      520/4-80-014, December 1980.

      This report develops a methodology for examining the potential
      economic impacts of the proposed environmental standards.

(11)  Environmental Pathway Models for Estimating Population Health
      Effects from Disposal of High-level Radioactive Waste in Geo-
      logic Repositories, Draft Report EPA 520/5-80-002, December
      1982.

      This report describee the mathematical models formulated to
      calculate the environmental dose commitments and population
      health effects (fatal cancers and first generation genetic
      defects) that could occur as a result of releases from geologic
      repositories.

(12)  Population Risks From Disposal of High-Level Radioactive Wastes
      in Geologic Repositories, Draft Report, EPA 520/3-80-006,
      December 1982.

      This report presents estimated population risks associated with
      disposal of the wastes in mined geologic repositories and
      describes the methods used to arrive at these estimates.

(13)  Draft Regulatory Impact Analysis for 40 CFR 191:  Environmental
      Standards for Management and Disposal of Spent Nuclear Fuel,
      High-Level and Transuranic Radioactive Wastes, EPA 520/1-82-
      024, December 1982.

      This report reviews th« projected costs associated with the
      management and disposal of high-level radioactive waste and
      evaluates the potential effects of the proposed 40 CFR 191
      environmental standards for disposal of these wastes.

(14)  Draft Environmental Impact Statement for 40 CFR 191:  Environ-
      mental Standards for Management and Disposal of Spent Nuclear
      Fuel and High-Level and Transuranic Radioactive Wastes, EPA
      520/1-82-025, December 1982.

      This report provides technical support information for the
      proposed environmental standards (40 CFR Fart 191).
                                 1-8

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(15)   State of Geological Knowledge Regarding Potential Transport  of
      High-Level Radioactive Waste From Deep Continental Repositor-
      ies,  EPA 520/4-78-004.

      This  report contains an evaluation by an ad hoc panel of earth
      scientists concerning the adequacy of basic knowledge in the
      pertinent earth sciences for reliably estimating environmental
      impacts.

(16)   Population Risks From Uranium Ore Bodies, EPA 520/3-80-009,
      October 1980.

      This  report presents a methodology for estimating the radiolog-
      ical  releases  and potential impact of deep-lying uranium ore on
      people.
                                 1-9

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                                REFERENCES


 AEC70     AEC Press Release No. N-102,  June 17,  1970.

 Ca80      The White House,  President J.  Carter,  The President's Program
           on Radioactive Waste Management,  Fact  Sheet,  February 12,  1980.

 Ca81      Campbell J.  E., D.  E. Longsine, and R.  M. Cranwell,  Risk
           Methodology  for Geologic  Disposal of Radioactive Waste:  The
           NWFT/DVM Computer Code Users Manual, Sandia National Labora-
           tories,  Report SAND81-0886 (NUREG/CR-2081), November 1981.

 DOE79     Department of  Energy, Report to the President by the Inter-
           agency Review  Group on Nuclear Waste Management,  Report  No.
           TID-29442, March  1979.

 Do72      Doub W.  0. Commissioner,  USAEC, Statement before the Science,
           Research and Development  Subcommittee of  Committee on Science
           and Astronautics, U.S.  House of Representatives,  U.S.  Congress,
           Washington, D.C., May 11  and 30,  1972.

 En77a     English  T. D.,  et al.,  An Analysis  of the Back End of  the
           Nuclear  Fuel Cycle  With Emphasis  on High-level Waste Manage-
           ment, JPL Publication 77-59, Volumes 1  and II, Jet Propulsion
           Laboratory, Pasadena,  California, August  12,  1977.

 En77b      English  T. D.,  et al.,  An Analysis  of the Technical  Status of
           High-level Radioactive  Waste and  Spent  Fuel Management Systems,
           JPL Publication 77-69,  Jet Propulsion Laboratory, Pasadena,
           California, December  1, 1977.

EPA76      Environmental Protection  Agency,  Environmental Radiation Pro-
           tection Standards for High-Level  Radioactive Waste,  Advance
          Notice of Proposed  Rulemaking, Federal  Register, 41(235);53363.
          Monday, December 6,  1976.

EPA77a    Environmental Protection Agency,  Proceedings:  A Workshop on
          Issues Pertinent to the Development of  Environmental Protection
          Criteria for Radioactive Wastes,  Reston, Virginia, February 3-5,
          1977, Office of Radiation Programs, Report No. ORP/CSD-77-1,
          Washington, D.C.,  1977.
                                     1-10

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EPA77b
EPA76a
EPA78b
EPA78C
EFA81
EFA82
EPA83a
EPA83b
EPA83c
EPA84
FERC76
Environmental Protection Agency, Proceedings:  A Workshop on
Policy and Technical Issues Pertinent to the Development of
Environmental Protection Criteria for Radioactive Wastes,
Albuquerque, New Mexico, April 12-17, 1977, Office of Radiation
Programs, Report No. ORP/CSD-77-2, Washington, D.C., 1977.

Environmental Protection Agency, Background Report - Considera-
tion of Environmental Protection Criteria for Radioactive
Waste, Office of Radiation Programs, Washington, D.C., February
J.7/8*

Environmental Protection Agency, Proceedings of a Public Forum
on Environmental Protection Criteria for Radioactive Wastes,
Denver, Colorado, March 30-April 1, 1978, Office of Radiation
Programs, Report No. ORP/CSD-78-2, Washington, D.C., May 1978.

Environmental Protection Agency, Recommendations for Federal
Radiation Guidance, Criteria for Radioactive Wastes, Federal
Register, 43_(221):53262-53268, Wednesday, November 15, 1978.

Environmental Protection Agency, Withdrawal of Proposed
Regulations, Federal Register, 46(53):17567, Thursday,
March 19, 1981.                —

Environmental Protection Agency, Proposed Rule, Environmental
Standards for the Management and Disposal of Spent Nuclear
Fuel, High-Level and Transuranic Radioactive Wastes, Federal
Register, 47/250):58196-58206, Wednesday, December 29, 1982.

Environmental Protection Agency, Environmental Standards for
Management and Disposal of Spent Nuclear Fuel, High-Level and
Transuranic Wastes—Notice of Public Hearings, Federal
Register, 48(63):13444-13446, Thursday, March 31, 1983.

Environmental Protection Agency, Environmental Standards for
the Management and Disposal of Spent Nuclear Fuel, High-Level
and Transuranic Radioactive Wastes—Request for Post-Hearing
Comments, Federal Register, 48(103):23666, Thursday, May 26,
1983.

Environmental Protection Agency, Science Advisory Board - Open
Meeting; High-Level Radioactive Waste Disposal Subcommittee,
Federal Register, 48(3):509, Wednesday, January 5, 1983.

Environmental Protection Agency, Environmental Standards for
the Management and Disposal of Spent Nuclear Fuel, High-Level
and Transuranic Radioactive Wastes—Notice of Availability,
Federal Register, 49(90):19604-19606, Tuesday, May 8, 1984.

Federal Energy Resources Council, Management of Commercial
Radioactive Nuclear Wastes - A Status Report, May 10, 1976.
                                     1-1L

-------
 Fo76      The White House,  President G.  Ford,  President's Nuclear Waste
           Management Plan,  Fact Sheet,  October 28,  1976.

 Ly76      Memorandum from J.  T. Lynn,  OMB to R.  Train,  EPA;  R.  Peterson,
           CEQ; R.  Seamans,  ERDA,  and W.  Anders,  NRC;  March 25,  1976,
           concerning Establishment  of an Interagency  Task Force on Com-
           mercial  Nuclear Wastes.

 Mc70      McClain  W.  C.  and R.  L. Bradshaw,  Status  of Investigations  of
           Salt Formations for Disposal  of Highly Radioactive
           Power-Reactor  Wastes, Nuclear  Safety,  11(2);130-141,
           March-April 1970.

 NAS57      National Academy  of Sciences - National Research Council,
           Disposal of Radioactive Wastes on  Land, Publication 519,
           Washington,  D.C., 1957.

 NAS70      National Academy  of Sciences - National Research Council,
           Committee on Radioactive Waste Management,  Disposal of  Solid
           Radioactive Wastes  In Bedded Salt  Deposits, Washington,  D.C.,
           November 1970.

           The  White House, President R.  Nixon, Reorganization Plan No.  3
           of  1970,  Federal Register, 35(194):15623-15626,  October  6,
           1970.

           Nuclear  Waste Policy  Act of 1982,  Public Law 97-425,  January  7,
           1983.

           Reeves M. and R. M.  Cranwell,  User's Manual for  the Sandia
           Waste-Isolated Flow and Transport Model (SWIFT)  Release  4.81,
           Sandia National Laboratories,  Report SANDS1-2516
           (NUREG/CR-2324), November 1981.

SAB84     Report on the Review  of Proposed Environmental Standards for
           the Management and Disposal of Spent Nuclear Fuel, High-level
          and Transuranic Radioactive Wastes (40 CFR  191), by the High-
          Level Radioactive Waste Disposal Subcommittee, Science Advisory
          Board, USEPA, January 1984.

Sm85      Smith J.  M., T. W. Fowler and A. S. Golden,  Environmental
          Pathway Models for Estimating Population Health Effects From
          Disposal of High-level Radioactive Waste in Geologic Reposi-
          tories, U.S. E.P.A.  Report EPA-520/5-85-026, August 1985.
Ni70



NWPA83


Re81
                                     1-12

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         Chapter 2:  CURRENT REGULATORY PROGRAMS AND STRATEGIES
2.1  Introduction

     People have always been exposed to ionizing radiations from cosmic
rays and the naturally-occurring radionuclides in the earth that make up
the natural radiation background.  Awareness of radiation and radio-
activity dates back only to the end of the last century—to the discov-
ery of x-rays in 1895 and the discovery of radioactivity in 1896.  These
discoveries marked the beginning of radiation science and the deliberate
use of radiation and radionuclides in science, medicine, and industry.

     The findings of radiation science rapidly led to the development of
medical and industrial radiology, nuclear physics, and nuclear medicine.
By the 1920's, the use of x-rays in diagnostic medicine and industrial
applications was widespread, and radium was being used by industry for
luminescent dials and by doctors in therapeutic procedures.  By the
1930's, biomedical and genetic researchers were studying the effects of
radiation on living organisms, and physicists were beginning to under-
stand the mechanisms of spontaneous fission and radioactive decay.  By
the 1940's, a self-sustaining fission reaction was demonstrated, which
led directly to the construction of the first nuclear reactors and
atomic weapons.

     Developments since the end of World War II have been rapid.  Today
the use of x-rays and radioactive materials is widespread and includes:

     0    .Nuclear reactors, and their supporting fuel-cycle facilities,
          which generate electricity and power ships and submarines;
          produce radiolsotopes for research, space, defense, and medi-
          cal applications; and are used as research tools for nuclear
          engineers and physicists.

     0    Particle accelerators which produce radioisotopes and are used
          as research tools for studying the structure of materials and
          atoms.

          The radiopharmaceutical industry which provides the radioiso^-
          topes needed for biomedical research and nuclear medicine.
                                   2-1

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      0    Nuclear medicine which has developed as a recognized medical
           specialty in which radioisotopes are used in the diagnosis and
           treatment of numerous diseases.

      0    X-rays which are widely used as a diagnostic tool in medicine
           and in such diverse industrial fields as oil exploration and
           nondestructive testing.

           Radionuclides which are used in such common consumer products
           as luminous-dial wristwatches and smoke detectors.

      The following sections of this chapter provide a brief history of
 the evolution of radiation protection philosophy and an outline of the
 current regulatory programs and strategies of the government agencies
 responsible for assuring that radiation and radionuclides are used
 safely.

 2*2  The International Commission on Radiological Protection and the
      National Council on Radiation Protection and Measurements

      Initially,  the dangers and risks posed by x-rays and radioactivity
 were poorly understood.   By 1896,  however,  "x-ray burns" were being
 reported in the  medical  literature,  and by 1910,  it  was understood that
 such "burns" could  be caused by radioactive materials.  By the 1920's,
 sufficient  direct evidence  (from the experiences  of  radium dial painters,
 medical radiologists,  and miners)  and indirect evidence (from biomedical
 and genetic experiments  with animals)  had  been accumulated to persuade
 the scientific community that an official  body should be established  to
 make recommendations  concerning human protection  against exposure to
 x-rays  and  radium.

      At the Second  International Congress of Radiology meeting in
 Stockholm,  Sweden,  in  1928,  the first  radiation protection commission
 was created.   Reflecting the  uses  of radiation and radioactive materials
 at  the  time,  the body was named the  International X-Ray and Radium
 Protection  Commission and was charged with  developing  recommendations
 concerning  protection from  radiation.   In  1950, to reflect better its
 role  in a changing world, the commission was reconstituted and renamed
 the International Commission on  Radiation Protection  (ICRP).

     During the Second International Congress  of  Radiology,  the newly
 created  commission suggested to  the nations represented at  the Congress
 that they appoint national advisory  committees  to represent  their view-
points before the ICRP, and to act in concert with the Commission in
developing and disseminating recommendations on radiation  protection.
This suggestion led to the formation, in 1929, of the Advisory Committee
on X-Ray and Radium Protection as the U.S. advisory group.  This  Advis-
ory Committee, after a series of reorganizations and name  changes,
emerged in 1964 in its present form as the Congressionally-chartered
                                   2-2

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National Council on Radiation Protection and Measurements (NCRP).   The
Congressional charter provides for the NCRP to:

     0    Collect, analyze, develop, and disseminate in the public
          interest information and recommendations about radiation
          protection and radiation quantities! units, and measurements.

     0    Develop basic concepts about radiation protection and radia-
          tion quantities, units, and measurements, and the application
          of these concepts.

     0    Provide a means by which organizations concerned with radia-
          tion protection and radiation quantities, units, and measure-
          ments may cooperate to effectively use their combined re-
          sources, and to stimulate the work of such organizations.

     0    Cooperate with the ICRP and other national and international
          organizations concerned with radiation protection and radia-
          tion quantities, units, and measurements.

Throughout their existence, the ICRP and the NCRP have worked together
closely to develop radiation protection recommendations that reflect the
current understanding of the dangers associated with exposure to ioniz-
ing radiation.

     The first exposure limits adopted by the  ICRP and the NCRP (ICRP34,
ICRP38, NCRP36) established 0.2 roentgen/day*  as the "tolerance dose"
for occupational exposure to x-rays and gamma  radiation from radium.
This limit, equivalent to approximately 25 rads/year as measured in air,
was established to guard against the known effects of ionizing radiation
on superficial tissue, changes in the blood, and "derangement" of  inter-
nal organs, especially the reproductive organs.  At  the time the recom-
mendations were made, high doses of radiation  were known  to cause  obser-
vable effects and even to induce cancer.  However, no such effects were
observed at lower doses, and the epidemiological evidence at the time
was inadequate to even imply the carcinogenic  induction effects of
moderate or low doses.  Therefore,  the aim of  radiation protection was
to guard against known effects, and the "tolerance dose"  limits that
were adopted were believed  to represent the  level of radiation that a
person in normal health could tolerate without suffering  observable
effects.  The concept of a  tolerance dose and  the recommended occupa-
tional exposure limit of 0.2 R/day  for x- and  gamma  radiation remained
in effect until the  end of  the  1940's.  The  recommendations of the ICRP
and the NCRP made no mention of exposure of  the general populace.

     By the end of World War II,  the widespread use  of  radioactive
materials and scientific evidence of  genetic and  somatic  effects  at
lower doses and dose rates  suggested  that  the radiation protection
   The  NCRP's  recommendation was 0.1  roentgen(R)/day measured in air.
   This limit  is  roughly equivalent to the ICRP limit,  which was conven-
   tionally measured at the point of  exposure and included back-scatter.
                                    2-3

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recommendations of the NCRP and the ICRP would have to be revised down-
ward.

     By  1948, the NCRP had formulated its position on appropriate new
limits.  These limits were largely accepted by the ICRP in its recommen-
dations  of  1950 and formally issued by the NCRP in 1954 (ICRP51, NCRP54).
The immediate effect was to lower the basic whole-body occupational dose
limit to 0.3 rad/week (approximately 15 rads/year); the revised recom-
mendations  also embodied several new and important concepts in the
formulation of radiation protection criteria.

     First, the recommendations recognized the differences in the ef-
fects of various types and energies of radiation; both ICRP's and NCRP's
recommendations included discussions of the weighting factors that
should be applied to radiations of differing types and energies.  The
NCRP advocated the use of the "rem" to express the equivalence in bio-
logical  effects between radiations of differing types and energy.*
Although the ICRP noted the shift toward the acceptance of the rem, it
continued to express its recommendations in terms of the rad, with the
caveat that neutrons should carry a quality factor of ten.

     Second, the recommendations of both organizations introduced the
concept  of  critical organs and tissues.  The intent of this concept was
to assure that no tissue or organ, with the exception of the skin, would
receive  a dose in excess of that allowed for the whole body.  At the
time, scientific evidence was lacking on which to base different recom-
mended limits for the various tissues and organs.  Thus, all blood-
forming  organs were considered critical organs and were limited to the
same exposure as the whole body.  The skin was allowed an exposure of 30
rads/year and the extremities were allowed a limit of 75 rads/year.

     Third, the recommendations of the NCRP included the suggestion that
individuals under the age of 18 receive no more than one-tenth the expo-
sure allowed for adults.  The reasoning behind this particular recommen-
dation is interesting as it reflects clearly the limited knowledge of
the times.   The scientific evidence indicated a clear relationship
between  accumulated dose and genetic effect.  However, this evidence was
obtained exclusively from animal studies that had been conducted with
doses ranging from 25 to thousands of rads.  There was no evidence from
exposures less than 25 rads accumulated dose, and the interpretation of
* The exact relationship between roentgens, rads, and rems is beyond the
  scope of this work.  In simple terms, the roentgen is a measure of the
  degree of ionization induced by x- and gamma radiations in air.  The
  rad (radiation absorbed dose) is a measure of the energy imparted to
  matter by radiation.  And the rem (roentgen equivalent man) is a mea-
  sure of equivalence for the relative biological effect of radiations
  of different types and energies on man.  Over the range of energies
  typically encountered, the relationship of roentgens to rads to rems
  for x- and gamma radiation is essentially equality.   For beta radia-
  tion, rads are equivalent to rems, and for alpha radiation one rad
  equals 10 to 20 rems.

                                   2-4

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 the animal data and the implications  for humans was  unclear and did not
 support a specific permissible dose.   The data did suggest that genetic
 damage was more dependent on accumulated dose  than previously believed*
 but experience showed that exposure for prolonged periods to the permis-
 sible dose (1.0 R/week) did not result in any  observable genetic ef-
 fects.  The NCRP decided that it was  not necessary to  change the occupa-
 tional limit to provide additional protection  beyond that provided by
 the reduction in the permissible dose limit to 0.3 R/week.  At the same
 timei it recommended limiting the exposure of  individuals under the age
 of 18 to assure that they did not accumulate a genetic dose that would
 later preclude their employment as radiation workers.   The factor of ten
 was rather arbitrary, but was believed to be sufficient to protect the
 future employability of all individuals (NCRP54).

      Fourth* the concept of a tolerance dose was  replaced by the concept
 of a maximum permissible dose.  The change in  terminology reflected the
 increasing awareness that any radiation exposure might involve some risk
 and that repair mechanisms might be less effective  than previously be-
 lieved.  Therefore* the concept of a  maximum permissible dose was adopted
 because it better reflects the uncertainty in  our knowledge  than does
 the concept of tolerance dose.  The maximum permissible dose was defined
 as the level of exposure that entailed a small risk compared with those
 posed by other hazards in life (ICRP51).

      Finally* in explicit recognition of the inadequacy of  our knowledge
 regarding the effects of radiation and of the  possibility  that any  expo-
 sure might have some potential for harm, the recommendations included an
 admonition that every effort should be made to reduce exposure  to all
 kinds of ionizing radiation to the lowest possible level.   This  concept*
 known originally as ALAP (as low as practicable)  and later as ALARA (as
 low as reasonably achievable), would become a cornerstone of radiation
 protection philosophy.

      During the 1950's, a great deal of scientific evidence on the
 effects of radiation became available from studies of the radium dial
 painters, radiologists, and the survivors of the atomic bombs dropped on
 Japan.  This evidence suggested that genetic effects and long-term
 somatic effects were more important than previously considered.   Thus*
 by the late 1950's, the ICRP and NCRP recommendations were  again revised
 (ICRP59, NCRP59).  These revisions included the following major  changes:
 the annual maximum permissible dose for whole-body exposure and  the most
 critical organs (blood-forming organs, gonads, and the lens  of  the  eye)
 waa lowered to 5 rems, with a quarterly limit  of  3 rems; the limit  for
 exposure of other organs was set at 30 rems/year; internal  exposures
 vere controlled by a comprehensive set of maximum permissible concentra-
 tions of radionuclides in air and water based  on the most  restrictive
 case of a young worker; and recommendations vere Included  for some
 rtonoccupational groups and for the general population (for  the first
 time).

     The lowering of the annual maximum  permissible  whole-body dose to 5
reme, with a quarterly llait of 3 rems, reflects both  the new evidence
                                   2-5

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and the uncertainties of  the  time.  Although no adverse effects were
observed  among workers who had  received  the earlier maximum permissible
dose of 0.3 rad in a week, there was concern that the lifetime accumu-
lation of as much as 750  rads (15 rads/year times 50 years) was too
much.  Lowering the maximum permissible  dose by a factor of three was
believed  to provide a greater margin of  safety.  At the same time,
operational experience showed that an annual dose of 5 rems could be met
in most instances, particularly with the additional operational
flexibility provided by expressing the limit on an annual and quarterly
basis.

     The  recommendations  given  for nonoccupational exposures were based
on concerns of genetic effects.  The evidence available suggested that
genetic effects were primarily  dependent on the total accumulated dose.
Thus, having sought the opinions of respected geneticists, the ICRP and
the NCRP  adopted the recommendation that accumulated gonadal dose to age
30 be limited to 5 rems from sources other than natural background and
medical exposure.  As an  operational guide, the NCR? recommended that
the maximum annual dose to any  individual be limited to 0.5 rem, with
maximum permissible body burdens of radionuclides (to control internal
exposures) set at one-tenth that allowed for radiation workers.  These
values were derived from consideration of the genetically significant
dose to the population, and were established "primarily for the purpose
of keeping the average dose to the whole population as low as reasonably
possible, and not because of the likelihood of specific injury to the
individual" (NCRP59).

     During the 1960's, the ICRP and NCRP again lowered the maximum
permissible dose limits (ICRP65, NCRP71).  The considerable scientific
data on the effects of exposure to ionizing radiation were still incon-
clusive with respect to the dose-response relationship at low exposure
levels; thus, both organizations continued to stress the need to keep
all exposures to the lowest possible level.

The NCRP and the ICRP  made the following similar recommendations:

     0     Limit the dose to the  whole-body, red bone marrow,  and gonads
          to 5 rems in any year, with a retrospective limit of 10 to 15
          rems in any  given year as long as total accumulated dose did
          not exceed  5x(N-18), where N is age in years.

     e     Limit the annual dose  to the skin,  hands,  and  forearms to 15,
          75,  and  30 rems, respectively.

     0     Limit the annual dose  to any other  organ or tissue  to 15 rems.

     0     Limit the annual dose  to any nonoccupationally exposed
          individual in the population to 0.5 rem.

     0     Limit the annual average dose to the  population to  0.17  rem.
                                   2-6

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     The scientific evidence and the protection philosophy on which the
above recommendations were based were set forth in detail in NCRP71.  In
the case of occupational exposure limits, the goal of protection was to
ensure that the risks of genetic and somatic effects were small enough
to be comparable to the risks experienced by workers in other indus-
tries.  The conservatively derived numerical limits recommended were
based on the linear, nonthreshold, dose-response model, and were be-
lieved to represent a level of risk that was readily acceptable to an
average individual.  For nonoccupational exposures, the goal of protec-
tion was to ensure that the risks of genetic or somatic effects were
small compared with other risks encountered in everyday life.  The deri-
vation of specific limits was complicated by the unknown dose-response
relationship at low exposure levels and the fact that the risks of radi-
ation exposure did not necessarily accrue to the same individuals who
benefited from the activity responsible for the exposure.  Therefore, it
was necessary to derive limits that gave adequate protection to each
member of the public and to the gene pool of the population as a whole,
while still allowing the development of beneficial uses of radiation and
radionuclides.

     In 1977, the ICRP made a fundamental change in its recommendations
when it abandoned the critical organ concept in favor of the weighted
whole-body dose equivalent concept for limiting occupational exposure
(ICRP77).  The change, made to reflect our increased understanding of
the differing radiosensitivity of the various organs and tissues, did
not affect the overall limit of 5 rems per year and is not intended to
be applied to nonoccupational exposures.

     Also significant is the fact that ICRP's 1977 recommendations
represent the first explicit attempt to relate and justify permissible
radiation exposures with quantitative levels of acceptable risk.  Thus.
the risks of average occupational exposures (approximately 0.5 rem/year)
are equated with risks in safe industries, given as  It)"1* annually.  At
the maximum limit of 5 rems/year, the risk is equated with that expe-
rienced by some workers in recognized hazardous occupations.  Similarly,
the risks implied by the nonoccupational limit of 0.5 rem/year are
equated to levels of risk of less than 10~a in a lifetime; the general
populace's average exposure is equivalent to a lifetime risk on the
order of 10 s to 10"1*.  The ICRP believed these levels of risk were in
the range that most individuals find acceptable.

     The NCRP has not formally changed its recommendations for occupa-
tional exposure to correspond to the 1977 recommendations of the  ICRP.
It has been working diligently, however, to review its recommendations,
and has circulated a draft of proposed changes to various interested
scientists and regulatory bodies for their comment.  The relevant non-
occupational exposure limits are:

     0    0.5 rem/year whole-body dose equivalent, not including  back-
          ground or medical radiation, for individuals in the population
          when the exposure is not continuous.
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      0     0.1  rem/year whole-body  dose  equivalent, not  including back-
           ground  or medical  radiation,  for  individuals  in the population
           when the exposure  is  continuous.

      0     Continued use of a total dose limitation system based on
           justification of every exposure and application of the ALAKA
           philosophy  to every exposure.

      The NCR?  equates continuous exposure at the level  of 0.1 rem/year
 to a  lifetime  risk of developing cancer of  about one in a thousand.  The
 NCRP  has not formulated exposure limits for specific organs, but it
 notes that the permissible limits  will  necessarily be higher than the
 whole-body limit  in inverse  ratio  of the risk for a particular organ to
 the total  risk for whole-body exposure.

 2.3   Federal Guidance

      The ICRP  and the NCRP function as nongovernmental  advisory bodies.
 Their recommendations are not binding on any user of radiation or radio-
 active materials.   The wealth of new scientific information on the
 effects of radiation that became available  in the 1950's prompted Presi-
 dent  Eisenhower to establish an official government entity with responsi-
 bility for formulating radiation protection criteria and coordinating
 radiation protection activities.  Thus, the Federal Radiation Council
 (FRC) was established in 1959 by Executive Order 10831.  The Council
 included representatives from all of the Federal agencies concerned with
 radiation protection, and acted as a coordinating body  for all of the
 radiation activities conducted by the Federal government.  In addition
 to its coordinating function, the Council's major responsibility was to
 "t..advise the President with respect to radiation matters, directly or
 indirectly affecting health,  including guidance for all Federal agencies
 in the formulation of radiation standards and in the establishment and
 execution of programs of cooperation with States..." (FRC60).

     The Council's first recommendations concerning radiation protection
standards for Federal agencies were approved by the President in 1960.
Based largely on the work and recommendations of the ICRP and the NCRP,
the guidance established the  following limits for occupational expo-
sures :

     0    Whole body, head and trunk, active blood forming organs,
          gonads,  or lens of  eye—not to exceed 3 rems in 13 weeks and
          total accumulated dose limited to 5 times the number of years
          beyond age 18.

     0    Skin of  whole body  and thyroid—not to exceed 10 rems  in 13
          weeks or 30 rems per year.

     0    Hands, forearms, feet, and ankles—not to exceed 25 rems in 13
          weeks or 75 rems per year.
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     0    Bone—not to exceed 0.1 microgram of radium-226 or its  biolog-
          ical equivalent.

          Any other organ—not to exceed 5 rems per 13 weeks or 15 rems
          per year.

     Although these levels differ slightly from those recommended by
NCRP and ICRP at the time,  the differences do not represent any greater
or lesser protection.  In fact, the FRC not only accepted the levels
recommended by the NCRP for occupational exposure, it adopted the NCRP's
philosophy of acceptable risk for determining occupational exposure
limits.  Although quantitative measures of risk were not given in the
guidance, the prescribed levels were not expected to cause appreciable
bodily injury to an individual during his or her lifetime.  Thus, while
the possibility of some injury was not zero, it was so low as to be ac-
ceptable if there was any significant benefit derived from the exposure.

     The guidance also established exposure limits for members of the
public.  These were set at 0.5 rem per year (whole body) for an indivi-
dual, and an average of 5 rems in 30 years (gonadal) per capita.   The
guidance also provided for developing a suitable sample of the popula-
tion as an operational basis for determining compliance with the limit
when doses to all individuals are unknown.  Exposure to this population
sample was not to exceed 0.17 rem per capita per year.  The population
limit of 0.5 rem to any individual per year, was derived from considera-
tion of natural background exposure.

     In addition to the formal exposure limits, the guidance also estab-
lished as Federal policy that there should be no radiation exposure
without an expectation of benefit, and that "every effort should be made
to encourage the maintenance of radiation doses as far below this guide
as practicable."  The inclusion of the requirements to consider benefits
and keep all exposure to a minimum was based on the possibility that
there is no threshold dose for radiation.  The linear nonthreshold dose
response was assumed to place an upper limit on the estimate of radia-
tion risk. -However, the FRC explicitly recognized that it might also
represent the true level of risk.  If so, then any radiation exposure
carried some risk, and it was necessary to avoid all unproductive expo-
sures and to keep all productive exposures as "far below this guide as
practicable."

     In 1967, the Federal Radiation Council issued guidance for the
control of radiation hazards in uranium mining  (FRC67).  The need for
such guidance was clearly indicated by the epidemiological evidence that
showed a higher Incidence of lung cancer in adult males who worked in
uranium mines compared with the incidence in adult males from the same
locations who had not worked in mines.  The guidance established  specif-
ic exposure limits and recommended that all exposures be kept as  far
below the guide limits as possible.  The limits chosen  represented a
trade-off between the risks incurred at various exposure  levels,  the
technical feasibility of reducing the exposure, and  the benefits  of  the
                                   2-9

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 activity responsible for the exposure.   The guidance also applied to
 nonuranium mines.

      In 1970,  the  functions of the Federal Radiation Council were trans-
 ferred to the  U.S.  Environmental Protection Agency (EPA).  In 1971,  the
 EPA revised the Federal guidance for the control of radiation hazards  in
 underground uranium mining (EPA71).  Based on the risk levels associated
 with the exposure  limits established in 1967, the upper limit of  expo-
 sure was reduced by a factor of three.   The EPA has also provided fed-
 eral guidance  for  the diagnostic use of x-rays (EPA78).  This guidance
 established maximum skin entrance doses for various types of routine
 x-ray examinations.  It also established the requirement that all x-ray
 exposures be based  on clinical indication and diagnostic need,  and that
 all exposure of patients should be kept as low as reasonably achievable
 consistent with the diagnostic need.

      In 1981,  the EPA proposed new Federal guidance for occupational
 exposures to supersede the 1960 guidance (EPA81).   The 1981 recommended
 guidance follows the principles set forth by the  ICRP in 1977,  with  re-
 spect to combining  internal and external doses.   The basic occupational
 limit suggested in  the guidance is 5 rems per year.   This recommended
 guidance has not yet been adopted as Federal policy.   The proposals  in
 the guidance were issued for public comment in 1981  and are currently
 being reviewed and  revised in light of  the comments received.

 2.4  The Environmental Protection Agency

      In addition to the statutory responsibility  to provide Federal
 guidance on radiation protection,  the EPA has various statutory author-
 ities and responsibilities regarding regulation of exposure to  radi-
 ation.   The standards and the regulations that EPA has promulgated and
 proposed with  respect to controlling radiation exposures are summarized
 here.

      The U.S.  Atomic Energy Act  of  1954,  as amended,  and Reorganization
 Plan  No.  3  granted  EPA the authority  to establish  generally applicable
 environmental  standards for exposure  to radionuclides.   Pursuant  to  this
 authority,  in  1977  the EPA issued standards limiting exposure from oper-
 ations  of  the  light-water reactor nuclear fuel cycle  (EPA77b).  These
 standards  cover  normal operations of  the  uranium fuel cycle,  excluding
mining  and  waste disposal.   The  standards limit the  annual dose equiva-
 lent  to  any  member  of  the public from all phases of  the uranium fuel
 cycle  (excluding radon and its daughters)  to  25 mrems to the whole body,
 75 mrems  to  the  thyroid,  and 25  mrems to  any  other organ.   To protect
against  the  buildup  of  long-lived  radionuclides in the environment,  the
 standard also  sets normalized emission  limits for  krypton-85, iodine-
 129, and plutonium-239  combined  with  other  transuranics with a  half-life
exceeding one year.   The  dose limits  imposed  by the  standard cover all
exposures resulting  from radiation  and  radionuclide  releases to air  and
water from  operations  of  fuel-cycle  facilities.
                                   2-10

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     The development of this standard took into account both the maximum
risk to an individual and the overall effect of releases from fuel-cycle
operations on the population and balanced these risks against the  costs
of effluent control in a primarily qualitative way.

     Under the authority of the Uranium Mill Tailings Radiation Control
Act, the EPA promulgated standards limiting public exposure to radiation
and restricting releases of materials from uranium tailings piles
(EPA83a).  Cleanup standards for land and buildings contaminated with
residual radioactive materials from inactive uranium processing sites
were also established.  In these actions, the Agency sought to balance
the radiation risks imposed on individuals and the population in  the
vicinity of the pile against the feasibility and costs of control.

     The Agency first established regulations and criteria for the
disposal of radioactive waste into the oceans in 1973 under the author-
ity of the Marine Protection, Research and Sanctuaries Act of 1972.
These regulations (40 CFR Parts 220-229), which were revised in 1977,
prohibit ocean disposal of high-level radioactive wastes and
radiological warfare agents and establish requirements for obtaining
ocean disposal permits for other radioactive waste (EPA77a).

     In 1982, EPA issued effluent limitations guidelines for the  ore
mining and dressing point source category under the Clean Water Act.
Subpart C - Uranium, Radium and Vanadium Ores Subcategory of 40 CFR
Part 440 limits, among other items, the concentrations of radium and
uranium in effluent discharges from such mines and prohibits the dis-
charge of process wastewater from uranium mills in dry climates.

     Under the authority of the Safe Drinking Water Act, the EPA issued
interim regulations covering the permissible levels of radium, gross
alpha, manmade beta, and photon-emitting contaminants in community water
systems (EPA76).  The limits are expressed in picocuries/liter.  The
limits chosen for manmade beta- and photon-emitters equate to approxi-
mately 4 mrems/year whole-body or organ dose to the most exposed indi-
vidual.  In the background information for the standard, the 4 mrems/
year exposure through a single pathway that the standard permits is
explicitly compared with the overall population standard of 170 mrems/
year, and the conclusion is expressed that the roughly 40-fold decrease
is appropriate for a single pathway.

     Section 122 of the Clean Air Act amendments of 1977 (Public Law
95-95) directed the Administrator of EPA to review all relevant informa-
tion and determine if emissions of hazardous pollutants into air will
cause or contribute to air pollution that may reasonably be expected to
endanger public health.  In December 1979, EPA designated radionuclides
as hazardous air pollutants under Section  112 of the Act.  On February
6, 1985, and April 17, 1985, EPA published Kational Emission Standards
for radionuclides for selected sources  (EPA85a, 85b).

     In 1982, under the authority of the U.S. Atomic Energy Act of  1954,
as amended, the EPA proposed standards for disposal of  spent fuel,  high-
level wastes, and transuranic elements  (EPA82).  The proposed  standards

                                   2-11

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establish two different limits:   (1) during the active waste-disposal
phase, operations at the repository must be conducted so that no member
of the public receives a dose greater than that allowed for other phases
of the uranium fuel cycle; and  (2) once the repository is closed, expo-
sure is to be controlled by limiting releases.  The release limits were
derived by summing, over long time periods, the estimated risks to all
persons exposed to radioactive materials released into the environment.
The uncertainties involved in estimating the performance of a theoret-
ical repository led to this unusual approach, and the proposed standards
admonish the agencies responsible for constructing and operating such
repositories to take steps to reduce releases below the upper bounds
given in the standards to the extent reasonably achievable.

2.5  Nuclear Regulatory Commission

     Under the authority of the Atomic Energy Act of 1954, as amended,
the U.S. Nuclear Regulatory Commission (NRC) is responsible for licen-
sing and regulating the use of byproduct, source, and special nuclear
material, and for assuring that all licensed activities are conducted in
a manner that protects public health and safety.  The Federal guidance
on radiation protection applies directly to the NRC; therefore, the NRC
must assure that none of the operations of its licensees exposes an
individual of the public to more than 0.5 rem/year from all pathways.
The dose limits imposed by the EPA's standard for uranium fuel-cycle
facilities (40 CFR Part 190) also apply to the fuel-cycle facilities
licensed by the NRC.  These facilities are prohibited from releasing
radioactive effluents in amounts that would result in doses greater than
the 25 mrems/year limit imposed by that standard.

     Also NRC facilities are required to operate in accordance with the
requirements of the Clean Air Act (40 CFR Part 61), which limits radio-
nuclide emissions to air to that amount which will cause a dose equiva-
lent of 25 mrems/year to the whole body or 75 mreos/year to the critical
organ of any member of the public.

     The NRC exercises its statutory authority by imposing a combination
of design criteria, operating parameters, and license conditions at the
time of construction and licensing.  It assures that the license condi-
tions are fulfilled through inspection and enforcement.  The NRC licenses
more than 7000 users of radioactivity.

2.5.1  Fuel Cycle Licenses

     The NRC does not use the term "fuel cycle facilities" to define its
classes of licensees.  The term is used here to coincide with the EPA
use of the term in its standard for uranium fuel cycle facilities.  As a
practical matter, this term includes the NRC's large source and special
nuclear material, and production and utilization facilities.  The NRC's
regulations require an analysis of probable radioactive effluents and
their effects on the population near fuel cycle facilities.  The NRC
also assures that all exposures are as low as reasonably achievable by
imposing design criteria and specific equipment requirements on the
licensees.  After a license has been issued, fuel-cycle licensees must

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monitor their emissions and take environmental measurements to assure
that the design criteria and license conditions have been met.  For
practical purposes, the NRC adopted the maximum permissible concen-
trations developed by the NCRP to relate effluent concentrations to
exposure.

     In the 1970's, the NRC formalized the implementation of as low as
reasonably achievable exposure levels by issuing a regulatory guide for
as low as reasonably achievable design criteria.  This coincided with a
decision to adopt, as a design criterion, a maximum annual permissible
dose of 5 mrems from a single nuclear electric generating station.  The
5-mrem limit applies to the most exposed individual actually living in
the vicinity of the reactor, and refers to whole-body doses from ex-
ternal radiation by the air pathway (NRC77).

2.5.2  Radioactive Waste Disposal Licenses

     The NRC1 s requirements for radioactive waste disposal are contained
in 10 CFR Part 60, Disposal of High-Level Radioactive Wastes in Geologic
Repositories:  Technical Criteria (NRC83); 10 CFR Parts 2, 19, 20, 21,
30, 40, 51, 60, and 70, Disposal of High-Level Radioactive Wastes in
Geologic Repositories:  Licensing Procedures  (NRC81); 10 CFR Part 61,
Licensing Requirements for Land Disposal of Radioactive Waste  (NRC82);
and 10 CFR Part AO, Uranium Mill Licensing Requirements (NRC80).  NRC is
expected to make certain revisions to 10 CFR Part 60 to bring  them into
full consistency with the AO CFR Part 191 issued by EPA.

2.6  Department of Energy

     The U.S. Department of Energy  (DOE) operates a complex of national
laboratories and weapons facilities.  These facilities are hot licensed
by the NRC.  Under the U.S. Atomic Energy Act of  195A, as amended, the
DOE is responsible for keeping radionuclide emissions at these facil-
ities as low as reasonably achievable (ALARA).  The EPA has promulgated
a final standard, consistent With the requirements of the Clean Air  Act,
that limits radionuclide air emissions from DOE facilities to  that
amount which will cause a dose equivalent of  25 mrems/year to  the whole
body or 75 mrems/year to the critical organ of any member of  the public.
These limits generally reflect current emission levels achieved by
existing control technology and operating practices at DOE facilities
(EPA85a).

     For practical purposes, the DOE has adopted  the NCRP's maximum
permissible concentrations in air and water as a workable way to assure
that the annual dose limits of 0.5  rem whole-body and 1.5 rems to  any
organ are being observed.  The DOE  also has a requirement that all doses
be kept as low as is reasonably achievable, but the contractors  that op-
erate the various DOE sites have a  great deal of  latitude  in implement-
ing policies and procedures to assure that  all  doses are  kept to the
lowest possible level.
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      The DOE assures that its operations are within its operating guide-
 lines by requiring its contractors to maintain radiation monitoring
 systems around each of its sites and to report the results in an annual
 summary report.  New facilities and modifications to existing facilities
 are subject to extensive design criteria reviews and require the prepara-
 tion of environmental impact statements pursuant to the National Environ-
 mental Policy Act of 1970 (NEPA70).  Since the mid-1970's, the DOE
 initiated a systematic effluent-reduction program that resulted in the
 upgrading of many facilities and effected a corresponding reduction in
 the effluents (including airborne and liquid radioactive materials)
 released to the environment.

      The DOE has developed and issued general guidelines in 10 CFR Part
 960 (DOE84) for the recommendation of sites for the disposal of high-
 level radioactive waste and spent nuclear fuel in geologic formations.
 The guidelines are to be used in various steps of the site selection
 process and are to be compatible with the regulations issued by the NRC
 in 10 CFR Part 60 and by the standards issued by the EPA in 40 CFR
 Part 191.   These guidelines establish performance objectives for a
 geologic repository system,  define the basic technical requirements that
 candidate sites must meet,  and specify how the DOE will implement the
 site-selection process.

 2.7  Department of Transportation

      The U.S.  Department of  Transportation (DOT)  has statutory responsi-
 bility for regulating the shipment and transportation of radioactive
 materials.   This authority  includes the responsibility to protect the
 public from exposure to  radioactive materials while they are in transit.
 For practical  purposes,  the  DOT  has implemented its authority through
 the specification of performance standards for shipment containers,  and
 by setting maximum exposure  rates from any package containing radioac-
 tive materials.   These limits  were set to assure  compliance  with the
 Federal guidance for occupational exposure,  and they are believed to be
 sufficient  to  protect the public from exposure.   The DOT also controls
 potential  public exposure by managing the routing of radioactive ship-
 ments  to avoid densely populated areas.

 2.8  State  Agencies

      States have  important authority  for  protecting the  public  from the
 hazards associated with  ionizing radiation.   Twenty-six  States  have
 assumed NRC's  inspection, enforcement,  and licensing responsibilities
 for users of source  and  byproduct materials  and users  of small  quanti-
 ties of special nuclear  material.   These  "NRC-agreement  States,"  which
 license and regulate more than 11,500 users  of  radiation and  radioactive
materialst are bound by  formal agreements  to  adopt  requirements consis-
 tent with those  imposed  by the NRC.   The NRC  continues  to perform this
 function for all  licensable uses of source, byproduct, and special
nuclear material  in  the  24 States  that are not agreement  States.
                                   2-14

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     State and public participation in the planning and development of
high-level waste repositories is essential in order to promote public
confidence in the safety of disposal of these wastes.  States which are
identified by the Secretary of Energy as having one or more acceptable
sites for a high-level waste repository may disapprove the site designa-
tion and submit to the Congress a notice of disapproval (NWPA83).   This
notice must be accompanied by a statement of reasons explaining why the
recommended repository site has been disapproved.

     Grants are available to States with acceptable sites so that a
State may:  1) determine potential economic, social, public health and
safety, and environmental impacts of the repository on the State and its
residents; 2) develop a request for impact assistance; 3) engage in any
monitoring, testing, or evaluation activities with respect to site
characterization programs; 4) provide information to its residences with
respect to the site; and 5) request information from, and make comments
and recommendations to, the Secretary of Energy with respect to the site
(NWPA83).

     After construction of a high-level waste repository is authorized,
financial and technical assistance will be provided  to the State by the
Secretary of Energy to mitigate the impact of the development of the
repository on the State.  The State must provide a report on economic,
social, public health and safety, and environmental  impacts that are
likely as a result of the development of a repository at the specified
site.  The State will be notified of the transportation of any high-
level radioactive waste or spent fuel that is brought into the  State  for
disposal at the repository site and can conduct  reasonable independent
monitoring and testing of activities on the  repository site  (NWPA83).

2.9  Indian Tribes

     If a recommended high-level radioactive waste  repository  site is
located on the reservation of an Indian tribe, the  tribe may disapprove
the site designation and  submit to Congress  a notice of disapproval.   As
with the State, grants are available to affected Indian tribes  so  that
they may:  1) determine any potential economic,  social, public  health
and safety, and environmental impacts of  the repository on the  reserva-
tion and its residents; 2) develop a request for impact assistance;  3)
engage in any monitoring,  testing, or evaluation activities with respect
to site characterization  programs; 4) provide  information  to  the resi-
dents of the reservation  with respect to  the site;  and  5)  request  infor-
mation from, and make  comments  and recommendations  to,  the Secretary of
Energy with respect  to the site (NWPA83).
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                                REFERENCES
 DOE84
 EPA71



 EPA76


 EPA77a


 EPA77b



 EPA78



 EPA81



 EPA82
EPA83a
EPA83b
 U.S. Department of Energy, Nuclear Waste Policy Act of 1982;
 General Guidelines for the Recommendation of Sites for the
 Nuclear Waste Repositories, TO CFR 960, Federal Register
 49.(236):47714-47770, Thursday, December 6, 1984.

 U.S. Environmental Protection Agency, Radiation Protection
 Guidance for Federal Agencies:  Underground Mining of Uranium
 Ore, Federal Register _3J3(132): 12921,  Friday,  July 9,  1971.

 U.S. Environmental Protection Agency, National Interim Primary
 Drinking Water Regulations, EPA-570/9-76-003, 1976.

 U.S. Environmental Protection Agency, Ocean Dumping,  Federal
 Register ^2_(7):2462-2490,  Tuesday,  January 11, 1977.

 U.S. Environmental Protection Agency, Environmental Radiation
 Protection Standards  for Nuclear Power Operations,  40 CFR 190,
 Federal  Register  42(9):2858-2861, Thursday, January 13,  1977.

 U.S. Environmental Protection Agency, Radiation  Protection
 Guidance to Federal Agencies  for Diagnostic X-Rays, Federal
 Register 43(22):4378-4380,  Wednesday,  February 1,  1978.

 U.S.  Environmental  Protection Agency,  Federal Radiation
 Protection  Guidance for Occupational  Exposure, Federal Reg-
 ister 46(15): 2836-2844, Friday*  January 23, 1981.

 U.S.  Environmental  Protection Agency,  Environmental Standards
 for  the  Management  and Disposal  of Spent Nuclear Fuel, High-
 Level and Transuranic Radioactive Wastes, 40  CFR 191, Federal
 Register  47(250):58196-58206,  Wednesday, December 29, 1982.

 U.S. Environmental  Protection Agency, Standards for Remedial
Actions at  Inactive Uranium Processing Sites, 40 CFR  192
Federal Register 48(3):590-604, Wednesday, January 5,  1983.

U.S. Environmental  Protection Agency, Environmental Standards
 for Uranium and Thorium Mill Tailings at Licensed Commercial
Processing Sites;  40 CFR 192 Federal Register 48_(196)  :45926-
45947, Friday, October 7,  1983.
                                   2-16

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EPA85a    U.S. Environmental Protection Agency, National Emission
          Standards for Hazardous Air Pollutants, Standards for
          Radionuclides, Federal Register j>()(25) :5190-5200, Wednesday,
          February 6, 1985.

EPA85b    U.S. Environmental Protection Agency, National Emission
          Standards for Hazardous Air Pollutants, Standard for Radon-222
          Emissions From Underground Uranium Mines, Federal Register
          5(K74):15386-15394, Wednesday, April 17, 1985.

FRC60     Federal Radiation Council, Radiation Protection Guidance for
          Federal Agencies, Federal Register 25(102):4402-4403, Wednes-
          day, May 18, 1960.

FRC67     Federal Radiation Council, Guidance for the Control of
          Radiation Hazards in Uranium Mining, Report No. 8, September
          1967.

ICRP34    International" X-Ray and Radium Protection Commission,
          International Recommendations for X-Ray and Radium Protection,
          British Journal of Radiology ]_, 695-699, 1934.

ICRP38    International X-Ray and Radium Protection Commission,
          International Recommendations for X-Ray and Radium Protection,
          Amer. Journal of Roent and Radium 40.  134-138,  1938.

ICRP51    International Commission on Radiological Protection,
          International Recommendations on Radiological Protection  1950,
          British Journal of Radiology  24, 46-53,  1951.

ICRP59    International Commission on Radiological Protection,  Recommen-
          dations of  the ICRP 1958,  ICRP Publication  1,  Pergamon Press,
          Oxford, 1959.

ICRP65    Internationa^ Commission on Radiological Protection,  Recommen-
          dations of  the ICRP 1965,  ICRP  Publication  9,  Pergamon Press,
          Oxford,  1965.

ICRP77    International Commission  on Radiological Protection,  Recommen-
          dations of  the International Commission on  Radiological Protec-
          tion,  ICRP  Publication 26, Pergamon Press,  Oxford,  1977.

NCRP36    National  Council on Radiation Protection and  Measurements,
          Advisory  Committee on X-ray and Radium Protection,  X-ray
          Protection, NCRP Report No.  3,  1936.

NCRP54    National  Council on Radiation Protection and  Measurements,
          Permissible Dose From External  Sources of Ionizing  Radiation,
          National  Bureau  of Standards  Handbook 59,  1954.
                                    2-17

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 NCRP59    National Council on Radiation Protection and Measurements,
           Maximum Permissible Body Burdens and Maximum Permissible
           Concentrations of Radionuclides in Air and in Water for
           Occupational Exposure,  National Bureau of Standards Handbook
           69,  1959.

 NCRP71    National Council on Radiation Protection and Measurements,
           Basic  Radiation Protection Criteria,  NCRP Report  No.  39,  1971.

 NEPA70    National Environmental  Policy Act of  1970,  Public Law 91-190,
           January 1,  1970.

 NRC77      U.S. Nuclear Regulatory Commission,  1977,  Appendix I:   10 CFR
           50, Federal Register 44,  September 26.  1979.

 NRC80      U.S. Nuclear Regulatory Commission, Uranium Mill  Licensing
           Requirements,  10  CFR 40,  Federal  Register  4£,  October  3,  1980.

 NRC81      U.S. Nuclear Regulatory Commission, Disposal of High-Level
           Radioactive Wastes  in Geologic  Repositories:   Licensing Pro-
           cedures, Federal  Register ^6_(37): 13971-13988,  Wednesday,
           February 25, 1985.

 NRC82     U.S. Nuclear Regulatory  Commission, Licensing  Requirements for
          Land Disposal of  Radioactive Waste, 10 CFR  61, Federal  Regis-
           ter 47_(248): 57446-57482, Monday, December 27,  1982.

 NRC83     U.S. Nuclear Regulatory Commission, Disposal of High-Level
          Radioactive Wastes  in Geologic Repositories, Technical  Cri-
          teria, 10 CFR 60, Federal Register 48  (120):28194-28229,
          Tuesday, June 21, 1983.

NWPA83    Nuclear Waste Policy Act of 1982, Public Law 97-425,
          January 7, 1983.
                                   2-18

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 Chapter 3:  QUANTITIES, SOURCES,  AND CHARACTERISTICS OF SPENT NUCLEAR
               FUEL AND HIGH-LEVEL AND TRANSURANIC WASTES
3.1  Introduction

     Presented in this chapter are current inventories of commercial
spent fuels, commercial and U.S. Department of Energy (DOE) high-level
radioactive wastes, and DOE transuranic wastes.  These inventories were
compiled from the most reliable government information.  Estimates of
generated wastes and spent fuel to the year 2000, based on the latest DOE
information and projected U.S. commercial power growth, are also present-
ed.  The spent fuel and wastes are characterized according to their
volumes (or masses) and their nuclear, physical, and chemical properties.

     The radioactive wastes and spent nuclear fuel originate from the
commercial nuclear fuel cycle and DOE defense-related activities.  The
wastes are broadly characterized as high-level waste (HLW) and trans-
uranic (TRU) waste.  In addition, an inventory of commercial reactor
spent fuel also may require an expansion of current storage or the con-
struction of additional facilities for interim storage, pending the
availability of commercial reprocessing facilities, permanent disposal
facilities, or monitored retrievable storage.

     Both spent fuel and high-level radioactive wastes from reprocessing
are intensely radioactive and generate substantial quantities of heat.
The radioactivity and heat production continue for long periods of time
because the wastes contain a number of long-lived radionuclides.  The
transuranium elements in particular have long radiological half-lives,
generate very little heat, and present a possible hazard to people for
tens of thousands of years.

3.2  Spent Nuclear Fuel (EPA82, DOE84a, Bu82, Li79, St79)

     In this standard, spent nuclear fuel is defined as fuel that has
been withdrawn from a nuclear reactor following irradiation and whose
constituent elements have not been separated by reprocessing.

     Spent fuel from government, industrial, and commercial sources can
be categorized as  1) fuel discharged from commercial light-water reactors
(LwR's); 2) fuel elements generated by government-sponsored research and
demonstration programs, universities, and industry; 3) fuels from experi-
mental reactors [viz., liquid metal fast breeder reactor  (LMFBR) and
                                     3-1

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 high-temperature gas-cooled reactors (HTGR)]; 4) U.S. Government-con-
 trolled nuclear weapon production reactors; and 5) naval reactor fuels
 and other Department of Defense (DOD) reactor fuels.

      Most (95 percent) of the spent fuels from commercial power reactors
 are stored at the reactor sites.  The re%t are stored at the Nuclear Fuel
 Services (NFS) plant at West Valley, New York, and at the Midwest Fuel
 Recovery Plant (MFRP) at Morris, Illinois.  The NFS plant is now being
 decommissioned, and the residual fuel quantities stored there are being
 transferred to other sites.  Special fuels are stored at the Savannah
 River Plant (SRP) in South Carolina and the Idaho Chemical Processing
 Plant (ICPP) in Idaho.  The LMFBR fuel from the Fast Flux Test Facility
 (FFTF) is stored at Hanford, Washington (HANF), and HTGR spent fuel
 discharged from the Fort St. Vrain reactor is stored at ICPP.  Production
 and naval reactor fuels are stored at SRP, ICPP, and Hanford, awaiting
 reprocessing by government-owned facilities.

      The fuel currently used in commercial light-water reactors consists
 of a mixture of uranium-238 and uranium-235 dioxides encased in zirconium
 alloy (zircaloy)  or stainless steel tubes.  During reactor operation,
 fission of the uranium-235 produces energy, neutrons, and fission products.
 The neutrons produce further fission reactions and thus sustain the chain
 reaction.  The neutrons also convert some of  the uranium-238 into pluto-
 nium-239, which can fission as uranium-235 can.   In time, the fissile
 uranium-235, which originally constituted some 3 to A 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.   Typical removal rate is one-
 third of the fuel,  or 30 tons/year per  reactor.   Reprocessing of com-
 mercial spent fuel has been proposed to  recover the unfissioned
 uranium-235 and the plutonium for  reuse  as a  fuel resource,  but such
 reprocessing is not currently taking place.

      The radioactive materials in  spent  fuel  fall into two major cate-
 gories:   fission  products and actinide elements.   Typically,  fresh spent
 fuel  contains more  than 100 radionuclides  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  an  atomic number greater than
 92, Including  plutonium,  curium, americium, and  neptunium 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  spent-fuel  sample depends on the reactor type,  the initial fuel
 compositioni  the  length of time the  fuel was  irradiated,  and  the elapsed
 time since its removal  from the reactor  core.

 3.2.1  Spent Fuel Inventory  and Projection (DOE84a)

     As of December 31,  1983,  there were 10,140 metric  tons  (t)  of  spent
fuel in inventory from  commercial  reactor  operation.   Of  this amount,  159

                                     3-2

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t are stored at the NFS facility and 322.5 t are stored at the MFRP.   The
remainder is at the reactor sites.  The oldest light-water-reactor spent
fuel in inventory was discharged in 1970.   The historical and projected
buildup of the spent fuel inventory and accumulated radioactivity are
given in Table 3.2-1.  These values do not include the relatively small
amount of spent fuel reprocessed by the NFS facility.


    Table 3.2-1,  Historical and projected mass and radioactivity of
                     commercial spent fuel (DOE84a)
End of
calendar
year
1970
1975
1980
1983
1985
1990
1995
2000
Mass
accumulated
(t)
28
1,449
6,496
10,140
12,449
21,121
31,559
42,812
Radioactivity
accumulated
(106 Ci)
134
4,057
10,236
12,879
13,178
23,176
29,456
35,674
     The activity of spent fuel depends primarily on its age.  As the
spent fuel ages, many of the short-lived fission products decay.  Calcu-
lations of waste activities 10 years after removal from the reactor, with
consideration being given only to radionuclides (fission products and
heavy elements) with half-lives greater than 20 years, show that the 1983
activity of the 10,400 t of spent fuel corresponds to about 1.6 billion
curies.

     The projected inventory of spent fuel (Table 3.2-1) was based on the
projected installed nuclear capacities given in Table 3.2-2.


        Table 3.2-2.  Historical and projected installed nuclear
                     electric power capacity (DOE84a)
End of
calendar
year
1960
1965
1970
1975
1980

Total
GW(e)
0.2
0.8
4.7
34.9
49.8
End of
calendar
year
1983
1985
1990
2000


Total
GW(e)
59.3
83.5
109.6
121.5

                                     3-3

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      The spent fuel from special research and test  reactors  Is  shipped  to
 either the SRP or the ICPP for indefinite storage or eventual reprocess-
 ing.   The production and naval reactor fuels are stored at SRP,  ICPP, and
 Hanford for routine reprocessing.   As of  December 31,  1983,  the  special
 spent fuel inventory was over  4500 Kg U-235, whereas the special Fort St.
 Vrain HTGR fuel in storage was 2.5 metric tons.

 3.3  High-level Radioactive Wastes (EPA82,  DOE84a,  Li79, St79, DOE80)

      The standard defines high-level radioactive wastes as the highly
 radioactive materials resulting from the  reprocessing  of spent nuclear
 fuel, including liquid waste produced directly in reprocessing and any
 solid material derived from such liquid waste.   This definition  is the
 same  as listed in the NWPA;  however,  it is  slightly different from the
 previous EPA definition  in the ocean dumping regulations, 40 CFR Parts
 220-227, and the NRC definition in 10 CFR Part 60.   Federal  regulations
 require that commercial  high-level waste  generated  in  the future be
 converted to a solid within 5  years.

      The fission products,  actinides,  and neutron-activated  products of
 particular importance are the  same for HLW  as those listed for the spent
 fuel  assemblies.

      Weapons program reactors  are  operated  (by DOE  contractors)  to pro-
 duce  plutonium.   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.

      High-level  radioactive  waste  that is generated by the reprocessing
 of  spent reactor fuel and targets  contains  more  than 99 percent  of the
 nonvolatile fission products produced  in  the fuel or targets during
 reactor operation.   It generally contains about  0.5 percent  of the
 uranium and plutonium in the original  fuel.   Most of the current HLW
 inventory,  which is the  result  of  DOE  national defense activities, is
 stored  at  the Savannah River Plant,  the ICPP at  the Idaho National Engi-
 neering Laboratory  (INEL), and  the Hanford  sites.   A small amount of
 commercial  HLW was  generated at  the Nuclear Fuel Services Plant  at West
 Valley, New York,  from 1966  to  1972.   These wastes  have been through one
 or more  treatment  steps  (i.e, neutralization,  precipitation, decantation,
 evaporation,  etc.).   Their volumes depend greatly on the steps they have
been  through.  They must  be  incorporated  into a  stable solid medium
 (e.g.,  glass)  for  final  disposal,  and  the volumes of these interim wastes
will be greatly  reduced  once this  has  been  accomplished.

     The DOE/defense  HLW at  INEL results  from reprocessing nuclear fuels
 from naval  propulsion  reactors and special  research and test reactors.
The bulk of  this waste, which is acidic,  has  been converted  to a stable,
granular solid (calcine).  At SRP  and  HANF,  the acidic waste from re-
processing  defense  reactor fuel  is  or  has been made  alkaline by  the
addition of  a caustic  and stored in tanks.  During  storage,  these alka-
line wastes  separate  into three  or  four phases:  liquid, sludge,  slurry,
                                     3-4

-------
and salt cake.  The relative proportions of liquid and salt cake  depend
on how much water is removed by waste evaporators during waste management
operations.  The condensed water may be recycled within the facility  or
decontaminated further and discharged.

     The commercial HLW at Vest Valley consists of both alkaline  and
acidic waste.  The alkaline waste was generated by reprocessing commer-
cial power reactor fuels and some Hanford N-Reactor fuels,  whereas the
acidic waste was generated by reprocessing a small amount of commercial
fuel containing thorium.

     The inventories of HLW in storage at the end of 1983 are listed  in
Table 3.3-1 (by volume) and Table 3.3-2 (by radioactivity).  Projected
volume and radioactivity data for DOE/defense, West Valley, and future
commercial HLW are given in Table 3.3-3.

3.3.1  HLW Inventories at SRP

     The approximately 111,000 m9 of alkaline HLW that has  accumulated at
the SRP over the past three decades is stored in high-integrity,  double-
walled, carbon-steel tanks.  The current inventories  (Tables 3.3-1 and
3.3-2) consist of alkaline liquid, sludge, and salt cake that were gener-
ated primarily by the PDREX reprocessing of nuclear fuels and targets
from plutonium production reactors.  As generated, most of the waste is
acid, and the sludge is formed after treatment with caustic and after
aging.  Salt cake results when the supernatant liquor is concentrated in
evaporators.

3.3.2  HLW Inventories at 1NEL

     The 9700 ms of HLW stored at INEL is  at  the  Idaho  Chemical Process-
ing Plant; it consists of 6900 m3 of liquid waste and 2800 m9 of calcine
(Tables 3.3-1 and 3.3-2).  Liquid HLW is generated at ICPP primarily by
the reprocessing of spent fuel from the national  defense (naval propul-
sion nuclear reactors) and reactor testing programs;  a  small amount is
generated by reprocessing fuel from nondefense research reactors.  This
acidic waste is stored  in large, doubly contained, underground,  stainless
steel tanks.  The waste is then  converted  to  a calcine, after which it is
stored in  stainless steel bins housed In reinforced concrete vaults.

3.3.3  HLW Inventories  at HANF

     The 203,000 m3 of  alkaline  HLW stored at HANF is In four phases:
liquid, sludge, slurry, and  salt cake.  This  waste, which  has been ac-
cumulating since  1944,  was  generated  by reprocessing  production  reactor
fuel for the  recovery of  plutonium,  uranium,  and neptunium for defense
and other  Federal programs.   Reprocessing  was suspended from 1972 until
November  1983.  Most  of the  high-heat-emitting isotopes (90Sr and 137Cs,
plus their daughters) have  been removed from the old  waste, converted to
solids as  strontium fluoride and cesium chloride, placed in double-walled
capsules,  and  stored  in water basins.   The liquid,  sludge, slurry, and
salt cake  wastes  (Tables  3.3-1 and  3.3-2)  are stored  in underground
concrete tanks with carbon steel liners.

                                      3-5

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            Table 3.3-1.  Current volume of HLW in storage by site through 1983 (DOE84a)
Volume (103 m3)
Site
Defense
Savannah River Plant
Idaho Chemical
Processing Plant
Hanford
Subtotal
Commercial
Nuclear Fuel Services
Acid waste
Alkaline waste
Subtotal
Grand total
Liquid

65.9
6.9
57.0
129.8


0.045
2.1
2.145
131.9
Sludge

12.8
(b)
47.0
59.8


(b)
0.17
0.17
60.0
Salt cake

32.7
(b)
95.0
127.7


(b)
(b)
(b)
127.7
Slurry

(b)
(b)
4.0
4.0


(b)
(b)
(b)
4.0
Calcine

(b)
2.8
(b)
2.8


(b)
(b)
(b)
2.8
fa)
Capsules

(b)
(b)
0.0049
0.0049


(b)
(b)
(b)
0.0049
Total

111.4
9.7
203.0
324.1


0.045
2.27
2.315
326.4
(b)
Capsules contain either strontium  (90Sr-90Y) fluoride or cesium  (137Cs-137mBa) chloride.
Not applicable.

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              Table 3.3-2.   Current  radioactivity  of HLW in storage by  site through  1983  (DOE84a)
Radioactivity*^
Site
Defense
Savannah River Plant
Idaho Chemical
Processing Plant
Hanford
Subtotal
Commercial
V Nuclear Fuel Services
Acid waste
Alkaline waste
Subtotal
Grand total
Liquid

85.9
16.2
33.0
135.1


2.95
15.2
18.15
153.2
Sludge

509.2
(b)
143.3
652.5


(b)
16.7
16.7
669.2
Salt cake

181.1
(b)
14.4
195.5


(b)
(b)
(b)
195.5
Slurry

(b)
(b)
0.22
0.22


(b)
(b)
(b)
0.22
(106 Ci)
Calcine

(b)
48.6
(b)
48.6


(b)
(b)
(b)
48.6

Capsules

(b)
(b)
283.3(C)
283.3


(b)
(b)
(b)
283.3

Total

776.2
64.8
474.2
1315.2


2.95
31.9
34.85
1350.0
(a)

(b)

(c)
Calculated values  allowing  for radioactive decay.

Not applicable.

Includes strontium and  cesium  in  capsules  and  separated  concentrates  that are awaiting encapsulation.
The quantity  of  90Sr-90Y  is 1.074 x  108 Ci and that  of  137Cs-137^a is  1.759 x 108 Ci.

-------
                             Table 3.3-3.  Historical and projected volume and associated
                             radioactivity of HLW in storage by site through 2000  (DOE84a)
00

End of
calendar
year

1980
1983
1985
1990
1995
2000

1980
1983
1985
1990
1995
2000

1980
1983
1985
1990
1995
2000

Liquid

59.8
65.9
55.4
51.4
37.6
29.0

9.34
6.9
7.0
5.2
5.7
2.9

39.0
57.0
62.0
57.0
57.0
57.0

Sludge

10.5
12.8
14.0
15.0
14.3
13.5

—
—
—
—
—
—

49.0
47.0
50.0
55.0
56.0
56.0

Salt
cake

26.4
32.7
41.2
50.3
44.3
36.6

—
—
—
—
—
—

95.0
95.0
95.0
95.0
95.0
95.0
Volume (103 m3)
Slurry Calcine Capsules ^ Glass ^b'
Savannah River Plant
— — — —
— — — —
— — — —
0.3
2.0
3.6
Idaho Chemical Processing Plant
2.07
2.8
3.0
4.9
— 6.8
11.0
Hanford
(c) — 0.0017
4.0 — 0.0049
6.0 — 0.0049
8.0 — 0.0103
9.0 — 0.0105
9.0 — 0.0105

Total

96.7
111.4
110.6
117.0
98.2
82.7

11.4
9.7
10.0
10.1
12.5
13.9

183.0
203.0
213.0
215.0
217.0
217.0
Radioactivity
(106 Ci)
	 Total 	

699.0
776.2
813.1
790.6
751.0
698.9

53.4
64.8
73.8
90.3
140.7
240.8

557.6
474.2
560.5
664.6
574.4
430.4
     (continued)

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    Table 3.3-3.   Historical and projected volume and associated radioactivity of  HLW  in  storage
                                  by site through 2000 (DOE84a)  (continued)

End of
calendar
year Liquid
1980 2.15
1983 2.145
1985 2.145
1990
1995
2000
Volume (10s «s)
Salt f ^ f *
Sludge cake Slurry Calcine Capsules Glass Total
Nuclear Fuel Services
0.047 ______ __ 2.2
0.170 — — — — — 2.315
0.170 — — -_ — — 2.315
— — — — — 0.159 0.159
— — — — — 0.159 0.159
— — — — _ — 0.159 0.159
Radioactivity
(10s Ci)
Total
37.7
103.5
98.0
86.6
76.5
67.5
(a)
(b)
(c)
Includes strontium and cesium in capsules and separated concentrates that are to be encapsulated.
Glass may be in storage at the site, in transit to a repository, or in a repository.
Slurry included with sludge.

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 3.3.4  HLW Inventories at NFS

      The 2315 m3 of HLW stored at NFS consists of 2270 m3 of alkaline
 waste and only 45 m3 of acid waste.   The alkaline waste was generated by
 reprocessing commercial and some Hanford N-Reactor spent fuels.   Initial-
 ly, all of the waste was highly acid; treatment with excess sodium hy-
 droxide led to the formation of an alkaline sludge.   The acid waste now
 in storage was generated by reprocessing a small batch of thorium-uranium
 fuel from the Indian Point-1 Reactor.  The alkaline  waste is stored in an
 underground carbon-steel tank, and the acid waste is stored in an under-
 ground stainless steel tank.  Reprocessing at the NFS plant was discon-
 tinued in 1972, and no additional HLW has been generated since then.   The
 current inventories of HLW at NFS are presented in Tables 3.3-1  and
 3.3-2.

 3.3.5  Waste Characterization

      A generic characterization of HLW at any site is difficult  because
 the wastes have been generated by several different  processes,  and sever-
 al methods have been used to condition the wastes for storage (e.g.,
 evaporation and precipitation).   In  some instances,  several different
 wastes have been blended.   Nonetheless,  representative chemical  and
 radionuclide compositions for HLW at SRP, ICPP,  HANF, and NFS can be
 found in some sources (DOE84a, Li79, DOE80).

      As with spent fuel,  HLW radioactivity levels depend on age.   To
 bring the activity into  perspective, calculations showed that fission
 products and heavy element radionuclides with half-lives exceeding 20
 years in the existing HLW are estimated  to be about  700 million  curies.

 3.3.6  Projections

      Projections  for  HLW (volume  and radioactivity)  by  source are pre-
 sented in Table 3.3-3.   The  projections  for SRP  are  based on the  restart-
 ing of the  L-Reactor  (fall of  1985)  and  initial  operation of the  Defense
 Waste Processing  Facility  (DWPF)  in  late 1989, with  the first radioactive
 glass to  be  made  in  1990.

      The  ICPP  projections  are  based  on predicted fuel deliveries  and
 estimates of  fuel  reprocessing and waste management  operations.   The  HANF
 projections  are based  on the  shutdown of  the  N-Reactor  in 1983 and  the
 restarting of  the  fuel reprocessing  plant in  November 1983,  with  opera-
 tion  projected  to  continue through 1993.   The HLW at  ICFP and HANF  are
not incorporated  in glass  because such processes  are  not  yet available
 there.  At NFS, vitrification  of  the waste is scheduled to begin  in
mid-1988 and to be completed by the  end  of  1989.

 3.4   Transuranic Wastes  (DOE84a, Li79, DOE80, Ja83, Br81,  DOE84b)

      The standard defines  transuranic wastes  as wastes  containing more
than  100 nanocuries of alpha-emitting transuranic  isotopes,  with half-
lives greater than 20 years, per grata of waste.  TRU waste was originally
                                     3-10

-------
defined by DOE as "...solid material that is contaminated to greater than
10 nCi/g with certain alpha-emitting radionuclides of long half-life and
highly specific radiotoxicity."  However, this definition was recently
revised by DOE to read that "TRU waste is material having no significant
economic value which, at the end of institutional control periods, is
contaminated with alpha-emitting radionuclides with atomic numbers greater
than 92 and half-lives greater than 20 years, in concentrations greater
than 100 nCi/g" (DOE84b).  Alpha-emitting transuranic nuclides represent
a special type of hazard because of their long half-lives and high radio-
toxicity.

     Most of the nuclides that make up TRU wastes have very long half-
lives and low specific activities.  Although a few daughter products have
energetic gamma emissions, their most significant hazard is due to alpha
radiation emissions.  Most TRU wastes can be handled with just the shield-
ing that is provided by the waste package itself.  These wastes are
classified as "contact-handled" TRU wastes.  A smaller volume may be
contaminated with sufficient beta, gamma, or neutron activity to require
remote handling.   Also, heat generation in stored TRU waste is not a
factor affecting how closely packages can be stored; however, avoiding
the production of a critical mass as a result of densely-stored material
must always be considered.

     Most TRU wastes are generated in DOE defense-related activities at
the Rocky Flats Plant  (RFP), Hanford Facilities, and the Los Alamos
National Laboratory  (LANL).  Nearly one-half of all TRU waste comes from
weapons components manufactured at RFP and  subsequent plutonium recovery
at these three sites.   Smaller amounts are  generated at the Oak Ridge
National Laboratory  (ORNL), SRP,  INEL, Argonne National Laboratory  (ANL),
Mound Facility, Bettis  Atomic Power Laboratory, Lawrence Livermore Labora-
tory, and Battelle-Columbus Laboratory.  The  second  largest source of  TRU
waste is decontamination and decommissioning  (DSD) projects, which account
for one-fourth of the  total.  About one-fifth of TRU wastes come  from
laboratory activities,  which can  produce exotic TRU  isotopes.

     The amounts of  TRU wastes  from fuel cycle activities  are quite  small
because of the current  moratorium on  reprocessing  and plutonium recycle.
The Nuclear Fuel Services1  reprocessing  of  nuclear fuel  at West Valley,
New York, produced  some TRU waste that was  disposed  of at  that site.   A
small amount  of TRU waste  is also being  generated  in industrial and
government-sponsored fuel  fabrication and  research.

3.4.1   Inventories  and Characterization

      As opposed  to  other radioactive  wastes,  TRU wastes  represent a group
of  liquid  and solid materials  with widely  varying chemical and physical
properties.   These  wastes are  categorized  as contact-handled (CH),  i.e.,
                                      3-11

-------
 having a surface dose rate of less than 200 mR/h; or remote-handled (RH),
 i.e., having a surface dose rate of greater than 200 mR/h.

      Before March 1970, low-level TRU wastes were disposed of by shallow
 land burial at AEC and commercial sites.  The estimated buried volume and
 mass of contained TRU elements at DOE sites are given in Table 3.4-1.

      Beginning in 1970, the AEC initiated a policy of retrievable storage
 for TRU wastes.  Storage facilities and emplaced waste containers were to
 have at least a 20-year lifetime, and during the storage period, a deci-
 sion was to be made regarding permanent disposal.  All of the retrievably
 stored waste is at the DOE sites shown in Table 3.4-2.   Also given in
 this table are the volume of the waste, the mass of TRU elements, and  the
 radioactivity as of December 31, 1983.   Estimates of the radioactivity of
 this waste are based upon emplacement records and a knowledge of the
 types of operations at the generation site.

      Over the years, some of the buried waste containers have been
 breached, and the surrounding soil has  been contaminated.   Accurately
 determining the volume of contaminated  soil is a difficult task, and the
 estimated amounts cover a rather broad  range (Table 3.4-3).   Also,  in  the
 early days at HANF,  ORNL,  and LANL,  some liquid wastes  containing TRU
 elements were spilled or drained to  the earth.   Further characterization
 is needed for better Identification  of  the volume of soil  that Is contam-
 inated with TRU elements.

      Through ongoing characterization studies,  the  DOE  sites  have esti-
 mated that their  buried  and  retrievable TRU solid waste Is composed
 primarily of the  physical  species given in Table 3.4-4.  Most of the
 storage sites have relatively  large  fractions  of combustible  material  and
 contaminated metal.

      Estimated  isotopic  compositions for projected  commercial wastes and
 for buried  and  retrievably stored wastes at  the  several  DOE sites where
 TRU wastes  are  emplaced are given in Table 3.4-5.  Background knowledge
 of  the  DOE  site operations and of the sources of commercial TRU wastes
 was used  to estimate  compositions when documented data were not  avail-
 able.   Separate composition data  for contact-handled and remotely handled
 waste were  available  for all sites that store both types of waste; however,
 composition data were not available for buried TRU waste at ORNL.  The
 radioactivity of ORNL buried waste was assumed to be the same  as  that of
 the  contact-handled waste.  These data represent the best  site estimates
 of  the isotopic compositions of existing TRU wastes at government sites.

 3.4.2  Projections

     The current inventory and projected accumulation at government
storage sites of buried TRU waste, as well as contact- and remotely-
handled waste from DOE/defense activities are given in Table 3.4-6.
                                     3-12

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Table 3.4-1.  Inventories and characteristics of DOE/defense
           TRU wastes buried through 1983 (DOE84a)
Values reported by burial site
as of Dec. 31. 1983
Burial
site
HANF
INEL
LANL
ORNL
SAND
SRP
Volume

29,230
73,267
6,580
272
«1
54.284
  Total
171,409
736
163.633
                             3-13

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      Table 3.4-2.   Inventories and characteristics  of  DOE/defense waste
               in TRU retrievable storage through  1983  (DOE84a)

Storage
site
HANF
INEL
LANL
NTS
ORNL
SRP
Subtotal
HANF
INEL
LANL
ORNL
Subtotal
Total
Values reported
of Dec.
Volume
(m3)
Contact
12,808
50,958
6,294.8
319.7
450
3,399
74,229.5
Remotely
21.8
50.69
26.6
653
752.09
74,981.6
by storage site as
31, 1983
Mass of TRU
elements
(kg)
handled (a)
340
524.4
247.5
7.775
12.33
98.5
1,230.5
handled (a)
5.4
0.319
1.27
0.613
7.602
1,238.1
Alpha
radioactivity
(Ci)
27,680
171,157
138,017
1,318.7
21,414
580,761
940,348
750
24.1
78
428
1,280.1
941,628
(a)
    Beginning with 1983,  TRU waste inventories are estimated on the basis of
    DOE Order 5820.2,  which defines TRU waste as 100 nCi/g.   Prior invento-
    ries were estimated on the basis of the earlier definition of TRU waste
    (10 nCi/g);  hence  a portion of the volume might be reclassified as
    non-TRU waste.
                                     3-14

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              Table 3.4-3.   Estimated  inventories  of  items  that might require special handling
                                   and/or treatment as  TRU  waste (DOE84a)
Volume (m3) of
contaminated soil
DOE
Burial
site
HANF
INEL
LANL
ORNL
SRP
Total
Solid waste
burial
27,900
56,640-156,000
1,000
12,000-60,000
Up to 38,000
135,540-282,900
Liquid
disposal/spills
32,000
0
140
1,000
Not reported
33,340
Mass (kg) of TRU elements in
contaminated soil
Solid waste
burial
350
Unknown
Unknown
Unknown
9.4
359.4
Liquid
disposal/spills
190
0
0.12
0.3
Not reported
190.42
Alpha radioactivity (Ci) of
contaminated soil
Solid waste
burial
30,000
Unknown
Unknown
Unknown
54,284
84,284
Liquid
disposal/spills
15,900
0
8.6
8
Not reported
15,916.6
(a)
    The mass of TRU elements and the radioactivity are included in the total inventory of burled waste (see
    Table 3.4-1).  There is no known method of estimating these values for the contaminated soil.

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         Table 3.4-4.  Physical composition of TRU wastes
                   at DOE/defense sites  (DOE84a)
Waste type

Absorbed liquids or sludges
Combustibles
Concreted or cemented sludge
Filters or filter media
Class
Metal
Other

Absorbed liquids or sludges
Alpha hot cell waste
Combustibles
Concreted or cemented sludges
Dirt, gravel, or asphalt
Filters or filter media
Glass
Laboratory waste
Metal
Solidified fuel
Other
Unknown

Absorbed liquids or sludges
Combustibles
Concreted or cemented sludges
Dirt, gravel, or asphalt
Filters or filter media
Glass
Metal
Other

Combustibles
Metal

Combustibles
Filters or filter media
Metal
Other

Combustibles
Noncombustible
Waste
Retrievably
Contact handled
BASF
5.8
23.3
2.4
0.5
3.6
48.7
15.7
ISEL
14.13

20.67
3.034
1.802
7.522
2.051

28.02

12.32
10.49
LANL
18.1
16.6
11.9
2.1
2.6
0.5
39.3
8.9
UTS
45
55
ORHL
56
0
16
28
SRP
70
30
composition (vol. X)
stored waste
Remotely handled


65.1
6.1


28.8



80.09
4.333


11.59

3.513

0.4684










100




39
2
40
19




Buried vaste

8
20
5
1
8
40
18

(a)












24

44




32




(a)




90
10
(a)
   Data not available to determine composition of burled waste.
                                3-16

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 Table 3.4-5.  Estimated isotopic composition of buried,
     retrievably  stored, and future TRU waste (DOE84a)
Isotope

238Pu
239Fu
2*°Fu
ZA1Pu
242-
Fu

233.J
238Pu
239,
Fu
2«°Pu
2*2Pu
241 Am

233U
236Pu
238Pu
239Pu
2*°Pu
24lPu
2A2Pu
i?
24AC«
237R
238PU
239Pu
240.
Pu
2AlPu
242Pu
241.
Am
244-
Isotopic composition (wt. X)
Retrievably stored waste
Contact handled Remotely handled
HANF
0.01 0.05
93.89 86.36
5.75 11.75
0.34 1.63
0.02 0.21

mtt
18.22
0.728 0.05417
72.36 92.86

4.552 7.001
0.0238 0.0860
A. 119
JAW,
1.7
2.8E-06
2.25 0.014
87.92 93.55
3.59 3.89
0.518 0.536
0.324 0.023
1.4

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Table 3.4-5.   Estimated  isotopic  composition of buried,  retrievably stored,
                    and future TRU  vaste (DOE84a) (continued)

Isotopic
composition 
-------
  Table 3.4-6.  Current Inventories and projections of DOE buried and stored
                  TRU waste from defense activities (DOE84a)
End of
calendar
year
Volume
(109 ras)
Accumulation
Radioactivity
(106 Ci)
Accumulation
Mass
(kg)
Accumulation
                                  Buried
                                         (a)
          1983
          1985
          1990
          1995
          2000
171.4
171.4
171.4
171.4
171.4
0.3
0.3
0.3
0.2
0.2
736.0
735.9
735.9
735.9
735.9
                                Stored(a>'(b>
1983
1985
1990
1995
2000
75.0
85.5
112.0
140.3
168.7
1.5
1.8
2.5
3.3
3.9
1238.1
1417.9
1816.4
2369.4
2878.8
( ^  Beginning with 1983,  TRU waste inventories have been estimated on the
    basis of DOE Order 5820.2, which defines TRU waste as 100 nCi/g.   Prior
    inventories were estimated using the earlier definition of TRU waste (10
    nCi/g);  hence a portion of the volume might be reclassified as non-TRU
    waste.
* '  Includes TRU wastes that will be shipped to the Waste Isolation Pilot
    Plant.
                                     3-19

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                               REFERENCES
Br81      Bryan G. H., Battelle Pacific Northwest Laboratory Character-
          ization of Transuranium-Contaminated Solid Wastes Residues,
          PNL-3776, April 1981.

Bu82      Burton B. W., et al.  Los Alamos National Laboratory, Overview
          Assessment of Nuclear Waste Management, LA-9395-MS, August
          1982.

DOE80     Department of Energy, Spent Fuel and Waste Inventories and
          Projections, ORO-778, August 1980.

DOE84a    Department of Energy, Spent Fuel and Radioactive Waste
          Inventories, Projections, and Characteristics, DOE/RW-0006,
          September 1984.

DOE84b    Department of Energy, Radioactive Waste Management, DOE Order
          5820.2, dated February 6, 1984.

EPA82     U.S. Environmental Protection Agency, Draft Environmental
          Impact Statement for 40 CFR 191:  Environmental Standards for
          Management and Disposal of Spant Nuclear Fuel, High-Level and
          Transuranlc Radioactive Wastes, EPA 520/1-82-025, December
          1982.

Ja83      Jensen R. T., and Wilkinson F.  J., II, Rockwell International
          Energy Systems Group, Characteristics of Transuranic Waste at
          Department of Energy Sites, RFP-3357, May 1983.

L179      A. D. Little, Inc., U.S. Environmental Protection Agency*
          Technical Support of Standards  for High-Level Radioactive Waste
          Management:   Volume A, Source Term Cheracterization, EPA
          520/4-79-007A, 1979.

St79      Storch S. N., and Prince B. E., Union Csrbide Corp-Nuclear
          Division, Assumptions and Ground Rules Used in Nuclear Waste
          Projections  and Source Term Data, ONWI-24, September 1979.
                                     3-20

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                    Chapter 4:  PLANNED DISPOSAL PROGRAMS
4.1  Introduction

     Since the inception of the nuclear age in the 1940's, the Federal
government has assumed ultimate responsibility for the care and disposal
of high-level radioactive wastes, regardless of their source, in order to
protect the public health and safety and the environment.

     The Civilian Radioactive Waste Management (CRWM) Program, formerly
called the National Waste Terminal Storage (NWTS) Program, was estab-
lished in 1976 by DOE's predecessor, the Energy Research and Development
Administration (ERDA), to develop technology and provide facilities for
the safe, environmentally acceptable, permanent disposal of high-level
nuclear waste.  Included in the HLW are wastes from both commercial and
defense sources, such as spent fuel from nuclear power reactors, accumu-
lations of wastes from production of nuclear weapons, and solidified
wastes from fuel reprocessing.

     The Federal laws defining DOE's responsibility for the long-term
management of HLW specify that the DOE must provide facilities for the
successful isolation of HLW from the environment in federally licensed
and federally owned repositories for as long as the wastes present a sig-
nificant hazard (AEA54, ERA74, DEOA74, NWPA82).  The Nuclear Waste Policy
Act of 1982 (NWPA), enacted January 7, 1983, as Public Law 97-425, con-
firmed the responsibility of the DOE for management of high-level radio-
active waste.  The NWPA also confirmed EPA1a role in setting general
standards and the Nuclear Regulatory Commission's (NRC's) role to act as
the licensing agent.  The NWPA directed the DOE to provide safe facili-
ties for isolation of high-level radioactive wastes from the environment.

     As directed by the NWPA, development work has been performed to
define methods for disposal of spent fuel and solidified high-level and
transuranic radioactive wastes at the direction of Congress.  The devel-
opment work is being concentrated on mined geological repositories.  Such
repositories would be constructed in suitable host media at depths great-
er than 300 meters by conventional mining techniques.  Suggested host
media Include granite, basalt, volcanic tuff, and salt.  Wastes In canis-
ters would be placed in holes in the mine floor.  When the repository is
full, the holes and shafts would be backfilled.  After a validation
period, during which the wastes could be retrieved, the site would be
                                     4-1

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 permanently sealed .   Protection
 ...  «... f.n.  . dur.bl.  c.,,1...                ,
 tlon poc.ntl.l  for  r.dtonuclld..  through tk. .nrlr^..^'     i°* *llr*'
 rock.   Klo.d ..olo.lc.l  r.po.ltorl«. .*. .™!rTj T^  "°UI"' tta ko"
 b.for.  .a, ..hn .qu.U, .litrtuX^LTESJ    'L'1*1  f" ""
      Th«  Vtest*  UoUtlon  Pilot  Plant
           K« M..I... .111  provld.  .                 .n
 d««on.tr.te  th«  i«ft dl«po.«l of  tr««ur«Uc r!  *^  facility to
 fro.  U.S.  def.n.e  .tctlvitl.t .nd  pro,rl!i.r  r*4i°*ctlT« •«•«•• r««ultlng
               Udloectiv.  a«.   nw^nt f-  (DOE82.
         CIWM  Proiru Mpb«*lx«« d««p Baargroua  dl«po«*l la
 r.po.lton.e loc.t.d in g«olo§ic«ll, atabl. bodla. of rock
 currently being  considered  Include bedded salt deposit.  e«lt
 bee.lt ,  tuff,  and crystalline  rocka.  Tteae rock typ«e are
 at different localltlea within the co-termlnmi.

                  proj'e"
      (I)  The Salt Kapoeltory Project  (for bedded aalt depo.it. and ..It
          dotjwa) .

      (2)  The Baaalt Uaate  laolatlon Project  (for baaalt).

      (3)  The Nevada Nuclear Uaate Storage loveatlgetlona (for tuff).

      (4)  The Cryatalllne tepoaltory Project  (for cryatallin* rock.).

     The proceaa for alt Ing the geologic repoaltoriea la defined In the
NVPA, Including a eequence of the etep. that  form the) baale for the
etrategy to achieve operation of a eafe. environmentally aouad.
geologic repoaitory by 1998.

*-Z.l  Flrat Cotjmerclal lepoeitorr (DOM 2, DOC»4«-d)

     The NUPA require, that th« DOE nominate at leaat ft»« altea to r
Prealdant and recommend three candidate elte. for duractarliar.!*. ..
poeeibla location, for the flrat Federal repcaltory.  The) rock
being considered e. potential hoete for the flrat rtpoaltorr
ba.alt (a fin.-gr.lnad rock formed by th. aoUdlflcaUon oP

                                     proct" to to
t. ntc
repository (see Figure 4.2-2).  Thay Include j    "*'** for tb«
          A Nevada site In tuff.
                                     4-2

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                                                       NORTHERN
                                                      APPALACHIANS
                COLUMBIA
              RIVER PLATEAU
                  FLOOD
                 BASALTS
                                                             CRN
                                                       APPALACHIANS
Flgur« 4.2-1.  teflon* Umtlfiad by DOC •• voter consideration for
     t«ologlc*l disposal of hl|h-l«vel nuel«*r wast* (DOES.
                                 4-3

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                                                     CYPRESS CREEK
                                                         DOME
Figure A.2-2.  Sites Identified by DOE as potentially acceptable
               for the first repository (DOE84d).

-------
     0    A Washington site in basalt.

     0    Two Texas sites in bedded salt.

     0    Two Utah sites in bedded salt.

     0    One Louisiana site in a salt dome.

     *    Two Mississippi sites In salt domes.

     Draft Environmental Assessments have been prepaied for all nine
potentially acceptable sites (DOE84e-m).  In December 1984, five sites
were tentatively nominated as being suitable for site characterization:
Yucca Mountain in Nevada; Richton Dome in Mississippi; Deaf Smith County
in Texas; Davis Canyon in Utah; and Hanford in Washington.  Three of
these (Yucca Mountain, Deaf Smith County, and Hanford) were recommended
by DOE as tentative choices for site characterization.

4.2.2  Second Commercial Repository (NWPA82, DOE82, DOE84b)

     In accordance with the NWPA, a separate process of nominations and
recommendations will be conducted for a second repository site, which is
to be identified by 1990.  The NWPA permits sites characterized for the
first repository to be nominated for the second repository if not select-
ed as the first site.  In addition to crystalline rocks, potential host
rocks for the second repository are salt, tuff, and basalt.

     As part of its efforts to determine potentially acceptable sites for
a second repository, DOE is conducting literature studies on crystalline
rock in the following 17 states:  Connecticut, Georgia, Maine, Maryland,
Massachusetts, Michigan, Minnesota, New Hampshire, New Jersey, New York,
North Carolina» Pennsylvania, Rhode Island, South Carolina, Vermont,
Virginia, and Wisconsin.

4.3  Geological Media

     The characteristics of the various geological media being considered
are important in understanding the issues of high-level radioactive waste
disposal.

4.3.1  Salt Media (DOE82, DOE83a, DOESAe-g,1-1, Bu82)

     Both bedded and domed rock salt are being investigated by DOE's CRWM
Program as a suitable host rock for the long-term isolation of high-level
radioactive waste.  Salt is suitable as a host rock because of its struc-
tural strength, radiation-shielding capability, high plasticity  (which
enables fractures to self-seal at repository depths), low moisture con-
tent, and low permeability.  In addition, salt deposits are abundant in
the United States, and the cost of mining is low.  A desirable feature of
many bedded salt basins is their relatively simple structure, from which
                                     4-5

-------
 the  stratigraphy  of  a wide  area near  the  repository can be projected.
 Although salt  deposits  are  widespread,  the  salt  Itself and associated
 deposits of  potash or hydrocarbons  are  resources that could increase the
 probability  of accidental human intrusion into a repository.  The solu-
 bility of rock salt  is  two  orders of  magnitude greater than any other
 potential host rock  and this  is Important in  the analysis of potential
 failure modes  for salt.

 4.3.2   Tuffs (DOE82, DOE84h,  Bu82,  K180)

     The two forms of tuff  considered for repository use are quite dif-
 ferent.  The first form is  densely  welded tuff (i.e., one in which the
 glass  shards became  fused because they  were hot  and plastic when deposit-
 ed).   This form has  high density, low porosity and water content, and the
 capability to  withstand the temperatures  generated by radioactive waste.
 The  compressive strength, thermal conductivity,  and thermal expansion of
 densely welded tuff  are comparable  to those of basalt.

     The second form of tuff  of interest  is a zeolltic tuff (i.e., a non-
 welded tuff  containing  zeolite, a hydrous silicate of open molecular
 structure).  This form  has  low density, high  porosity, low interstitial
 permeability,  high water content, extremely high aorptive properties,
 moderate compressive strength, and  moderate thermal conductivity.  Dehy-
 dration of some zeolites begins at  about  100°C,  and unless the water re-
 leased IB able to escape through the  rock,  it could contribute to changes
 in stress that could result in fracture.  An  increase in temperature can
 also cause some zeolites to decompose to  new  minerals of lower sorptive
 capacity.

     The repository  design  concept  is to  place radioactive waste in ther-
 mally  stable welded  tuff, where it  would  gain a  significant benefit from
 highly sorptive barriers of zeolitic  tuff underlying, and where possible,
 overlying the  welded tuff.

     Occurrences  of  welded  and zeolitic tuffs are widespread, and some
 occur  in thick sections  in  the western  states; however, their homogeneity
 and hydrologic properties have not  been characterized.  Most of these
 tuffs  are  relatively young  geologically;  they have been broken into
 blocks  tens  of kilometers in  size by  tectonic forces that were active
 during  and after  the time the tuffs were  formed  through volcanic erup-
 tions.

 4.3.3   Basalt  (DOE82, DOE84m, Bu82, K180)

     Basalt  is  the potential host rock at the Hanford Site in Washington,
where it occurs in a thick  section near the middle of the extensive
basalt  flows of the Columbia Plateau.  Thick basaltic sections also occur
 in Idaho and Oregon.   These basaltic  terrains are geologically young, and
earthquakes have caused possible surface manifestations;  however, no
faults  are known to jeopardize the Hanford area.   Deep drilling at
Hanford has shown that two thick basalt layers (one 55 m thick and the
other 36 m) occur at about 950 m below the surface that may be suitable
                                     4-6

-------
for repository construction.  Most openings within these layers are
filled with alteration products (predominantly clay minerals)  and thus
provide rock masses of low permeability.  These basaltic masses are among
the strongest of common rock types.  Basalt has moderate thermal conduc-
tivity and a high melting temperature; therefore, it can withstand a high
thermal load.

     Basalts of the Columbia Plateau commonly have zones of columnar
joints or rubble that are potential channels for water flow.  Water-
bearing sedimentary interbeds within the basalt section are also common.
The geologic section at Hanford thus comprises a system of alternating
aquifers and relatively impermeable zones.  The mineralogy and resulting
sorptive properties of the partially altered permeable basalt in the
sediments must be determined, as they will differ from those of the fresh
basalt.

4.3.4  Granite and Related Crystalline Rocks (DOE82, Bu82, K180)

     Granite and related crystalline Igneous and metamorphic rocks, such
as gneiss, have been proposed as potential host rocks for a repository.
These are the most abundant rocks in the upper 10 km of the earth's
continental crust.  Crystalline rocks underlie virtually all of the
United States; they occur at the surface in stable areas, in the cores of
many mountain ranges, and beneath all of the younger sedimentary cover.
Their strength, structural and chemical stability, and low porosity make
them attractive for waste repositories.  The water content of these rocks
is low and is held mainly in fractures and in hydrous silicate minerals.

     Because crystalline rocks are ubiquitous, they occur in various
tectonic settings.  In some areas of the United States, crystalline rocks
have been demonstrated to be stable for as long as 2.5 billion years.  In
other areas, crystalline rocks are involved in younger episodes of moun-
tain building that occurred only tens to hundreds of millions of years
ago.

     At depths in excess of several hundred meters, where vertical and
horizontal stresses increase, the permeability is reduced considerably by
closure of the fractures.  At some depth, granitic rocks probably become
nearly impermeable.  A principal goal in evaluating these rocks for
nuclear waste disposal will be to use geologic, geophysical, geochemical,
and hydrologic investigations to determine the depths at which a reposi-
tory should be placed so that fracture permeability will not represent an
escape pathway for the radionuclides.  The safe depth for a repository
probably will vary from region to region as a result of the influence of
tectonic history on fracture permeability.

4.4  Waste Isolation Pilot Plant (Bu82, DOEBOa, DOE83a,b, Le84)

     In 1974, DOE began a program to develop a Waste Isolation Pilot
Plant in the Los Medanos area of southeastern New Mexico to demonstrate
the safe disposal of TRU radioactive wastes from national defense pro-
                                     4-7

-------
 grains.  The HIPP, as authorized by Public Law 96-164, is specifically
 exempted  from licensing by the Nuclear Regulatory Commission.

     The  WIPP is located in a 610-meter-thick bedded salt formation.  The
 formation, which is first encountered at a depth of 260 meters below the
 surface,  is over 200 million years old.  The facility has a capacity of
 0.18 million cubic meters of contact-handled TRU and 7 thousand cubic
 meters of remotely handled TRD.  The facility, scheduled to begin oper-
 ation in  October 1988, will also contain a research and development area
 and a retrievable high-level waste experiment area.  The limited quantity
 of high-level waste emplaced for experimental purposes will be removed
 from the  WIPP before the facility is permanently sealed.

 4.5  Disposal of DOE Defense High-Level Wastes (NWPA82, DOE84b, DOEB5)

     The  NWPA of 1982 required an evaluation be made to determine the use
 of disposal capacity at civilian repositories for the disposal of high-
 level wastes generated by defense activities.  The NWPA further states
 that after factors relating to cost efficiency, health and safety, regu-
 lation, transportation, public acceptability, and national security are
 taken into account, unless the evaluation shows that the development of a
 separate  repository Is necessary, the Secretary of Energy shall proceed
 with arrangements for using the "civilian" repositories for both commer-
 cial and  defense high-level wastes.

     A draft evaluation was prepared by DOE, and because of the cost ad-
 vantage of disposing of defense wastes in a combined commercial and de-
 fense repository, DOE has recommended this option.  The NWPA clearly
 states that coats resulting from permanent disposal of defense high-level
vaete shall be paid by the Federal government.

 4.6  Alternative Disposal Methods (NWPA82, DOE80a,b)

     The NWPA also requests the DOE to continue a program of research,
developmenti and investigation of alternative means and technologies for
 the permanent disposal of high-level radioactive waste from civilian
nuclear activities and Federal research and development activities.

     Over the years* the DOE and Its predecessor agencies have and con-
tinue to study several disposal methods.  These Include:

     0     Very deep hole concept:  placement of containers of waste into
          holes 3000 to 10,000 meters deep.

     *     Rock melt concept:   placement of fresh liquids or slurries
          directly into rocks by melting;  the heat of the wastes would
          melt the rock,  and  thus become incorporated as an Integral
          component of the rock.

     0     Island-based concept:   emplacement of wastes within deep geo-
          logical formations  on a remote isolated island.
                                     4-8

-------
     0    Extraterrestrial concept:  disposal of selected fractions  of
          reprocessed waste into an earth escape trajectory or solar
          orbit.

     0    Transmutation concept:  reduction of selected fractions of
          reprocessed waste by transmuting it to stable Isotopes.

     0    Well injection concept:  disposal of fresh liquid wastes or
          slurries by deep veil injection at depths of 1,000 to 4,800
          meters or by shale/grout high-pressure injection at depths of
          300 to 480 meters.

     0    Ice sheet concept:  disposal of waste containers in remote
          continental ice sheets.

     0    Subseabed concept:  emplacement of waste containers on or  under
          the ocean floor.  (Present U.S. law and international treaties
          prohibit disposal of high-level wastes In the ocean.)

4.6.1  In-Place Tank Stabilization (DOE83a.c, Le84, An85)

     The DOE is considering the possibility of in-place stabilization of
various defense high-level wastes currently stored in single-walled
underground tanks if, after the requisite environmental documentation, it
is determined that the risks and costs of retrieval and transportation
outweigh the environmental benefits of disposal in a mined geologic
repository.

     The DOE performance assessment for disposal at Hanford of the sin-
gle-shell tank wastes by in-place stabilization would include stabilizing
the waste by drying it to a near solid form or grouting or mixing it with
stabilizing chemicals or .in situ vitrification, and then covering it with
substantial engineered barriers.  Current recommended plans call for
placing a monument at the surface of the burial sites, and placing perma-
nent records in public libraries, time capsules, computerized information
centers, etc., to reduce the probability of all records of the repository
being lost.
                                     4-9

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                                REFERENCES
 AEA54     Atomic  Energy  Act  of  1954, As Amended,  42 USC  2011 et  seq.

 An85      Anderson  B.  N.,  et al,  Single-Shell Tank Technology
           Demonstration; Waste  Management  '85, Volume  1, University of
           Arizona,  Tucson, Arizona, March  24-28,  1985.

 Bu82      Burton  B. W.,  et al,  Los Alamos  National Laboratory, Overview
           Assessment of  Nuclear Waste Management, LA-9395-MS, August
           1982.

 DEOA77     Department of  Energy  Organization Act of 1977, 42 USC  7101 et
           seq.

 DOESOa     Department of  Energy, Final Environmental Impact Statement,
           Waste Isolation Pilot Plant, 2 Volumes, DOE/EIS-0026,  October
           1980.

 DOESOb     Department of  Energy, Final Environmental Impact Statement,
           Management of  Commercially Generated Radioactive Waste, 3
           Volumes, DOE/EIS-0046F, October  1980.

 DOE82      Department of  Energy, Program Summary, Nuclear Waste Management
           and Fuel Cycle Programs, DOE/NE-0039, July 1982.

 DOE83a     Department of  Energy, The Defense Waste Management Plan,
           DOE/DP-0015, June  1983.

 DOE83b     Department of  Energy, Secretary's Annual Report to Congress,
           DOE/S-0010(83), September 1983.

 DOE83c     Department of Energy, Hanford Defense Waste Disposal Program,
           2 Volumes, for U.S. EPA Staff Site Visit, October 1983.

 DOE84a    Department of Energy, Mission Plan for the Civilian Radioactive
          Waste Management Program, DOE/RW-0005 (Draft), April 1984.

DOE84b    Department of Energy, Implementation of the Nuclear Waste
          Policy Act of 1982, Fact Sheet, DOE/RW-0008, October 1984.

DOE84c    Department of Energy, Department of Energy Announces Three
          Proposed Sites for Potential Disposal of High-Level Radioactive
          Waste, DOE News Release, R-84-150, December 19, 1984.
                                     4-10

-------
DOE84d    Department of Energy, Preliminary Selection of Candidate Nu-
          clear Waste Repository Sites for Field Characterization, Fact
          Sheet, December 1984.

DOE84e    Department of Energy, Draft Environmental Assessment, Lavender
          Canyon Site, Utah, DOE/RW-0009, December 1984.

DOE84f    Department of Energy, Draft Environmental Assessment, Davis
          Canyon Site, Utah, DOE/RW-0010, December 1984.

DOE84g    Department of Energy, Draft Environmental Assessment, Cypress
          Creek Dome Site, Mississippi, DOE/RW-0011, December 1984.

DOE84h    Department of Energy, Draft Environmental Assessment, Yucca
          Mountain Site, Nevada Research and Development Area, Nevada,
          DOE/RW-0012, December 1984.

DOE84i    Department of Energy, Draft Environmental Assessment, Richton
          Dome Site, Mississippi, DOE/RW-0013, December 1984.

DOE84J    Department of Energy, Draft Environmental Assessment, Deaf
          Smith County Site, Texas, DOE/RW-0014, December 1984.

DOE84k    Department of Energy, Draft Environmental Assessment, Swisher
          County Site, Texas, DOE/RW-0015, December 1984.

DOE841    Department of Energy, Draft Environmental Assessment, Vacherie
          Dome Site, Louisiana, DOE/RW-0016, December 1984.

DOE84m    Department of Energy, Draft Environmental Assessment, Reference
          Repository Location, Hanford Site, Washington, DOE/RW-0017,
          December 1984.

DOE85     Department of Energy, An Evaluation of Commercial Repository
          Capacity for the Disposal of Defense High-Level Waste,
          DOE/DP/0020/1, June 1985.

ERA74     Energy Reorganization Act of 1974, 42 USC 5811 et seq.

K180      Klingsberg C., and Duguid J., Status of Technology for Isolat-
          ing High-Level Radioactive Wastes in Geologic Repositories,
          DOE/TIC 11207 (Draft), October 1980.

Le84      Briefing by D. B. LeClaire, Director, Office of Defense Waste
          and Byproducts Management, U.S. DOE to M. Meigs, Staff member
          of Subcommittee on Energy Research and Development, on Defense
          Programs, Defense Waste and Byproducts Management, March 21,
          1984.

NWPA83    Nuclear Waste Policy Act of 1982, Public Law 97-425, January 7,
          1983.
                                     4-11

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                      Chapter 5:   RADIATION DOSIMETRY
5.1  Introduction

     Radionuclides transported through the environment may eventually
reach people.  Contact may occur through either external exposure to
radloactlvely contaminated air, water, and ground surfaces or internal
exposure from inhaling or Ingesting radioactively contaminated air,
water, or food.  Individuals in the population may absorb energy emitted
by the decaying radionuclides.  The quantification of this absorbed
energy is termed doslmetry.  This chapter describes the dosimetric
models for internal and external exposures, the EPA procedure for
implementing the dosimetric equations associated with the models, and
the uncertainties in dosimetric calculations.

     Mathematical models are used to calculate doses to specific human
body organs.  The models account for the amount of radionuclides enter-
ing the body, the movement of radionuclides through the body, and the
energy deposited in organs or tissues resulting from irradiation by the
radionuclides that reach the tissue.  These models provide the basis for
the computer codes, RADRISK and DARTAB, which EPA uses to calculate
doses and dose rates.  (See Appendix A.)

     Uncertainties in doslmetric calculations arise from assumptions of
uniform distribution of activity in external sources and source organs
and assumptions concerning the movement of the radionuclides In the
body.  The uncertainties associated with dosimetric calculations are
difficult to quantify because the data available for determining distri-
bution for the parameters used in the models are usually insufficient.
The major source of uncertainty in dosimetry is the real variation in
parameter values among Individuals in the general population while doses
and dose rates are calculated for a "typical" member of the general
population.  The three sources of dosimetric uncertainty assessed by EPA
are:  individual variation, age, and measurement errors.  The effects of
uncertainty are discussed in greater detail in Sections 5.5 and 5.6.

5.2  Definitions

5.2.1  Activity

     Radioactive decay is a process whereby the nucleus of an atom emits
excess energy.  The property possessed by atoms that emit this energy is
referred to as radioactivity.  The "activity" of a radioactive material
                                    5-1

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is characterized by the number of atoms that emit energy, or disinte-
grate, in a given period of time.  The unit of activity used in this
report is the curie (Ci), which equals 3.7 x 10   disintegrations
per second.  The excess energy is normally emitted as charged particles
moving at high velocities and photons.  Although there are many types of
emitted radiations, only three are commonly encountered in radioactive
material found in the general environment:  alpha radiation (nuclei of
helium atoms), beta radiation (electrons), and gamma radiation (photons).

     The primary mechanism for radiation damage is the transfer of
kinetic energy from the moving alpha and beta particles and photons to
living tissue.  This transfer leads to the rupture of cellular constit-
uents resulting in electrically charged fragments (ionization)   Al-
though the amount of energy transferred is small in absolute terms  it
is enough to disrupt the molecular structure of living tissue  and'
depending on the amount and location of the energy release, leads to the
risk of radiation damage.

5.2.2  Exposure and Dose

     The term "exposure" herein denotes the subjection of an organ or
person to a radiation field.  The term "dose" refers to the amount of
energy absorbed per gram of absorbing tissue as a result of the exno
sure.  An exposure, for example, may be acute, i.e., occur over a short
period of time, while the dose, for some internally deposited materials
may extend over a long period of time.                        materials,

     The dose is a measure of the amount of energy deposited by the
alpha and beta particles or photons and their secondary radiation«%«
the organ.  The only units of dose used in this chapter are ?he
rad—defined as 100 erg (energy units) per gram (mass unit)--and i-ho
millirad Onrad), which Is one one-thousandth of a rad.  The raf J^L
seats the amount, on average, of potentially disruptive energy trans-"
ferred by ionizing radiation to each gram of tissue.  Becausf it IT
necessary to know the yearly variation in dose for the calculation
described in this report, the quantity used will L M! calculatioi"J
dose (or dose rate) in rad or miUirad (per ££)     * aVerage annual
5.2.3  External and Internal Exposures

     Radiation doses may be caused by either external or internal
sures.  External exposures are those caused by radioactive
located outside the body, such as irradiation of the DoJy
material lying on the ground or suspended in the air   In
sures are caused by radioactive material that Ss entered
through the inhalation or consumption of radioactive mHera    H
once entered the body, the contaminant may be transmit to ith£
internal organs and tissues.                     u»»i«ea to other
     The external exposures considered in this report ar* *>,,«
ing from irradiation of the body by gamma rays only* Gamma rays
energy photons) are the most penetrating of those radiations considered
                                    5-2

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 and external gammas may contribute to the radiation dose affecting all
 organs in the body.  Beta particles (electrons), which are far less
 penetrating, normally deliver their dose to, or slightly below, the
 unshielded surface of the skin and are not considered because their
 impact is small, particularly on clothed individuals.  Alpha particles
 (helium nuclei), which are of major importance internally, will not
 penetrate unbroken skin and so are also excluded from the external dose
 calculations.  The internal exposures considered in this report origi-
 nate from all three types of radiation.

 5.2.4  Dose Equivalent

      Different types of charged particles differ in the rate at which
 their energy is transferred per unit of length traveled in tissue,  a
 parameter called the linear energy transfer (LET) of the particle.  Beta
 particles generally have a much lower LET than alpha particles.  Alpha
 particles are more damaging biologically, per rad,  than gamma rays  and
 beta particles.  In radiation protection, this difference is accounted
 for by multiplying the absorbed dose by a factor, Q,  the quality factor,
 to obtain a dose equivalent.   The quality factor Is Intended to correct
 for the difference in LET of the various particles.   At present, the
 International Commission on Radiological Protection (ICRP)  recommends
 the values Q"l for gamma rays and beta particles and  Q"20 for alpha
 particles (ICRP77).   The units for the dose equivalent,  corresponding  to
 the rad and millirad,  are rem and mlllirem.   Thus,  dose equivalents for
 gamma rays and beta  particles are numerically equal to  the dose  since
 the dose equivalent  (mrem)  -  (Q"l) x dose (mrad) while  alpha dose
 equivalents are twenty times  as large,  dose  equivalent  (mrem)  -  (Q-20) x
 dose (mrad).

 5.3  Dosimetric Models

      The radiation dose  has been defined,  in Section  5.2.2,  as the
 amount  of energy absorbed per unit mass of tissue.  Calculation of  the
 dose requires  the  use  of mathematical models such as  that shown later in
 Equation (5-2).  In  this equation, the  amount  of activity ingested, I,
 is  multiplied  by the fraction,  fj., going to  the  blood, and  the fract-
 ion,  f£,  going to  a  specific  tissue.  E is the amount of energy
 absorbed by the tissue for  each unit of activity so that the product of
 all these factors  divided by  the mass of the tissue is, by definition,
 the radiation  dose per unit activity.   The remaining term,
 [l-e~^tl/X,  indicates  how the  activity  deposited in the tissue
 changes with time.  All  these  factors together yield the dose rate.  A
 more comprehensive description of  the equations used is given in
Appendix A.

 5.3.1  Internal Doses

     Any  effort  at calculating dose and  risk must, of necessity, involve
 the use of models.  In Its  simplest form, a model is a mathematical
                                   5-3

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 representation of a physical or biological system.   If,  for example,  the
 amount of radioactive material in an organ is measured periodically,
 a graph of the activity in the organ, such as that  in Figure 5.3-1, is
 obtained.  In the simplest case, analysis of these  data  may indicate  that
 the fraction of the initial activity, R,  retained In the organ at  any
 time, t, is given by an equation of the form


                               R - e-K                         (5-1)
 where X is the elimination rate  constant.   (More  generally,  it may
 require the sum of two or more exponential  functions  to properly
 approximate the decrease of radioactivity in  the  organ.  This may be
 interpreted physically as indicating  the existence of two or more
 "compartments" in the  organ from which  the  radionuclide leaves at dif-
 ferent rates•)
      The  elimination rate  constant,  X,  is  the  sum of two terms, which
 nay be measured experimentally, one  proportional to the biological
 clearance half-time  and  the other proportional to the radioactive
 half-life.   The effective  half-life, ti/2» for these processes is the
 time required  for  one-half of the material originally present to be
 removed by biological clearance or radioactive decay.

      If radionuclides are  generally  found to follow this behavior, then
 this equation  may  be used  as a general  model for the activity in an
 organ following deposition of any initial activity.  In general, the
 models used by EPA are those recommended by the ICRP and are documented
 in  detail in the ICRP79.   A brief description  of each model is given
 below as  an aid to understanding the material  presented in the remainder
 of  this chapter.

      As mentioned  earlier, all radiations—gamma, beta, and alpha—are
 considered  in  assessing  the doses resulting from internal exposure, that
 is,  exposure resulting from the inhalation or  ingestion of contaminated
 material.   Portions  of the material inhaled or ingested may not leave
 the  body  for a considerable period of time (up to decades); therefore,
 dose  rates  are calculated  over a corresponding time interval.

      The  calculation of  internal doses requires the use of several
 models.   The most  important are the ICRP lung  model, depicted in Figure
 5.3-2,  and  the  gastrointestinal (61) tract model shown in Figure 5.3-3.
 The lung model  is  comprised of three regions,  the nasopharyngial (N-P),
 the tracheobronchial  (T-B), and the pulmonary  (P) regions.   A certain
portion of  the  radioactive material Inhaled is deposited in each of the
 three lung  regions (N-P, T-B, and P) indicated in Figure 5.3-2.  The
material is  then cleared (removed) from the lung to the blood and
gastrointestinal tract, as indicated by the arrows, according to the
 specified clearance parameters for the clearance class of the inhaled
material.
                                   5-4

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oc
o
                                     TIME
          Figure 5.3-1.
Typical pattern of decline of activity of a
radionuclide in an organ* assuming an initial
activity in the organ and no additional uptake
of radionuclide by the organ (ORNL81).
                                      5-5

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Ui
I
OS
Compartment
N-P a
(D3 = 0.30) b
T-B c
(D4 = 0.08) d
e
P f
(D, = 0.25) g
b h
L i
Class
D
T
0.01
0.01
0.01
0.2
0.5
n.a.
n.a.
0.5
0.5
F
0.5
0.5
0.95
0.05
0.8
n.a.
n.a.
0.2
1.0
W
T
0.01
0.4
0.01
0.2
50
1.0
50
50
50
F
0.1
0.9
0.5
0.5
0.15
0.4
0.4
0.05
1.0
Y
T
0.01
0.4
0.01
0.2
500
1.0
500
500
1000
F
0.01
0.99
0.01
0.99
0.05
0.4
0.4
0.15
0.9

B
L
0
0
D
T
a
^^f
c
e

L

*
\D,
N

r
U-P

£V.,-
f r i
b

h
p

b
d
— ^
— ^
g

G
I
T
R
A
C
T
                          Figure 5.3-2.  The ICRP Task Group lung model for particulates.

        The columns labeled D, W, and Y correspond, respectively, to rapid, intermediate, and slow
   clearance of the Inspired material (in days, weeks, or years).  The symbols T and F denote the
   biological half-time (days) and coefficient, respectively, of a term in the appropriate retention
   function.  The values shown for D_, D,, and D_ correspond to activity median aerodynamic diameter
   AMAD = 1 ym and represent the fraction of the inspired material depositing in the lung regions.

-------
                             INGESTION
                                i
RESPIRATORY
TRACT
1



B
L
0
0
D




4" /

ab
SI
Alll J
ULI
** xab
LLI
s

j

SI
\
ULI
1
LLI



X$ = 24 day"1


\SI - 6 day"1

XULI =1.85 day"1

XLLI - 1 day"
Figure 5.3-3.  Schematic representation of radionuclide
               movement among respiratory tract,
               gastrointestinal tract, and blood.
         S     - stomach
         SI    - small intestine
         ULI   - upper large intestine
         LLI   - lover large intestine
         \     - elimination rate constant
                          5-7

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     Deposition and clearance of inhaled materials in the lung are con-
trolled by the particle size and clearance class of the material.  The
particle size distribution of the airborne material is specified by
giving its Activity Median Aerodynamic Diameter (AMAD) in microns, u,
(one micron equals 10~^ meters).  Where no AMAD is known, a value of
1.0 micron is assumed.  Clearance classes are stated in terms of the
time required for the material to leave the lung, that is, Class D
(days), Class W (weeks), and Class Y (years).

     The gastrointestinal tract model consists of four compartments, the
stomach (S), small intestine (SI), upper large intestine (ULI), and
lower large intestine (LLI).  However, it is only from the small intes-
tine (SI) that absorption into the blood is considered to occur.  The
fraction of material that is transferred into blood is denoted by the
symbol f^.

     Radionuclides may be absorbed by the blood from either the lungs or
the GI tract.  After absorption by the blood, the radionuclide is dis-
tributed among body organs according to fractional uptake coefficients,
denoted by the symbol f^-  Since the radioactive material may be
transported through the body, dose rates are calculated for each organ
or tissue affected by using a model of the organ that mathematically
simulates the biological processes involved.  The general form of the
model for each organ is relatively simple.  It postulates that the
radioactive material which enters the organ is removed by both radio-
active decay and biological removal processes.

5.3.2  External Doses

     The example just described for modeling the activity of a radio-
nuclide in an organ pertains to estimating doses from internal
exposure.  In contrast, the external immersion and surface doses are
calculated as follows.  First, the number of photons reaching the body
is determined.  The model used here is a set of equations governing the
travel of photons (gamma radiation) in air.  The simplifying assumptions
used in these calculations are that the medium (air) is an infinite
half-space and is the only material present.  This makes the calculation
relatively straightforward.  In the second portion of the calculation,
the photons reaching the body are followed through the body using a
"Monte Carlo" method.  The "phantoms," i.e., the models of the body, are
those used by the Medical Internal Radiation Dose Committee (MIRD69).
The Monte Carlo method is a procedure in which the known properties of
the radiation and tissues are employed to trace (simulate) the paths of
a large number of photons in the body.  The amount of energy released at
each interaction of the radiation with body tissues is recorded and,
thus, the dose tp each organ or tissue is estimated by evaluating a
large number of photon paths.

5.3.3  Effects of Decay Products

     In calculating doses from internal and external exposures, the
occurrence of radioactive decay products (or daughters) must be
                                   5-8

-------
considered.  When some atoms undergo radioactive decay,  the new atom
created in the process may also be radioactive and may contribute to the
radiation dose.  Although these decay products may be treated as
Independent radionuclides in external exposures, the decay products  of
each parent must be followed through the body in Internal exposures.
The decay product contributions to the dose rate are included in the
dose calculations, based on the metabolic properties of the element  and
the organ in which they occur.

5.3.4  Dose Rate Estimates

     For each external and internal exposure, dose rates to each of  the
organs listed in Table 5.3-1 are calculated for each radloisotope.
These organ dose rates serve as input to the life table calculations
described in Chapter 6.

          Table 5.3-1.   Organs for which dose rates  are  calculated
             Red bone marrow                    Intestine
             Bone                               Thyroid
             Lung                               Liver
             Breast                             Urinary tract
             Stomach                            Other**'
             Pancreas
     (a'Esophagus, lymphatic system, pharynx, larynx, salivary gland,
brain.

5.4  EPA Dose Calculation

5.4.1  Dose Rates

     The models described in Section 5.2 are used by EPA to calculate
radiation dose rates resulting from internal and external exposures to
radioactive materials.  A more complete description of the methodology,
equations, and parameters used is given in Du84, ORNL80, and ORNL81.  EPA
has adopted two refinements to the ICRP-recommended protocol for these
calculations.  The first is to track the movement of internally produced
radioactive daughters by assuming that their movement is governed by
their own metabolic properties rather than those of the parent.  Although
not enough information is available to allow a rigorously defensible
choice, this appears to be more accurate for most organs and radionu-
clides than the ICRP assumption that daughters behave exactly as the
parent.  In the second departure from ICRP recommendations, age-dependent
values of the parameters governing the uptake of transuranic radionu-
clides have been taken from two sources deemed appropriate to the
                                    5-9

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 general population, the National Radiological Protection Board (NRPB82)
 and the EPA. transuranic guidance document (EPA77).

      The internal dose equations given by ICRP may  be used to calculate
 either radiation doses (rad),  I.e.,  the total dose  over a given time
 period, or radiation dose rates (rad/yr), i.e., the way in which the dose
 changes with time after intake.  The integral of the  dose rates is, of
 course, the total dose.  EPA calculates dose  rates  rather than doses,
 because EPA considers age when assessing the  effects  of radiation on the
 population.

      External Irradiation does not result in  any residual internal
 material.   Therefore,  external dose  rates to  a given  organ are constant
 for as long as the external  radionucllde is present.  That is,  the dose
 rate caused by a given amount  of radionuclide present in air  or on a
 ground surface becomes zero  when the radionucllde is  removed.

      The calculation of dose rates,  rather than integrated doses, allows
 the use of age-dependent metabolic parameters more  appropriate to the
 general population to  be taken into  account.   In the  vast majority of
 cases,  however,  there  is not now sufficient information available to make
 such calculations.   The effect of using age-dependent metabolic
 parameters is discussed in Section 5.2  for some radionuclides  for which
 sufficient information  is available.

 5.4.2  Exposure  and  Usage

      The ICRP dosioetric equations used by EPA are  linear,  i.e.,  an
 intake  of  10  picocuries  (pCi)  will result  in  dose rates  ten times as
 large as those from  an  intake  of  1 picocurie.   In similar fashion, ex-
 posure  to  ten times  as  large an air  or  ground surface concentration will
 increase the  external doses  by a  factor of ten.  EPA  uses this linearity
 to  avoid having  to calculate radiation  dose rates for a  range  of  concen-
 trations.   The standard  EPA  procedure is  to use unit  Intakes of 1 pCi/yr
 and air and ground surface concentrations  of  1 pCi/cm^ and  1 pCi/cm*,
 respectively.  The doses  for other intakes and concentrations  may then be
 scaled  up or  down as required.

      In most  cases,  it is necessary  to make certain assumptions regard-
 ing the exposure conditions in order to perform an assessment.  EPA cal-
 culates dose  rates for lifetime exposure to the unit Intakes and concen-
 trations.  Chapter 6 describes the different ways in which these rates
can be  applied.  In addition,  the exposure assessment will usually
depend  on other usage conditions assumed for  the exposures.  For  the
general population, EPA assumes a breathing rate equal to the  ICRP-
recommended values (ICRP75),  based on 8 hours of heavy activity, 8 hours
of light activity, and 8 hours of rest per day.  When  required, EPA uses
a drinking water intake of 2 liters per day.   The quantities of food
Ingested are compiled from a variety of sources.  Because there may be
insufficient data for some food types, it may be necessary to  combine or
substitute types in some instances.
                                   5-10

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5.5  Uncertainty Analysis

     Uncertainty, in the dose, refers to the manner in which the  calcu-
lated dose changes when the parameters used in the calculation (intakes,
metabolic factors, organ masses, etc.) are changed.  The uncertainty
associated with the dosimetric calculations is extremely difficult  to
quantify because the term "uncertainty analysis" implies a knowledge of
parameter distributions that is usually lacking.  Internal doses, for
example, depend on the parameters used to characterize the physiological
and metabolic properties of an individual, while external doses must
consider parameters such as organ mass and geometry for a particular
individual.  The data available for most of these parameters is not
sufficient to define the form of the parameter distribution.  The major
source of uncertainty in calculating the dose to a distinct individual,
however, in most instances, does not result from errors in measuring  the
parameters but from the real variation in parameter values among  indi-
viduals in the general population.  Thus, a calculated dose is thought
to be representative of a "typical" member of the general population and
is probably reasonably precise for some large segment of that population.

     The basic physiological and metabolic data used by EPA in calcu-
lating radiation doses are taken from the ICRP Report of the Task Group
on Reference Man (ICRP75) and from the ICRP Limits for Intakes of
Radionuclides by Workers (ICRP79).  The "Reference Man" report is the
most comprehensive compilation of data available on the intake, metabo-
lism, internal distribution, and retention of radioisotopes in the human
body.  Its major purpose, however, is to "define Reference Man,  in the
first instance, as a typical occupational individual," although  diffe-
rences with respect to age and sex are indicated in some instances.

     The limitations inherent in defining Reference Man, and in  estimat-
ing uncertainties due to variations in individuals in the general popu-
lation, are recognized by the Task Group (ICRP75):

          "The Task Group agreed that it was not feasible to
     define Reference Man as an  'average' or a  'median1 indivi-
     dual of a specified population group and that it was not
     necessary that he be defined in any such precise statistical
     sense.  The available data certainly do not represent a
     random sample of any specified population.  Whether the
     sample is truly representative of a particular population
     group remains largely a matter of judgement which cannot be
     supported on the basis of  statistical  tests of the data
     since the sampling procedure is  suspect.   Thus the Task
     Group has not always selected the  'average',  or  the
     'median*, of the available measurements  in making its
     selection, nor has it attempted  to limit the  sample  to some
     national or regional group and  then  seek an average  or
     median value.  However,  the  fact that  Reference  Man is not
     closely related to an existing  population  is  not believed to
     be of any great importance.  If  one  did have  Reference Man
                                5-11

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      defined  precisely as having  for  each attribute the median
      value  of a  precisely defined age group In precisely limited
      locality (e.g., males 18-20  years  of age In Paris, France,
      on June  1,  1964), these median values may be expected to
      change somewhat with time, and In  a few years may no longer
      be the median values for the specified population.  More-
      over,  the Reference Man so defined would not have this rela-
      tion to  any other population group unless by coincidence.
      To meet  the needs for which  Reference Man is defined, this
      precise  statistical relationship to a particular population
      is not necessary.  Only a very few individuals of any popu-
      lation will have  characteristics which approximate closely
      those  of Reference Man, however  he is defined.  The impor-
      tance  of the Reference Man concept is that his characteris-
      tics are defined  rather precisely, and thus if adjustments
      for individual differences are to  be made, there is a known
      basis  for the dose estimation procedure and for the estima-
      tion of  the adjustment factor needed for a specified type of
      individual."

      With respect to the dosimetrlc calculations performed by EPA to
assess  the  impact of radioactive  pollutants on a general population,
three sources of uncertainty should be  considered:

      (1)  that due to  the variation in  individual parameters among
          adults in the general population

      (2)  that due to  the variation in  individual parameters with age

      (3)  that due to  experimental error in the determination of
          specific parameters

      Each of  these sources of uncertainty is discussed in this section.
As noted above,  the data required to  perform a rigorous uncertainty
analysis are  lacking, and a form of uncertainty analysis called sensi-
tivity analysis  is employed.  The sensitivity analysis consists of sub-
stituting known  ranges in the parameters for the recommended value and
observing the  resulting change in the calculated dose rate.

5.5.1  Dose Uncertainty Resulting from Individual Variation

     This section discusses the uncertainty in calculated radiation
doses occasioned by differences in physical size and metabolism among
Individuals in the general population.  In order to investigate the
effects of individual differences in intake, size, and metabolism, it is
necessary to consider the form of the equation used to calculate radia-
tion dose rates.   Equation (5-2) is a simplified form of the one used by
EPA to represent the ingestion of radioactive materials.
                i(t) - c I fj f£  |  i  U-e~Xt]               (5-2)
                                   5-12

-------
    where D       is the dose rate (mrem/yr)
          I       is the intake rate of radioactive material (pCi/yr)
          fl      is the fraction of I transferred to blood after ingestion

          f£      is the fraction transferred to an organ from the blood
          n       is the mass of the organ (g)
          X       is the elimination constant, which denotes how rapidly
                  the activity is removed from the organ (yr-1)
          E       is the energy absorbed by the  organ for each radioactive
                  disintegration (ergs)
          c       is a proportionality constant.
For simplicity, we will assume that dose rates at large times,  t,  are  to
be studied so that the term in the bracket is approximately unity.

     Although the actual equations used are considerably more complicated
because they must describe the lung model and the 61 tract, and also
treat all radioactive progeny, the essential features of the uncertainty
in dose calculations are reflected in the terms of Equation (5-2).  The
sensitivity of the dose rate to each of the terms in the equation may be
studied by substituting observed ranges of the quantities for the single
value recommended by Reference Man.  For some of these quantities, as
noted below, no range is cited because of insufficient data.

     Intake, I

     As an example, postulate that the ingestion mode to be calculated is
for fluid intakes.  The average daily fluid intake is about 1900 ml, with
an adult range of 1000 to 2400 for "normal" conditions.  Under higher
environmental temperatures, this range may be increased to 2840 to 3410
ml per day.  Thus, a dose calculated as 1.9, for example, could range
from 1.0 to 2.4.

     Transfer Fraction, fj
     The value of the transfer fraction to blood depends on the chemical
form of the element under study.  One of the most common naturally oc-
curring radionuclides is uranium, which is used here as an example.
ICRP79 cites values of f^ ranging from 0.005 to 0.05 for industrial
workers, but notes that a higher value of 0.2 is indicated by dietary
data from persons not occupationally exposed.  EPA has used the 0.2 value
for the general population but, based on the ICRP range above, a calcu-
lated dose determination could vary by a factor of 10.

     Organ Mass, m

     The range of organ masses depends primarily on the organ under
investigation.  For example,  reported values for the bloodless lungs
range from 461 to 676 grams.  Liver weights ranged from 1400 to 2300
grams for adult males and 1200 to 1820 grams for females.  Thus, because
                                    5-13

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 the organ mass appears in the denominator,  calculated lung doses might be
 expected to vary by a factor of 1.5 and liver doses by a factor of
 about 2.
      Remaining Terms,  £3,  X,  E

      There are few reported data on the ranges in values  to  be expected
 for the remaining variables.   They are all quantities which  are less
 directly observable than I, fj_,  and m and their influence on the dose
 calculation can only be estimated.   The discussion in Section 5.6 is
 intended to augment the uncertainty analysis by Introducing  the results
 of some direct observations on segments of the general  population.

 5.5.2  Dose Uncertainty Resulting from Age

      The dose  rates calculated by EPA are normally based  on  the metabo-
 lism and physical characteristics of Reference Man (ICRP75).   These pro-
 perties may obviously  be expected to depend on the age  of an individual.
 Most particularly,  for Infants and  children such factors  as  breathing
 rates,  liquid  and solid intakes,  organ size and growth  rates,  and body
 geometry are known to  vary  considerably from adult values.   The effect of
 such changes on the radiation dose  also depends on the  chemistry of the
 radioactive element under study.  For example,  rapid bone growth in
 children is of more importance when a "bone seeker" such  as  strontium is
 considered.  Although  the data available for most age and chemical ele-
 ment combinations are  insufficient  to allow estimation  of the  uncertainty
 in dose rate,  some organ/element  combinations,  for which  more  information
 IB available,  are discussed below.

      Iodine  and the Thyroid

      Iodine  is rapidly  and  virtually completely absorbed  Into  the blood-
 stream  following inhalation or ingestion.   From the blood, iodine enters
 the extracellular fluid and quickly becomes  concentrated  in  the salivary,
 gastric,  and thyroid glands.   It  is rapidly  secreted from the  salivary
 and gastric  glands,  but  it  is  retained in the thyroid for relatively long
 periods.

      The  Intake  and  metabolism of iodine have been reviewed extensively
 to  develop an  age-dependent model for  iodine  (ORNL84a).   In the model
 used here, ingested  iodine  is  assumed  to be almost completely absorbed by
 the blood.   The  remaining parameters are age dependent and are  shown in
 Table 5.5-1.   The fluid  intake varies  from 0.72  liters per day for a new-
 born to about  2.0 liters per day for an  adult.

     These age-dependent parameters may  then be  used in Equation  (5-2) to
 calculate the  dose rate resulting from a constant  concentration of iodine
 in water and air.  The resulting curves  for the  dose rate as a function
of age are shown  in Figures 5.5-1 and  5.5-2.  These may be compared to
 the dose rates obtained using Reference Nan parameters at all ages,
indicated by the dotted lines in the same figures.  Thus, for this
                                    5-14

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             AGE-DEPENDENT MODEL
  0.0   10.0  20.0   30.0  40.0   50.0   60.0   70.0   80.0  90.0  100.0
                             AGE  (YEARS)
Figure 5.5-1.  Dose rate from chronic ingestion of iodine-131 in
               water at a concentration of 1
                              5-15

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                  AGE-DEPENDENT MODEL
  0.0   10.0   20.0   30.0   40.0   50.0   60.0   70.0   80.0  90.0  100.0
                             AGE (YEARS)
Figure 5.5-2.  Dose rate from chronic inhalation of iodine-131  in
               air at a concentration of 1 yCi/m3.
                              5-16

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particular combination of organ and isotope,  the total (70-year) dose is
seen to increase by about 30 percent for ingestion and 35  percent for
inhalation when dependence on age is considered.

       Table 5.5-1.  Age-dependent parameters for iodine metabolism
                              in the thyroid
Age Fractional uptake
(days) to thyroid, f£
Newborn
100
365
1825
3650
5475
7300
0.5
0.4
0.3
0.3
0.3
0.3
0.3
Biological half-time
Thyroid mass in the thyroid
(g) (days)
	
_
1.78
3.45
7.93
12.40
20.00
15
20
30
40
50
65
80
     Strontium and Bone

     Because of the chemical similarities of strontium and calcium,
strontium tends to follow the calcium pathways in the body and deposits
to a large extent in the skeleton.  In fact, the fraction of Ingested
strontium eventually reaching the skeleton at a given age depends largely
on the skeletal needs for calcium at that age, although the body is able
to discriminate somewhat against strontium in favor of calcium after the
first few weeks of life.

     The ICRP model for bone is more complicated than that for the
thyroid because it consists of more than one compartment.  For purposes
of modeling the transport of strontium by the skeleton, it suffices to
view the mineralized skeleton as consisting of two main compartments:
trabecular (cancellous, porous, spongy) and cortical (compact) bone.
Two subcompartments, surface and volume, are considered within each of
these main compartments.  The four subcompartmenta of mineralised
skeleton and the movement of strontium among these compartments are shown
schematically in Figure 5.5-3.  The equations governing the age depen-
dence of the parameters are given in ORNL84a.  Dose rate curves for the
inhalation and ingestion of constant concentrations of strontium-90 are
given in Figures 5.5-4 and 5.5-5.  The comparable curves for Reference
Man are again indicated by dashed lines.  Thus, for this element and
organ combination, the dose rate resulting from ingestion is somewhat
higher, while the dose rate resulting from inhalation exhibits only minor
perturbations, when the age dependence of the parameters is considered.
The lifetime (70-year) dose resulting from Ingestion is about 7 percent
greater and the inhalation dose less than 1 percent different when age
dependence is considered.
                                    5-17

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                          BLOOD
    I
TRABECULAR
  SURFACE
    1
TRABECULAR
  VOLUME
   1
CORTICAL
 SURFACE
CORTICAL
 VOLUME
  Figure 5.5-3.  Compartments and pathways in model for
                 strontium in skeleton.
                         5-18

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         AGE-DEPENDENT MODEL
  0.0   10.0   20.0   30.0  40.0   50.0   60.0  70.0  80.0   90.0 100.0
                          AGE (YEARS)
Figure 5.5-4.   Dose rate from chronic ingestion of  strontium-90
               in water at a concentration of 1 yCi/£.
                             5-19

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 90.0
                          AGE-DEPENDENT MODEL


                       ADULT MODEL
             AGE-DEPENDENT MODEL
     0.0  10.0  20.0   30.0  40.0  50.0  60.0   70.0   80.0   90.0  100.0

                                 AGE  (YEARS)
Figure 5.5-5.   Dose rate from chronic inhalation of strontium-90 in
               air at a concentration of 1  yCi/m3.
                               5-20

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     Plutonium and Lung and Red Bone Marrow

     Apparently plutonium and Iron bear sufficient chemical resemblance
that plutonium is able to penetrate some iron transport and storage
systems.  It has been shown that plutonium in blood serum complexes with
transferrin, the iron-transport protein.  Thus, plutonium will partially
trace the iron pathway, with the result that a substantial fraction of
systemic plutonium is carried to the bone marrow and to the liver.   In
the skeleton, plutonium may be released mainly at sites of developing red
cells.  Plutonium that has reached the skeleton behaves very differently
from iron; its movement is governed by fairly complicated processes of
bone resorption and addition.  Because the total metabolic behavior of
plutoniuot is not closely related to that of any essential element,  any
retention model for plutonium as a function of age will involve much
larger uncertainties than the analogous model for strontium.  Still,
there is enough information concerning the metabolism of plutonium by
mammals to justify an examination of potential differences with age In
doses to radiosensitive tissues following intake of this radionuclide.

     The effect of age-dependent parameters on dose rate calculations is
most evident for the lung when the Inhalation pathway is considered.
Figure 5.5-6 exhibits the variation in dose rate to the total and pulmon-
ary portions of the lung both for the adult and age-dependent cases.  The
increased dose rate from age 0 to about 20 is typically caused by varia-
tions in the breathing rate-lung mass ratio for infants and juveniles.
For this model, the age-dependent pulmonary lung 70-year dose is about 9
percent greater than for the adult model.

     To describe retention of plutonium in the skeleton, the skeleton is
viewed as consisting of a cortical compartment and trabecular compart-
ment.  Each of these is further divided into three (rather than two as
for strontium) subcompartments:  bone surface, bone volume, and a trans-
fer compartment.  The transfer compartment, which includes the bone
marrow, may receive plutonium that is removed from bone surface or vol-
ume; plutonium may reside in this compartment temporarily before being
returned either to the bloodstream or to bone surfaces (Figure 5.5-7).
Because of the large amount of recycling of plutonium among the skeletal
compartments, blood, and other organs, recycling Is considered explicitly
in the model.  The age-dependent features of the model are described in
detail in ORNL84a.

     Red bone marrow dose rates for the age-dependent model are shown in
Figure 5.5-8, for ingestlon, and in Figure 5.5-9, for Inhalation.  The
dashed curves are the dose rates using non-age-dependent parameters.  As
in the corresponding curves for strontium, the difference is more pro-
nounced for the ingestlon pathway.  Because of the long physical
half-life and biological half-time of plutonium in the skeleton, the dose
rate, for a chronic intake, does not reach equilibrium within the one
hundred year time period of the figures.  The total lifetime (70-year)
dose to the red marrow is about 25 percent greater for Ingestlon, and
nearly unchanged for inhalation when the age-dependent parameters are
used.
                                   5-21

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   40.0
_ ,      vAGE-DEPENDENT DOSE RATES AND INTAKE RATES (PULMONARY LUNG) .
                'AGE-DEPENDENT DOSE RATES AND INTAKE RATES (TOTAL LUNG)

               ADULT DOSE RATES AND INTAKE RATE  (PULMONARY  LUNG)

             ADULT DOSE RATES AND INTAKE RATE (TOTAL LUNG)
       0.0  10.0  20.0   30.0  40.0  50.0   60.0   70.0  80.0  90.0  100.0
                                AGE (YEARS)
 Figure 5.5-6.  Dose rate from chronic inhalation of plutonium-239 in
                air at a concentration of 1 yCi/m9.
                                 5-22

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                             i
                            BLOOD
  TRABECULAR
     SURFACE
   TRABECULAR
     VOLUME
TRABECULAR
  MARROW
                             I
                                            CORTICAL
                                            SURFACE
                                             CORTICAL
                                              VOLUME
                                                CORTICAL
                                                 MARROW
Figure 5.5-7.  Compartments and pathways in model for plutonium
               in- skeleton.
                              5-23

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 20.0
           AGE-DEPENDENT  MODEL
 0.0
    0.0  10.0  20.0   30.0  40.0   50.0  60.0  70.0   80.0  90.0  100.0

                             AGE (YEARS)
Figure 5.5-8.   Dose rate from chronic ingestion of  plutonium-239  in
               water at a concentration of 1  yCi/A.
                               5-24

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   6000.0
<£
O
5 5000.0
DC
§
oc
   4000.0
   3000.0
g
   2000.0
o-
UJ
   1000.0
      0.0
                      I       I      I
                    .AGE-DEPENDENT MODEL



                ADULT MODEL






j	i	i	i      i       i      i
        0.0   10.0  20.0   30.0  40.0  50.0   60.0  70.0  80.0   90.0  100.0

                                   AGE (YEARS)
  Figure 5.5-9.  Dose rate from chronic inhalation of plutonium-239 in

                 air at a concentration of 1
                                  5-25

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      In summary, It is difficult to make generalizations concerning the
 uncertainty involved in neglecting age dependence in the dose rate calcu-
 lations.  Although the examples given indicate higher dose rates for the
 ingestlon pathway, with smaller changes for Inhalation, when using
 age-dependent parameters, this results from the complex interaction
 between parameters in the dose rate equation and depends on the
 element/organ combination under consideration.

 5.5.3  Dose Uncertainty Caused by Measurement Errors

      The last potential source of uncertainty in the dose rate calcula-
 tions is the error involved in making measurements of fixed quantities
 (ORNL84b).  The radioactive half-life of an isotope, for example,  may be
 measured independently of any biological system, but the measurement is
 subject to some error.  The organ mass of a given organ may also be
 measured with only a small error.   Repeated determinations of these quan-
 tities, in addition, can reduce the error.   Although this source of
 uncertainty may be of importance in other aspects of an environmental
 assessment, it is  of little consequence in the dosimetry, because  it is
 overwhelmed by the magnitude of the uncertainties resulting from indivi-
 dual variations.

      Although consideration of the factors  described above Implies large
 uncertainties in calculated doses,  the actual variation is expected to  be
 considerably smaller.   The reasons for this,  and some supporting studies
 on real populations, are presented in Section 5.6.

 5.6  Distribution  of Doses in the  General Population

      Although the  use of extreme parameter  values in a sensitivity analy-
 sis indicates that large uncertainties in calculated doses  are  possible,
 this uncertainty is not  usually reflected in the general population.
 There are  several  reasons for this:   the  parameter values chosen are
 intended to be typical of an individual in  the population;  it is improba-
 ble that the "worst case"  parameters  would  be chosen for all terms  in the
 equation;  and not  all  of the  terms are mutually  independent, e.g.,  an
 Increased  intake may be  offset  by more rapid excretion.

      This  smaller  range  of uncertainty in real populations  is demon-
 strated  by  studies  performed  on  various human and  animal  populations.   It
 should be noted that there is always  some variability in observed doses
 that  results  primarily from differences in  the characteristics of indi-
 viduals.  The usual way  of specifying  the variability of  the dose, or
activity, in an organ is in terms of  the deviation from the average
value.   In  the following studies, it should also be noted that, in
addition to the variability resulting from individual  characteristics,
 the exposure levels of individuals may also have varied appreciably -
another factor tending to increase the dose uncertainty.  The following
studies are representative of those carried out on real populations:
                                   5-26

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     (1)  An analysis of the thyroid from 133 jackrabbits  in a nuclear
fallout area found that in only 2 did the iodine-131 content exceed  three
times the average (Tu65).

     (2)  Measurements of the strontium-90 content of adult whole skele-
tons showed that only about 5 percent of the population would exceed
twice the average activity, with only about 0.1 percent exceeding four
times the average (Ku62).

     (3)  In another study, the cesium-137 content of 878 skeletal muscle
samples was measured (E164a,b).  This radioisotope is also the  result of
nuclear tests so that the muscle content depends not only on the varia-
tion in individual parameters but also on the pathways leading  to inges-
tion or inhalation of the isotope.  Nevertheless, analyses of these
samples indicated that only 0.2 percent exceeded three times the average
activity at a 95 percent confidence level.

     (4)  A study of the variability in organ deposition among individuals
exposed under relatively similar conditions to toxic substances has also
been performed (Cu79).  In eleven exposure situations (Table 5.6-1), the
geometric standard deviation of the apparently lognormal distribution of
organ doses ranged from 1.3 to 3.4.  From the table, for example, 68
percent of the bone doses resulting from ingestion of strontium-90 would
lie between 0.56 and 1.8 times the average.

     In all but two of the situations examined,  there is the complicating
factor that there was probably a great deal of variation in the  exposure
levels experienced by members of the population.  The magnitude  of  geome-
tric standard deviations of  the studies listed in Table 5.6-1 may be the
evidence  of this variation since, except for  the two beagle studies, the
exposure  was not uniform.  Despite  these nonuniform exposures, however,
the organ dose is not  greatly affected probably  because of differences in
metabolic processes.   For  example,  there is  probably some  "self-adjust-
ment"  in  the amount  of strontium-90 absorbed from the  small intestine to
blood  of  different persons,  since strontium-90 tends to vary with calcium
in food;  if a person has a low calcium  intake, then he may absorb a
higher fraction  of the calcium and  strontium-90  than a person with  a high
calcium intake.

      In the  beagle studies,  the geometric  standard deviation is  1.8 for
inhaled metals  in bone or liver,  but is only 1.3 for ingested  strontium-
90 in bone.  An important difference is that all dogs ingesting  stron-
tium-90 at  a given level were administered the same amount, whereas, in
the  inhalation  studies, the exposure air concentrations were controlled
but  the dogs inhaled variable amounts depending upon their individual
characteristic  breathing patterns.

      Thus,  in  real situations, the overall uncertainty in dose is seen to
be considerably smaller than would be expected solely on a basis of the
 "worst case" sensitivity analyses.
                                     5-27

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Table 5.6-1.  Distributions of organ doses^ from inhalation and
             ingest!on of metals
Population    Exposure
                           Principal
                         exposure mode
       Geometric standard
Target    deviation of
 organ    organ doses (a'
Beagle
Humans
Humans
Humans
Humans
Metals
Plutonium
(fallout)
Titanium
(soil)
Aluminum
(soil)
Vanadium
Inhalation
Inhalation
Inhalation
Inhalation
Inhalation
Bone or liver 1.8
Lung 3(b)
Lung 3.4**'*
Lung 3.4^
Lung 3.4^b)
          (fossil  fuel
           combustion)
Beagles
Humans
Humans
(smokers)
Humans
(nonsookers)
Humans
Humans
Strontium-90
Strontium-90
(fallout)
Cadmium
Cadmium
Lead
Lead
Ingestion
Ingestion
Inhalation and
Ingestion
Inhalation and
Ingestion
Inhalation and
Ingestion
Inhalation
Bone
Bone
Kidney
Kidney
Bone
Lung
1.3
1.8(b)
1.8CW
i.a(b)
2.2(b>
1.7CD)
       stable element organ doses used In compiling this table
   were generally expressed in parts-per-million of organ mass.
        that exposure levels may vary considerably among
   individuals; If this factor could be eliminated, geometric
   standard deviations probably would be smaller.

Source:  (Cu79).
                               5-28

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5.7  Summary

     This chapter presents an overview of the methods used by EPA to es-
timate radiation doses.  The chapter defines the basic quantities report-
ed by EPA and describes briefly the models employed.  The chapter also
points out departures from the occupational parameters and assumptions
employed in the basic IGRP methodology and gives the reasons for the
deviations outlined.

     Many of the physiological and metabolic parameters recommended in
methods for calculating radiation doses are based on a limited number of
observations, often on atypical humans or on other species*  EPA has
attempted to bound the uncertainty associated with the ranges observed
for some of the more Important parameters used.  In fact, some empirical
data on population doses mentioned here indicate that actual dose uncer-
tainties are much less than is implied by this "worst case" analysis.
For the sources of uncertainty discussed, the large dose ranges possible
because of variation in individual characteristics must be modified by
consideration of the narrower ranges indicated by studies of real popula-
tions; the dose range resulting from age dependence appears to be small
for lifetime exposures, and the range resulting from experimental error
is negligible by comparison.  Based on these observations, it is reason-
able to estimate that EPA's calculated doses should be accurate within a
factor of 3 or 4.  It should be emphasized that much of the "uncertainty"
in the dose calculation is not caused by parameter error but reflects
real differences in individual characteristics within the general popula-
tion.  Therefore, the uncertainty in the dose estimates cannot be disso-
ciated from specification of the segment of the population to be
protected.

     More complete derivations and explanations for the EPA methodology
are given in the references cited in the text, and a technical descrip-
tion of the dose rate equations and their use in conjunction with the
life table risk evaluation is given in Appendix A.
                                    5-29

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                                 REFERENCES
 Cu79       Cuddihy R. G., McClellan R. D.,  and Griffith W.  C.,
            Variability In Target Organ Deposition among Individuals
            Exposed to Toxic Substances, Toxicology and Applied
            Pharmacology 49, 179-187, 1979.

 Du84       Dunning D. E.  Jr.,  Leggett R. W.,  and Sullivan R. E.,  An
            Assessment of Health Risk from Radiation Exposures,  in Health
            Physics £6 (5),  1035-1051, May 1984.

 E164a      Ellett W.  H. and Brownell G. L., Caeslum-137 Fail-Out  Body
            Burdens, Time Variation and Frequency Distributions, Nature
            203 (4940), 53-55,  July 1964.

 E164b      Ellett W.  H. and Brownell G. L., The  Time Analysis and
            Frequency Distribution of Caeslum-137 Fall-Out In Muscle
            Samples, IAEA Proceedings Series,  STI/PUB/84,  Assessment of
            Radioactivity  in Man,  Vol. II, 155-166,  1964.

            U.S.  Environmental  Protection Agency,  Proposed Guidance on
            Dose  Limits for  Persons Exposed to Transuranium  Elements in
            the General Environment,  EPA 520/4-77-016,  1977.

            International  Commission on Radiological Protection, Report of
            the Task Group on Reference Man, ICRP Publication No.  23,
            Pergamon Press,  Oxford,  1975.

            International  Commission on Radiological Protection, Recommen-
            dations  of the International Commission  on  Radiological
            Protection,  ICRP Publication No. 26,  Pergamon  Press, Oxford,
            1977.

ICRP79      International  Commission  on Radiological  Protection, Limits
            for Intakes  of Radionuclides by Workers,  ICRP  Publication
            No. 30,  Pergamon Press, Oxford, 1979.

Ku62        Kulp J. L. and Schulert A. R., Strontium-90 in Man V, Science
            136 (3516), May  1962.

MIRD69      Medical Internal Radiation  Dose Committee, Estimates of
           Absorbed Fractions for Monoenergenetlc Photon  Sources
            Uniformly Distributed in Various Organs of a Heterogeneous
            Phantom, MIRD Supplement No. 3,  Pamphlet 5, 1969.
EPA77
ICRP75
ICRP77
                                   5-30

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NRPB82    National Radiological Protection Board, Gut Uptake Factors for
          Plutonium, Americium, and Curium, NRPB-R129, Her Majesty's
          Stationery Office, London, England, 1982.

ORNL80    Oak Ridge National Laboratory, A Combined Methodology for
          Estimating Dose Rates and Health Effects for Exposure to
          Radioactive Pollutants, ORNL/RM-7105, Oak Ridge, Tennessee,
          1980.

ORNL81    Oak Ridge National Laboratory, Estimates of Health Risk from
          Exposure to Radioactive Pollutants, ORNL/RM-7745, Oak Ridge,
          Tennessee, 1981.

ORNL84a   Oak Ridge National Laboratory, Age Dependent Estimation of
          Radiation Dose, to be published.

ORNLSAb   Oak Ridge National Laboratory, Reliability of the Internal
          Dosimetric Models of ICRP-30 and Prospects for Improved Models,
          to be published.

Tu65      Turner F. B., Uptake of Fallout Radionuclides by Mammals and a
          Stochastic Simulation of the Process, in Radioactive Fallout
          from Nuclear Weapons Tests, U.S. AEC, Division of Technical
          Information, November 1965.
                                    5-31

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            Chapter 6:  ESTIMATING THE RISK OF HEALTH EFFECTS
                       RESULTING FROM RADIONUCLIDES
6.1  Introduction

    This chapter describes how EPA estimates the probability of fatal
cancer, serious genetic effects, and other detrimental health effects
caused by exposure to Ionizing radiation.  Such risk estimates are
complex and uncertain, even though much scientific effort has been ex-
pended to Increase the understanding of radiation effects.

    Because the effects of radiation on human health are known more
quantitatively than for most other environmental pollutants, it is
possible to make numerical estimates of the risk from a particular
source of radioactivity.  Such numbers may give an unwarranted aura of
certainty to estimated radiation risks.  Compared to the baseline inci-
dence of cancer and genetic defects, radiogenic cancer and radiation-
induced genetic defects do not occur very frequently.  Even among
heavily irradiated populations, the number of cancers and genetic
defects resulting from radiation is not known with either accuracy or
precision simply because of sampling variability.  In addition, exposed
populations have not been followed for their full lifetime, so that
information on ultimate effects is limited.  Moreover, when considered
In light of information gained from experiments with animals and from
various theories of carcinogenesls and mutagenesis, the observational
data on the effects of human exposure are subject to a number of Inter-
pretations.  This in turn leads to differing estimates of radiation
risks by both individual radiation scientists and expert groups.
Readers should bear in mind that estimating radiation risks is not a
mature science and that the evaluation of radiation hazards will change
as additional Information becomes available.  In this chapter, a number
of simple mathematical models are presented that may describe the main
features of the human response to radiation.  However, most scientists
would agree that the underlying reality is quite complicated and largely
unknown, so that such models should not be taken too literally but
rather as useful approximations that will someday be obsolete.

    EPA*s estimates of cancer and genetic risks in this report are based
on the Biological Effects of Ionizing Radiation (BEIR-3) report prepared
by the National Academy of Sciences (NAS) Committee in 1980 (NAS80).
This report was prepared for the purpose of assessing radiation risks at
the low exposure levels of interest in standard setting.  As phrased by
the President of the Academy, "We believe that the report will be helpful
                                    6-1

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 to the EPA and other agencies as they reassess radiation protection
 standards.  It provides the scientific bases upon which standards may be
 decided after nonscientific social values have been taken into account."

      In the sections below, we outline the various assumptions made in
 calculating radiation risks based on the 1980 HAS report and compare
 these risk estimates with those prepared by other scientific groups,
 such as the 1972 NAS BEIR Committee, the United Nations Scientific
 Committee on the Effects of Atomic Radiation (UNSCEAR), and the Interna-
 tional Commission on Radiation Protection (ICRP).  We recognize that
 information on radiation risks is incomplete and do not argue that the
 estimates made by the 1980 HAS BEIR Committee are highly accurate.
 Rather, we discuss some of the deficiencies in the available data base
 and point out possible sources of bias in current risk estimates.
 Nevertheless, we believe the risk estimates made by EPA are
 "state-of-the-art."

      In the sections below, we first consider the cancer risk resulting
 from whole-body exposure to low-LET* radiation,  i.e.,  lightly ionizing
 radiation like the energetic electrons produced  by X-rays or gamma
 rays.  Environmental contamination by radioactive materials also leads
 to the ingestion or inhalation of the material and subsequent concentra-
 tion of the radioactivity in selected body organs.   Therefore, the
 cancer risk resulting from low-LET irradiation of specific organs is
 examined next.   Organ doses can also result from hlgh-LET radiation,
 such as that associated with alpha particles.  The estimation of cancer
 risks for situations where hlgh-LET radiation is distributed more or
 less uniformly within a body organ Is the third  situation considered,
 Section 6.3.   In Section 6.4,  we  review the causes of  uncertainty in the
 cancer risk estimates and the magnitude  of this  uncertainty so that  the
 public as well  as EPA decision makers have a proper understanding of the
 degree of confidence to place in  them.   In Section 6.5, we review and
 quantify  the hazard of deleterious  genetic effects  from radiation and
 the  effects of  exposure in utero  on the  developing  fetus.   Finally,  in
 Section 6.6, we calculate  cancer  and genetic risks  from background
 radiation using the models described in  this chapter.

 6.2   Cancer Risk  Estimates for Low-LET Radiations

     Most  of the  observations of  radiation-induced  carcinogenesis  in
humans are on groups exposed  to low-LET  radiations.  These groups
include the Japanese A-bomb survivors and medical patients treated with
X-rays for ankylosing spondylitis in England from 1935 to 1954  (Sm78).
The UNSCEAR and the  HAS BEIR-3 Committee have provided knowledgeable
reviews of these and other data on the carcinogenic effects of human
exposures  (UNSCEAR77, NAS80).
     *Linear Energy Transfer (LET) — the energy deposited per unit of
distance along the path of a charged particle.
                                    6-2

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     The most important epidemiological data base on radiogenic cancer
 is the A-bomb survivors.  The Japanese A-bomb survivors have been
 studied for more than 38 years and most of them, the Life Span Study
 Sample, have been followed in a carefully planned and monitored
 epidemiological survey since 1950 (Ka82, Wa83).  They were exposed to a
 wide range of doses and are the largest group that has been studied.
 Therefore, they are virtually the only group providing information on
 the response pattern at various levels of exposure to low-LET
 radiation.  Unfortunately, the doses received by various individuals in
 the Life Span Study Sample are not yet known accurately.  The 1980 BEIR
 Committee's analysis of the A-bomb survivor data was prepared before
 bias in the dose estimates for the A-bomb survivors (the tentative 1965
 dose estimates, T65) became widely recognized (Lo8l).  It is now clear
 that the T65 doses tended to be overestimated so that the BEIR
 Committee's estimates of the risk per unit dose are likely to be too
 low (Bo82, RERF83.84).  A detailed reevaluation of current risk
 estimates is indicated when the A-bomb survivor data have been
 reanalyzed on the basis of new and better estimates of the dose to
 individual survivors.

     Uncertainties in radiation risk estimates do not result just from
 the uncertainties in the Japanese data base and In other epidemio-
 logical studies.  Analyses of these data bases require a number of
 assumptions that have a considerable effect on the estimated risk.
 These assumptions are discussed below.

 6-2.1  Assumptions Needed to Make Risk Estimates

     A number of assumptions must be made about how observations at
 high doses should be extrapolated to low doses and low dose rates for
 radiation of a given type i.e., high- or low-LET (LET).  These
 assumptions Include the shape of the dose response function and
 possible dose rate effects.  A dose response function expresses the
 relationship between dose and the probability that a radiogenic cancer
 is induced.  Observed excess cancers have occurred, for the most part,
 following relatively high doses of Ionizing radiation compared to those
 likely to occur as a result of the combination of background radiation
 and environmental contamination from controllable sources of
 radiation.  Therefore, a dose response model provides a method of
 interpolating between the number of radiogenic cancers observed at high
 doses and the number of cancers resulting from all causes including
 background radiation.

     The range of interpolation is not the same for all kinds of cancer
 because It depends upon the radiosensltlvity of a given tissue.  For
 example, the most probable radiogenic cancer for women is breast
 cancer.  As described below, breast cancer appears not to be reduced
when the dose is delivered over a long period of time.  For example,
 the number of excess cancers per unit dose among Japanese women who
 received acute doses, is about the same per unit dose as women exposed
 to small periodic doses of X-rays over many years.  If this Is actually
 the case, background radiation is as carcinogenic for bceast tissue as
                                                       .»>-•
                                   6-3

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 the acute exposures from A-bomb gamma radiation.  Moreover, the female
 A-bomb survivors show an excess of breast cancer at doses below 20 rads
 which is linearly proportional to that observed at several hundred rads
 (To84).  Women in their 40's, the youngest age group in which breast
 cancer is common, have received about 4 rads of whole-body low-LET
 background radiation and usually some additional dose Incurred for
 diagnostic medical purposes.  Therefore, for this cancer, the
 difference between observed dose producing radiogenic cancer, less than
 20 rads, and the dose resulting from background radiation Is less than
 a factor of five, not several orders of magnitude as is sometimes
 claimed.  However, it should be noted that breast tissue is a
 comparatively sensitive tissue for cancer Induction and that for most
 cancers, a statistically significant excess has not been observed at
 doses below 100 rads, low-LET.  Therefore, the range of dose
 interpolation between observed and calculated risk is often large.

 6.2.2  Dose Response Functions

      The 1980 NAS report examined three dose response functions in
 detail:  (1) linear, in which effects are directly proportional to dose
 at all doses; (2) linear quadratic,  in which effects are very nearly
 proportional to dose at very low doses and proportional to the square
 of the dose at high doses;  and (3) a quadratic dose response function,
 where the risk varies as the square  of the dose at all doses (NAS80).

      We believe the first two of these functions are compatible with
 most of the data on human cancer.  Information which became available
 only after the BEIR-3 report was published indicates that a quadratic
 response function is inconsistent with the observed excess risk of
 solid cancers at Nagasaki, where the estimated gamma-ray doses are  not
 seriously confounded by an assumed neutron dose component.   The chance
 that a quadratic response function underlies the excess  cancer observed
 in the Nagasaki incidence data has been reported as only 1 in 10,000
 (Wa83).   Although a quadratic response function is  not Incompatible
 with the Life Span Study Sample  data on leukemia incidence at  Nagasaki,
 Beebe and others have  pointed out  how unrepresentative these data are
 of the total  observed  dose response  for leukemia in that city (Be78,
 £177).   There is  no  evidence  that  a  quadratic  response function
 provides  a better fit  to  the  observed leukemia excess among all A-bomb
 survivors in  the  Life  Span Study  Sample  than a simple linear model
 (NAS80).  Based on these  considerations, we  do not  believe  a quadratic
 response  can  be used in a serious  effort to  estimate cancer risks due
 to ionizing radiation.  EPA notes  that neither the National Council on
Radiation Protection and Measurements  (NCRP),  the 1CRP, nor other
authoritative scientific groups, e.g., National Radiological Protection
Board  (NRPB) and UNSCEAR, have used a quadratic response function to
estimate  the risks due to ionizing radiation.

     The 1980 NAS BEIR Committee considered only the Japanese mortality
data in their analysis of possible dose response functions.  Based on
the T65 dose estimates, this Committee showed  that the excess  incidence
of solid cancer and leukemia among the A-bomb  survivors is  compatible
                                   6-4

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with either a linear or linear quadratic dose response to the low-LET
radiation component and a linear response to the high-LET neutron
component (HAS80).  Although the 1980 BEIR report indicated low-LET
risk estimates based on a linear quadratic response were "preferred"  by
most of the scientists who prepared that report, opinion was not
unanimous, and we believe the subsequent reassessment of the A-bomb
dose seriously weakens the Committee's conclusion.  The Committee's
analysis of dose response functions was based on the assumption that
most of the observed excess leukemia and solid cancers among A-bomb
survivors resulted from neutrons.  Current evidence, however, is
conclusive that neutrons were only a minor component of the dose in
both Hiroshima and Nagasaki (Bo82, RERF83,84).  Therefore, it is likely
that the linear response attributed to neutrons was caused by the gamma
dose, not the dose from neutrons.  This point is discussed further in
Section 6.4.

     Reanalysis of the Japanese experience after completion of the dose
reassessment may provide more definitive information on the dose
response of the A-bomb survivors, but it is unlikely to provide a
consensus on the dose response at environmental levels, i.e., about 100
mrads per year.  This is because at low enough doses there will always
be sampling variations in the observed risks so that observations are
compatible, in a statistical sense, with a variety of dose response
functions.  In the absence of empirical evidence or a strong
theoretical basis, a choice between dose response functions must be
based on other considerations.

     Although there is evidence for a nonlinear response to low-LET
radiations in some, but not all, studies of animal radiocarcinogenesls
(see below), we are not aware of any data on human cancers that are
incompatible with a simple linear model.  In such a case, it may be
preferable to adopt the simplest hypothesis that adequately models the
observed radiation effect.  Moreover, EPA believes that risk estimates,
for the purpose of assessing radiation impacts on public health, should
be based on scientifically creditable risk models that are unlikely to
understate the risk.  The linear model fulfills this criteria.  Given
the current bias in the doses assigned to A-bomb survivors (see Section
6.5.1 below), such an approach seems reasonable, as well as prudent.
Therefore, in this chapter, EPA has used the BEIR-3 linear dose
response model as one of two dose response models for discussing the
risk of radiogenic cancer due to low-LET radiations.  For low-LET
radiations, we have also included discussions of risk that are based on
the BEIR-3 linear quadratic dose response model.  While in the dose
range of interest (environmental levels) the dose squared term in this
model is insignificant, the linear term is about 2.5 times smaller than
that in the BEIR-3 linear response model.  That is, for the same dose,
risk estimates based on the BEIR-3 linear quadratic dose response model
are only 40 percent of those based on the BEIR-3 linear model.

     Many of the risk estimates needed to evaluate the effect of
radionuclide releases must be made on an organ specific basis.  The
BEIR-3 report provides risk coefficients for individual solid cancers
                                   6-5

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 only for the linear model in Tables V-14 and V-15.  (Tables identified
 with a V refer to original tables in NAS80 and are not reproduced in
 this report).  We have therefore divided BEIR-3 organ risk estimates
 for a linear response by a factor of 2.5 to obtain organ specific
 linear quadratic risk coefficients.

      The underlying basis for a linear quadratic response is thought to
 be that repair of radiation damage mitigates the effect of small doses
 of radiation or those which occur over a long time period, the reduced
 linear term being indicative of this repair.  Use of a linear quadratic
 dose response function, as formulated by the BEIR-3 Committee, is
 equivalent to the use of a dose rate effectiveness factor (DREF) of 2.5
 (see below).

      The discussions of both the linear and the linear quadratic dose
 response models for low-LET radiations are Included in this chapter to
 compare the risk estimates obtained for given doses using both models.
 The more conservative of these two models is the linear model.  We  have
 used this model for the calculation of the fatal cancers per curie
 released to the accessible environment.   This policy was thoroughly
 reviewed and accepted by the High-level Radioactive Waste Disposal
 Subcommittee of the EPA Science Advisory Board (EPA84).

 6.2.3  The Possible Effects of Dose Rate on Radiocarcinogenesis

      The BEIR-3 Committee limited its  risk estimates to  a minimum dose
 rate of 1 rem per year and stated that it "does not know if dose rates
 of gamma rays and X-rays  of about 100  mrad/y are detrimental to man."
 At dose rates comparable  to the annual dose  that everyone receives  for
 naturally-occurring radioactive materials, a considerable body of
 scientific opinion holds  that  the effects of radiation are reduced.
 The NCRP Committee 40 has suggested that carcinogenic  effects of
 low-LET radiations may be a factor of  from 2 to 10  times  less for small
 doses and dose  rates  than have  been observed at high doses (NCRP80).

     The  low dose  and low dose  rate  effectiveness factors  developed by
 the NCRP  Committee 40 are based on their analysis of a large  body of
 plant and animal data that  showed reduced effects at low  doses  for a
 number  of biological  endpoints,  including radiogenic cancer in animals,
 chiefly rodents.   However, no data  for cancer in humans confirm  these
 findings  as  yet.  A few human studies contradict them.  Highly
 fractionated  small  doses  to human breast  tissue  are  apparently as
 carcinogenic  as large  acute doses  (NAS80, La80).  Furthermore, small
 acute (less  than 10 rads) doses  to the thyroid are as  effective per rad
 as much larger doses  in initiating thyroid cancer (UNSCEAR77, NAS80).
Moreover, the increased breast  cancer resulting  from chronic low dose
 occupational gamma ray exposures among British radium  dial painters is
 comparable to, or larger  than that expected on the basis of acute high
dose exposures (Ba81).

     While none of these examples is persuasive by Itself, collectively
they indicate that it may not be prudent to assume that all kinds of
                                   6-6

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cancer are reduced at low dose rates and/or low doses.   However, It may
be overly conservative to estimate the risk of all cancers  on the  basis
of the linearity observed for breast and thyroid cancer. The ICRP and
the UNSCEAR have used a dose rate effectiveness factor  of about  2.5 to
estimate the risks from occupational and environmental  exposures
(ICRP77, UNSCEAR77).  Their choice of a DREF is fully consistent with
and equivalent to the reduction of risk at low doses obtained by
substituting the BEIR-3 linear quadratic response model for their
linear model.  Use of both a DREF and a linear quadratic model for risk
estimation is inappropriate (NCRP80).

     The difference between risk estimates obtained with the BEIR-3
linear and linear quadratic dose response models is by  no means  the
full measure of the uncertainty in the estimates of the cancer risk
resulting from ionizing radiation.  (Section 6.4 below  summarizes
information on uncertainty).  Using two models serves as a  reminder
that there is more than one creditable dose response model  for
estimating radiation risks and that it is not known if  all  radiogenic
cancers have the same dose response.

6.2.4  Risk Projection Models

     None of the exposed groups have been observed long enough to
assess the full effects of their exposures, if, as currently thought,
most radiogenic cancers occur throughout an exposed person's lifetime
(NAS80).  Therefore, another major decision that must be made in
assessing the lifetime cancer risk due to radiation is  to select a risk
projection model to estimate the risk for a longer period of time than
currently available observation data will allow.

     To estimate the risk of radiation exposure that is beyond  the
years of observation, either a relative risk or an absolute risk
projection model (or a suitable variation) may be used.  These models
are described at length in Chapter 4 of the 1980 HAS report  (NAS80).  A
relative risk projection model projects the currently observed
percentage increase in cancer risk per unit dose into future years.  An
absolute risk model projects the average observed number of excess
cancers per unit dose into future years at risk. -

     Because the underlying risk of  cancer increases rapidly with age,
the relative risk model predicts a larger probability of excess cancer
toward  the end  of a person's lifetime.  In contrast, the absolute risk
model predicts  a constant  incidence  of  excess  cancer across  time.
Therefore, given the  incomplete data we have now,  a relative  risk model
projects somewhat greater  risk than that  projected using an absolute
risk model.

     The National Academy  of  Sciences  BEIR Committee and other
scientific groups,  e.g., UNSCEAR, have not  concluded which projection
model is  the appropriate choice  for most  radiogenic cancers.  However,
evidence  is  accumulating which favors the relative risk projection
                                    6-7

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  model for most solid cancers.   As pointed  out  by  the  1980 MAS BEIR
  Committee,

           "If  the  relative-risk  model applies,  then the age of the
           exposed  groups,  both at  the time  of exposure and as they
           move through life,  becomes very important.   There is now
           considerable evidence  in nearly all the  adult human
           populations studied that persons  Irradiated  at higher
           ages have,  in general, a greater  excess  risk of cancer
           than those  irradiated  at lower ages, or  at least they
           develop  cancer sooner.   Furthermore, if  they are
           irradiated  at a  particular age, the excess risk tends to
           rise pari passu  (at equal pace) with the risk of the
           population  at large.   In other words, the relative-risk
           model with  respect  to  cancer susceptibility at least as
           a function  of age, evidently applies to some kinds of
           cancer that have been observed to result from radiation
           exposure."  (NAS80, p.33)

      This observation is confirmed by the Ninth A-bomb Survivor Life
 Span Study, published two years after the 1980 Academy report.   This
 latest report indicates that, for solid cancers, relative risks have
 continued to remain constant in recent years while absolute risks have
 increased substantially (Ka82).   Smith and Doll have  reached similar
 conclusions on the trend in excess cancer with time among the
 irradiated spondylitic patients  (Sm78).

      Although we believe considerable weight should be given to the
 relative risk model for  most solid cancers  (see below), the  model does
 not necessarily give an  accurate projection of  lifetime risk.   The mix
 of tumor types varies with age so  that  the  relative frequency of some
 common radiogenic  tumors,  such as  thyroid cancer,  decreases  for  older
 ages.   Land has pointed  out  that this may result in overestimates of
 the lifetime risk  when they are  based on a  projection  model using
 summed sites relative risks  (La83). While  this  may turn  out  to  be true
 for estimates  of cancer  incidence  that include  cancers less likely to
 be fatal,  e.g.,  thyroid, it  may  not be too  important in estimating the
 lifetime risk  of fatal cancers since the incidence of  most of the
 common fatal cancers, e.g.,  breast and lung cancers, increases with  age.

     Leukemia  and  bone cancer are  exceptions to  the general validity of
a  lifetime expression period  for radiogenic cancers.   Most, if not all,
of the  leukemia  risk  has apparently already been expressed in both the
A-bomb  survivors and  the spondylitics (Ka82, Sm78).  Similarly, bone
sarcoma from acute exposure appears to have a limited  expression  period
(NAS80, Ma83).  For these  diseases, the BEIR-3 Committee believed that
an absolute risk projection model  with a limited expression period is
appropriate for estimating lifetime risk (NAS80).

     Note that, unlike the NAS72 (BEIR-1) report, the BEIR-3
Committee's relative and absolute  risk models are age dependent.  That
is, the risk coefficient changes, depending on the age of the exposed
                                   6-8

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persons.  Observation data on how cancer risk resulting from radiation
changes with age is sparse, particularly so in the case of childhood
exposures.  Nevertheless,  the explicit consideration of the variation
in radiosensitivity with age at exposure is a significant improvement
In methodology.  It is important to differentiate between age
sensitivity at exposure and the age dependence of cancer expression.
In general, people are most sensitive to radiation when they are
young.  In contrast most radiogenic cancers occur late in life,  much
like cancers resulting from other causes.  In this chapter we present
risk estimates for a lifetime exposure of equal annual doses.  The
cancer risk estimated is lifetime risk from this exposure pattern.
However, age-dependent analyses using BEIR-3 risk coefficients Indicate
that the risk from one year of exposure varies by a factor of at least
five, depending on the age of the recipient.

6.2.5  Effect of Various Assumptions on the Numerical Risk Estimates

     Differences between risk estimates made by using various
combinations of the assumptions described above were examined in the
1980 NAS report.  Table 6.2-1, taken from Table V-25, shows the range
of cancer fatalities that are induced by a single 10-rad dose as
estimated using linear, linear quadratic, and quadratic dose response
functions and two risk projection models, relative and absolute (NAS80).

     As illustrated in Table 6.2-1, estimating the cancer  risk for a
given risk projection model on the basis of a quadratic as compared to
a linear dose response reduces the estimated risk of  fatal cancer by a
factor  of nearly 20.  Between the more  credible linear and linear
quadratic response functions, the difference is less, a  factor of about
two and a half.  For a given dose response model, results  obtained with
the two projection models,  for solid  cancers, differ  by  about a factor
of three.

     Even though the 1980  NAS analysis  estimated  lower  risks for  a
linear  quadratic response,  it  should  not be  concluded that this
response  function always  provides  smaller  risk  estimates.   In contrast
to the  1980  NAS analysis  where  the  proportion of  risk due to the  dose
squared term (e.g.,  03  in footnote  c  of Table 6.2-1)  was constrained
to positive  values,  the  linear quadratic function (which agrees best
with Nagasaki cancer incidence data)  has a negative coefficient for  the
dose  squared term  (Wa83).   Although this negative coefficient is  small
and  indeed may not  be  significant,  the computational result is  a  larger
linear term  which leads to higher risk estimates at low doses than
would be  estimated using a simple linear model (Wa83).   Preliminarily,
the  BEIR-3 analyses of the mortality, which were not restricted to
positive  coefficients of the dose squared  terms,  yielded similar
results.

      Differences in the estimated cancer risk introduced by the choice
of the risk projection model are also appreciable.  As pointed out
 above, the 1980 NAS analysis indicates that relative risk estimates
 exceed absolute risk estimates by about a factor of three.  However,
                                    6-9

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  relative  risk  estimates are quite  sensitive  to how the risk resulting
  from  exposure  during  childhood persists  throughout life.  This question
  Is addressed in  the next  section,  where  we compare risk estimates made
  by the 1972 and  1980  NAS  BEIR Committees with those of the ICRP and
  UNSCEAR.
    Table 6.2-1.  Range of cancer fatalities induced by 10 rads of low-LET
         radiation (Average value per rad per million persons  exposed)
     Dose response                      Lifetime risk projection model

      functions                          _ ,     fa\         ..
                                         Relative^8'         Absolute
Linear(b>
Linear Quadratic
Quadratic(d)
501
226
28
167
77
10
      '^Relative risk projection for all solid cancers  except
         leukemia and bone cancer fatalities,  which are  projected by
         means of the absolute risk model.
         Response R varies as a constant times the dose,  i.e., R-CiD.
 Source:   NAS80,  Table V-25.


 6.2.6 Comparison of  Cancer Risk Estimates  for Low-LET Radiation

      A number  of estimates of  the risk of fatal cancer following
 lifetime  exposure are compared in Table 6.2-2.  Although all of these
 risk  estimates assume a linear response function, they differ
 considerably because  of other  assumptions.  In contrast with absolute
 risk  estimates,  which have increased since  the 1972 NAS report (BEIR-1)
 was prepared,  the 1980 NAS BEIR-3 Committee's estimates of the relative
 risk,  as  shown in Table 6.2-2, have decreased relative to those in the
 BEIR-1 report.   This  illustrates the sensitivity of risk projections to
 changes in modeling assumptions.  In NAS80, the relative risk observed
 for ages  10 to 19 was  substituted for the considerably higher relative
 risk observed  for those exposed during childhood, ages 0 to 9.  In
 addition, the  relative risk coefficients used in the BEIR-3 analysis
 are based on excess cancer in  the Japanese A-bomb survivors compared to
 U.S. population  cancer mortality rates.  In NAS72, this excess was
 compared  to cancer mortality in Japan.  Moreover, the difference
 Introduced by  these two changes, particularly the former, is somewhat
greater than indicated in the 1980 NAS report.  The relative
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risk estimate attributed to the BEIR-1 Committee in the NAS  1980  report
is incorrect.  Therefore, two BEIR-l relative risk estimates are  listed
in Table 6.2-2;  the risk estimate in NAS80 attributed to the BEIR-1
Committee and an estimate which is based on the risk coefficients in
NAS72.  The BEIR-3 estimate did not use the relative risk coefficient
for childhood exposure given in the BEIR-1 report, which for solid
cancers is a factor of 10 larger than adult values (p. 171 in HAS72),
but rather used the adult risk for all ages including children.   The
estimate in Table 6.2-2 labeled NAS72 uses the relative risk
coefficients actually given in the BEIR-1 report.
Table 6.2-2.  A comparison of estimates of the risk of fatal cancer from a
  lifetime exposure at 1 rad/year (low-LET radiation)
Source of
estimate
BEIR-1 (NAS72)Jaj
BEIR-1 (NAS80)
BEIR-3 (NAS80) 100 rads.
None. Low dose/dose rate.
None. Occupational —
  low dose/dose rate.
UNSCEAR77 — without A-bomb
  data
     (a>BEIR-l relative risk model.
     triable V-4 in NAS80, linear dose response.             _
     (C;L_L absolute risk model for bone cancer and leukemia; L-L relative
        risk model for all other cancer.
     Jd>Table V-4 in NAS80 linear-quadratic dose response.
     tej Paragraphs 317 and 318 In UNSCEAR77.
     By comparing the three relative risk estimates in Table 6.2-2, it is
apparent that the relative risk estimates are fairly sensitive to the
assumptions made as to what extent the observed high relative risk of
cancer from childhood exposure continues throughout adult life.  The Life
Span Study indicates that the high-risk adult cancer caused by childhood
exposures is continuing, although, perhaps, not to the extent predicted by
the NAS BEIR-1 Committee (Ka82).

     The major reason the two sets of risk estimates In  Table 6.2-2 differ
is because of the underlying assumption in each set.  The NAS BEIR
estimates are for lifetime exposure and lifetime  expression of induced
                                    6-11

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  cancers (NAS72.80).   Neither the age distribution of the population at
  risk nor the projection models (if any) have been specified by either the
  UNSCEAR or the ICRP.   UNSCEAR apparently presumes the same  age
  distributions as occurred in the epidemiological studies they  cited,
  mainly the A-bomb survivors,  and a 40-year period of cancer expression.
  The ICRP risk estimates are  for adult workers,  presumably exposed between
  ages 18 and 65,  and a similar expression period.   These  are essentially
  age-independent  absolute risk models with less  than  lifetime expression of
  induced cancer mortality.  For these reasons alone,  risks estimated by
  ICRP and UNSCEAR are  expected to be smaller than those made on the basis
  of  the BEIR-3 report.

       The last entry in  Table  6.2-2  is of  interest because it specifically
  excludes the  A-bomb survivor  data based on T65 dose  estimates  (Ch83).  The
  authors reanalyzed the  information  on radiogenic cancer in  UNSCEAR77 so as
  to  exclude  all data based on  the Japanese  experience.  Their estimate of
  fatalities  ranges from 100 to 440 per 106  person rad for high doses and
  dose  rates.  As  Indicated in Table  6.2-2,  this is somewhat greater but
  comparable  to the UNSCEAR estimate, which  includes the A-bomb survivor
  data.   The mean number of fatalities given in Ch83 is 270 per 106
  person-rem, which is nearly identical to the value EPA has used for a
  linear  dose response model—280 fatalities per 106 person rad (see
  below).

  6.2.7  EPA Assumptions About Cancer Risks Resulting from Low-LET
        Radiation

      EPA1s discussion of radiation risks in this chapter is  based on
 presumed linear and linear quadratic dose response functions.  We believe
 these are the most credible dose response functions for estimating risks
 to exposed populations.   Using the BEIR-3 linear quadratic model is
 equivalent, at low dose, to using a dose rate effectiveness  factor of
 2.5.  As stated earlier,  we  have used a linear  dose  response function for
 low-LET radiation in  computing the fatal cancers per  curie released  to the
 accessible environment.

      Except for leukemia and  bone cancer,  where  we use a  25-year
 expression period for  radiogenic cancer,  we use  a lifetime expression
 period, as was done in the NAS report (NAS80).  Because the  most recent
 Life Span Study Report Indicates absolute risks  for solid cancers are
 continuing to  increase 33 years after exposure,  the 1980  NAS Committee
 choice of a lifetime expression period appears to be  well founded (Ka82).
 We do  not believe limiting cancer expression to  40 years  (as has been done
 by the ICRP and UNSCEAR)  is compatible with the  continuing increase in
 solid  cancers  that has occurred  among irradiated  populations (Ka82).
Analyses of  the spondylitic data have led  others  to similar  conclusions
 (Sm78).

     To  project the number of  fatalities resulting from leukemia and bone
cancer,  EPA uses  an absolute risk model, a  minimum induction period of 2
years, and a 25-year expression  period.  To estimate  the number  of fatali-
ties resulting from other cancers, EPA uses  the arithmetic average of
absolute and relative risk projection models.  For these cancers, we

                                     6-12

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assume a 10-year minimum induction period and expression of radiation-
induced cancer for the balance of an exposed person's lifetime after the
minimum Induction period.

6.2.8  Methodology for Assessing the Risk of Radiogenic Cancer

     EPA uses a life table analysis to estimate the number of fatal
radiogenic cancers in an exposed population of 100,000 persons.  This
analysis considers not only death due to radiogenic cancer, but also the
probabilities of other competing causes of death which are, of course,
much larger and vary considerably with age (BuBl, Co78).  Basically, it
calculates for ages 0 to 110 the risk of death due to all causes by
applying the 1970 mortality data from the National Center for Health
Statistics to a cohort of 100,000 persons (NCHS75).  Additional
information on the details of the life table analysis is provided in
Appendix A.  It should be noted that a life table analysis is required to
use the age-dependent risk coefficients in the BEIR-3 report.  For
relative risk estimates, we use age-specific cancer mortality data also
provided by NCHS (NCHS73).  The EPA computer program we use for the life
table analysis was furnished to the NAS BEIR-3 Committee by EPA and used
by the Committee to prepare its risk estimates.  Therefore, we believe
that the population base and calculational approach are similar in both
the NAS and EPA analyses.

     To project the observed risks of most solid radiogenic cancers beyond
the period of current observation, we use both absolute and relative risk
models, but usually present an arithmetic average based on these
projections.  Using a single estimate, instead of a range  of values, does
not mean that our estimate is precise.  As indicated in Table  6.2-2, the
range of estimated fatal cancers resulting from  the choice of  a particular
projection model and its internal assumptions is about a factor of  three.
Although we think it is  likely that the relative risk model is the  best
projection model for most solid cancers, it has  been tested rigorously
only for lung and breast cancer (La78).  Until It has more empirical
support, we prefer to use an average risk based  on both projection
models.  A second reason for this choice is  to avoid overly conservative
risk estimates caused by the compounding of multiplicative conservative
assumptions.

     To estimate  the cancer risk  from  low-LET, whole-body, lifetime
exposure with the linear model, we  use the_arithmetic average  of  relative
and absolute risk projections  (the  BEIR-3 L-L model) for  solid cancers and
an absolute risk  projection for leukemia and  bone  cancer  (the  BEIR-3 L-L
model).  For a dose to  the whole-body, this  yields an estimated  280
fatalities per million  person rad.   For the  BEIR-3 linear quadratic model,
which  is equivalent to  assuming a DREF of  2.5, a low-LET whole body dose
yields an estimated life risk of  about 110  fatalities  per million person
rad.

     These risk  estimates are not unduly conservative.   More than 235 of
the  280 fatalities  estimated  with the  BEIR-3 linear model result from
cancers in soft  tissues for which we have used the BEIR-3 L-L model.  As
                                     6-13

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 explained on page 187 of NAS80, the L-L model is not derived from the
 observed risk of solid cancers alone but rather includes parameters based
 on the Committee's analysis of the leukenia mortality data.  Therefore, as
 outlined in Section 6.4, the BZIR-3 Committee's analysis of the Japanese
 leukemia data depended heavily on the assumption that aost of the leukemia
 observed at Hiroshima was caused by neutrons.  In contrast, Table V-30 in
 the BEIR-3 report estimates the risk of cancer incidence in soft tissues
 directly, without the additional assumptions contained in the BEIR-3 L-L
 model.  By using the weighted Incidence mortality ratios given in the
 Table V-15, the results given in Table V-30 can be expressed in terms of
 mortality to yield, for lifetime exposure, an absolute risk estimate of
 about 200 fatalities per 106 person rad and about 770 fatalities per
 10" person rad when a relative risk projection model is used to estimate
 lifetime risk.  The arithmetic mean of the fatalities projected by these
 two models is almost 500 per 106 person rad,  more than twice as many
 fatal soft tissue cancers as predicted by the BEIR-3 L-L model and about 5
 times as many as estimated using the BEIR-3 linear quadratic model.

 6«2.9  Organ Risks

      By a whole-body dose,  we mean a uniform dose to every organ in the
 body.  In practice,  such exposure situations  seldom occur,  particularly
 for ingested or inhaled radioactivity.   This  section describes how we
 apportion this risk estimate for whole-body exposure when considering the
 risks following the exposure of specific organs.

      For most sources  of environmental contamination,  inhalation and
 ingestion of radioactivity  are more common than direct exposure.   In many
 cases, depending on the chemical and physical characteristics of the
 radioactive material,  inhalation and ingestion result  in a  nonuniform
 distribution of radioactive materials within  the  body  so that some organ
 systems receive ouch higher doses than  others.  For example,  iodine
 isotopes concentrate in the thyroid gland,  and the  dose to  this organ can
 be  orders of magnitude  larger  than the  average dose to the  body.

      Fatal Cancer at Specific  Sites

      To  determine the probability  that  fatal  cancer occurs  at  a particular
 site,  we  have  performed life table analyses for each cancer type using  the
 information  on  cancer incidence and mortality in NAS80.   For  cancer other
 than  leukemia and bone cancer, we  used  NAS80  Table  V-14 (Age Weighted
 Cancer Incidence by Site Excluding Leukemia and Bone Cancer) and NAS80
 Table  V-15, which lists the  BEIR Committee's  estimates  of the  ratio of
 cancer fatality to cancer incidence for these various organs.   The
proportions of leukemia and  fatal  bone  cancer caused by  low-LET radiation
were estimated using the results given  in Tables V-17 and V-20  of NAS80.
Normalized results, which give the proportion of fatal  cancer caused by
radiogenic cancer at a particular  site, are listed  in Table 6.2-3.  As
noted  above, these proportions are assumed to be the same for the BEIR-3
linear quadratic dose response model.
                                    6-14

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   Table 6.2-3.  Proportion of the total risk of fatal radiogenic cancer
                 resulting from cancer at a particular site


                                          Proportion of
       Site                                total
Lung                                           0.21
Breast                                       0.11
   (a^NASSO—Lifetime exposure and cancer expression; results are
      rounded to two figures.
   (b)Average for both sexes.
   (°^Leukemia.
   (d>Xotal risk for all other organs, including the esophagus,
      lymphatic system, pharynx, larynx, salivary gland, and brain.
   Information on  the proportion of fatal cancers resulting from cancer
 at a  particular organ is not precise.  One reason is that the data in
 NAS80 (and Table 6.2-3) are based on vhole-body exposures, and it is
 possible  that the  incidence of radiogenic cancer varies depending on
 the number of exposed organs.  Except for breast and thyroid cancer,
 very  little  information is available on  radiogenic  cancer resulting
 from  exposure of only one region in the  body.  Another reason is that
 most  epidemiology  studies use mortality  data from death certificates,
 which often  provide  questionable information on the site of the primary
 cancer.   Moreover, when the existing data are  subdivided into specific
 cancer sites, the  number of cases becomes small, and sampling
 variability  is increased.  The net result of these  factors Is that
 numerical estimates  of the total cancer  risk are more reliable than
 those for most single sites.

    The 1977  UNSCEAR  Committee's estimated risks to  different organs are
 shown in Table 6.2-4.  For all of  the  organs,  except the breast, a high
 and low estimate was made.  This range varies  by a  factor of 2 or more
 for most organs.   Other  site-specific  estimates  show a  similar degree
 of uncertainty,  and  it is clear  that  any system  for allocating  the  risk
 of fatal cancer  on an organ-specific  basis  is  inexact  (Ka82).  Table
 6.2-5 compares proportional risks  by  the HAS  BEIR-3 Committee, UNSCEAR,
 and the ICRP.   ICRP  Report  26 provides organ-specific  weights for
 assessing combined genetic  and cancer risks from occupational exposure
 (ICRP77).  In Table  6.2-5, we have renormalized ICRP risks so that they
 pertain to  cancer  alone.

                                   6-15

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   Table 6.2-4.   UNSCEAR77  estimates  of cancer risks at  specified  sites
Fatalities per
Site person rad
Lung
Breast 'a'
Red bone marrow^"'
Thyroid
Bone
Liver
Stomach
Intestines
Pancreas
Kidneys and urinary
tract
Other 
25-50
25
15-25
5-15
2-5
10-15
10-15
14-23
2-5

2-5
4-10
Average per
organ rad
37.5
25.0
20.0
10.0
3.5
12.5
12.5
18.5
3.5

3.5
7.0
Proportion
of total risk
0.24
0.16
0.13
0.065
0.23
0.081
0.081
0.12
0.023

0.023
0.046
     (a)Average for
     ^Leukemia.
     ^'Includes esophagus and lymphatic tissues.
   Table 6.2-5,
Comparison of proportion of the total risk of radiogenic
      cancer fatalities by body organ
Site
  NAS80
                        (a)
UNSCEAR77
                                            (b)
ICRP77
Lung
Breast
Red Marrow
Thyroid
Bone
Liver
Stomach
Intestine
Pancreas
Kidneys and
urinary tract
Other
.21
.13
.16
.099
.009
.085
.084
.039
.058

.025
.11 d)
.24
.16
.13
.065
.023
.081
.081
.12
.023

.023
.046
.16
.20
.16
.04
.04 ,
(,08)(c>
(.08)
(.08)
(.08)

(.08)
~~
     (^Lifetime exposure  and  cancer  expression.
     (k)Normalized  for  risk  of fatal  cancer  (see text).
     (c)pive additional organs which  have  the highest dose are assigned
        0.08  for a total  of 0.4.
               include  esophagus, lymphatic  system, pharynx, larynx,
        salivary gland, and brain.
                                    6-16

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     Considering that the cancer risk for a particular site is usually
uncertain by a factor of 2 or more, as indicated by the range of UNSCEAR
estimates in Table 6.2-4, we would not expect perfect agreement in
apportionment of total body risks.  Table 6.2-5, however,  does indicate
reasonable agreement among the three sets of estimates considered here.

     The differences between the proportions of the total  risk of fatal
cancer shown in Table 6.2-5 are, for the most part, small  in comparison
to their uncertainty.  We have used the BEIR-3 organ risks in prefer-
ence to those made by other groups such as UNSCEAR or the  ICRP for
several reasons.  BEIR estimates of organ risk are based on a projec-
tion of lifetime risk using age-specific risk coefficients, rather than
just observations to date.  Moreover, the 1980 BEIR Committee consider-
ed cancer Incidence data as well as mortality data.  This  gives added
confidence that the diagnostic basis for their estimates is correct.
And, finally, because we apply these proportional organ risk estimates
to the NAS80 cancer risk estimates for whole-body exposures, we believe
it is consistent to use a single set of related risk estimates.  The way
we have used NAS80 to estimate mortality resulting from cancer at a
particular site is outlined in the next section.

6.2.10  Methodology for Calculating the Proportion of Mortality
        Resulting from Leukemia

     Application of NAS80 to particular problems is straightforward but
requires some familiarity with the details of that report.  In this
section we provide sample calculations based on the BEIR-3 linear dose
response model for the case of fatal leukemia resulting from irradia-
tion of the bone marrow throughout an average person's lifetime.  We
then compared this number to the average number of all fatal radiogenic
cancers to obtain the proportion due to leukemia (Table 6.2-3).

     The NAS80 estimates in Table 6.2-3 differ from the others in that
they include both a consideration of age at exposure and a full
expression of radiogenic cancer resulting from lifetime exposure.  For
example, Table V-17 in NAS80 gives explicit age- and sex-dependent
mortality coefficients for leukemia and bone cancer together.

     The ratio of leukemia to bone cancer fatalities is given by the
coefficient in the dose response relationship listed in Table V-17,
i.e., 2.24/0.05.  For lifetime exposure at a dose rate of one rad per
year, Table V-17 lists 3,568 leukemia (and bone) deaths per 10° males
and 2,709 deaths per 10* females  (HAS80).  Using a male-female birth
ratio of 1.05 to 1.0, this averages to 3,149 fatal cancers per million
persons in the general population.  The total person rads causing these
excess fatalities is the product of one rad per year, 106 persons, and
70.7 years (the average age of this population at death).  Dividing the
total number of fatalities by this product yields 44.5 fatalities per
10° person rad of which about 43.5 are due to leukemia.  As noted
above, for total body exposure, the average of  the absolute and relative
risk projection models yielded 280 premature cancer deaths per 106
person rad.  Therefore, P, the proportion of the whole-body risk caused
                                   6-17

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  j v            ,                 V.WC.1.H u,uc uu j-iietime exoostirp r»f *Vi«a
red bone marrow, is:                                     exposure or tne



                         1 .16  (cf. with Table 6.2-3)          (5^)


To obtain the proportional mortality for other cancers  w» h
site-specific, age-dependent risk coefficients in Table M IA   ^
mortality ratios in Table V-15 to calculate the risk of Hi V
from lifetime exposure at one rad per year (for each sL) ™  C"Cer
as in the example for leukemia outlined above.
    To apply the data shown in Table 6.2-3 to a particular
multiply the average of the relative and absolute SfeSme
estimates for whole body lifetime exposure for a lin*^ T
280 fatalities per 10* person rad and IS fatalities per 10?
rad for a linear quadratic response by the proportional iln
that cancer.  For example, using the linear mod*i       mortality for
(low-LET) to the kidnej (urinar?
 NJ.«™ ~~^,  ,.„  ,.1,^ n.j.uucjr  \ui.xaary  tract.; resulting from Hf
 is estimated  to cause a  lifetime  probability of death caused x!
 radiogenic cancer  that is equal to  (.025) x (280 x lO"*) or 7 *
 10~°, i.e., 7 chances in a million.                           x

    Iodine-131 has been  reported  to be only one tenth as
 X-rays or gamma rays in  inducing  thyroid cancer (NAS72  l
 this cancer a linear dose response and a DREF of 10 ia',,«, A *
 ing lifetime  probability of death.  For example, the risk L™ °alculat;
 dose to the thyroid from exposure to iodine-131 or LHI   T™  & One rad
 lated as follows:  (0.099) x  (0.10) x (280 x 10~6) «r aT  ?nS calcu"
 about 3 chances in a million.                         *'S x 10 b»

 6.2.11  Cancer Risks Due to Age-Dependent DnBon

    As noted  previously  in Chapter 5, almost all of th* H«O«   j ,
 have used are based on the ICRP "Reference Man."  ICRP do«?  ?T   Ve
 are appropriate for adult workers and do not take into accou r    models
 differences resulting from the changes in physiological nar  \
 between children and adults,  e.g., intake rates, metabolism  **
 size.  Although it is difficult to generalize for all radi   a^,or8an
 some cases these differences  tend to counterbalance each oth^r   w   ±n
 example, the  ratio of minute  volume to lung mass is relativ 1
with age, i.e., within a factor of two, so that the ICRP ad It COJsfant
 insoluble materials provides  a reasonably good estimate of ^t
annual dose throughout life.                   estimate of the average

    An exception is the  thyroid where the very young have - ™»i +*  ,
high uptake of radioiodine into a gland which is much smaller S    J7
adult thyroid, as  noted  in Table  5.5-1.  This results in a l«r»
childhood dose and an increased risk which persists throughout iif
Since this is a worst case situation, we have examined it with
                                   6-18

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 care,  using the age-specific risk coefficients  for thyroid cancer in
 Table  V-14 of NAS80 and the age-dependent dose  model in ORNL84.  For
 iodine-131 ingestion,  the estimated lifetime  risk  is increased by a
 factor of 1.56 due to  the 30 percent increase in lifetime dose over that
 obtained with the ORNL adult model, c.f.  Chapter 5.  Results are about
 the same for inhalation of iodine-131— the estimated lifetime risk of
 fatal  thyroid cancer is Increased by a factor of 1.63 for ORNL's
 age-dependent dose estimate.

     As noted in Chapter 5,  use of an age-dependent dosimetry for other
 radionuclides has yielded much smaller Increased doses relative to adult
 models and therefore has little effect on estimates  of lifetime risk.
 In particular, the lung dose  and  risk resulting from the inhalation of
 insoluble alpha particle emitters is nearly unchanged.  The lifetime
 dose for an age-dependent dose model is only  1.09  times greater than
 that calculated using  an adult model (Chapter 5);  the lifetime risk of
 lung cancer for this age-dependent model  Is a factor of 1.16 greater
 than we calculate for  life  exposure with  the  adult only model.

 6.3 Fatal Cancer Risk Resulting  from Hlgh-LET Radiation

     In this section we explain how EPA estimates  the risk of fatal
 cancer resulting  from  exposure to  hlgh-LET radiation.  In some cases,
 ingestion and inhalation of alpha  particle emitting  radionuclides can
 result in a relatively uniform exposure of specific  body organs by
 high-LET radiation.  Unlike exposures to  X-rays and  gamma rays where the
 resultant charged particle flux results in LET's of  the order of 0.2 to
 2 keV  per micron  in tissue, 5 MeV  alpha particles  result in energy
 deposition at a track  average rate  of more than 100 keV per micron.
 High-LET radiation have a larger biological effect per unit dose (rad)
 than low-LET radiation.  How much greater depends  on the particular
 biological endpoint being considered.  For cell killing and other
 readily  observed  endpoints, the relative  biological effectiveness (RBE)
 of  hlgh-LET alpha radiation is often 10 or more times greater than
 low-LET  radiations.

 6.3.1  Quality Factors  for Alpha Particles

     Charged particles  have been assigned quality factors, Q, to account
 for  their  efficiency In producing biological damage.  Unlike an RBE
value, which is for a specific and well-defined endpoint,  a quality
 factor is  based on  an average overall assessment by radiation protection
experts of potential harm of a given radiation relative to X or gamma
radiation.   In 1977, the ICRP assigned a quality factor of 20 to alpha
particle irradiation from radionuclides (ICRP77).   The reasonableness of
 this numerical factor for fatal radiogenic cancers  at a particular site
is not well known, but  it is probably conservative  for all sites and
highly conservative  for some.

     The dose equivalent, in rem,  is the dose, in rad,  times the appro-
priate quality factor for a specified kind of radiation.   For the case
of internally deposited alpha particle emitters the dose equivalent  from


                                  6-19

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a one-rad dose is equal to 20 rem.  It should be noted that prior to
ICRP Report 26 (ICRP77), the quality factor for alpha particle irradia-
tion was 10.  That is, the biological effect from a given dose of alpha
particle radiation was estimated to be 10 times that from an acute dose
of low-LET X-rays or gamma rays of the same magnitude in rad.  The ICRP
decision to increase this quality factor to 20 followed from their deci-
sion to estimate the risk of low-LET radiations, in occupational situa-
tions, on the assumption that biological effects were reduced at low
dose rates for low-LET radiation.  There is general agreement that dose
rate effects do not occur for high-LET (alpha) radiations.  The new ICRP
quality factor for alpha particles of 20 largely compensates for the
fact that their low-LET risks are now based on an assumed dose rate
reduction factor of 2.5.  This DREF has been addressed in preparing EPA
estimates of the risk per rad for alpha particle doses described below
in Section 6.6.3.                                                     '

    In 1980 the ICRP published a task group report "Biological Effects
of Inhaled Radionuclides" which compared the results of animal experi-
ments on radiocarcinogenesis following the inhalation of alpha particle
and beta particle emitters (ICRP80).  The task group concluded that "the
experimental animal data tend to support the decision by the ICRP to
change the recommended quality factor from 10 to 20 for alpha radiation."

6.3.2  Pose Response Function

    In the case of high-LET radiation, a linear dose response is
commonly observed in both human and animal studies and the response is
not reduced at low dose rates (NCRP80).  Some data on human lung cancer
indicate that the carcinogenic response per unit dose of alpha radiation
is higher at low doses than higher ones (Ar8l, Ho8l, Wh83); in addition
some studies with animals show the same response pattern (Ch8l  U182)
We agree with the MAS BEIR-3 Committee that, "For high-LET radiation
such as from internally deposited alpha-emitting radionuclldes  the '
linear hypothesis is less likely to lead to overestimates of the risk
and may, in fact, lead to underestimates" (NAS80).  However, at low
doses, departures from linearity are small compared to the uncertaintv
in the human epidemiological data, and we believe a linear response
provides an adequate model for evaluating risks in the general
environment.

    A possible exception to a linear response is provided by the data
for bone sarcoma (but not sinus carcinoma) among U.S. dial painters who
have ingested alpha-emitting radlum-226 (NAS80).  These data are
consistent with a dose squared response (Ro78).  Consequently, the HAS
BEIR-3 Committee estimated bone cancer risk on the basis of both linear
and quadratic dose response functions.  However, as pointed out in
NAS80, the number of U.S. dial painters at risk who received less than
1000 rad was so small that the absence of excess bone cancer at low
doses is not statistically significant.  Therefore, the consistency of
this data with a quadratic (or threshold) response is not remarkable
and, perhaps, not relevant to evaluating risks at low doses.  In
contrast to the dial painter data, the incidence of bone cancer
                                   6-20-

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 following radlum-224 irradiation, observed in spondylitics by Mays and
 Spless in a larger sample at much lower doses, is consistent with a
 linear response (Ma83, NAS80).  Therefore, for high-LET radiations EPA
 has used a linear response function to evaluate the risk of bone cancer.

     Closely related to the choice of a dose response function is what
 effect the rate at which a dose of high-LET radiation is delivered has
 on its carcinogenic potential.  This is a very active area of current
 research.  There is good empirical evidence, from both human and animal
 studies, that repeated exposures to radlum-224 alpha particles Is five
 times more effective in inducing bone sarcomas than a single exposure
 which delivers the same dose (Ha83,  MAS80).  The 1980 NAS BEIR Committee
 took this into account in their estimates of bone cancer fatalities
 which EPA is using.   We do not know to what extent, if any, a similar
 enhancement of carclnogeniclty may occur for other cancers resulting
 from internally deposited alpha particle emitters.  Nevertheless, we
 believe the ICRP quality factor of 20 Is conservative, even at low dose
 rates.

 6.3.3  Assumptions Made by EPA for Evaluating the Pose from Alpha
        Particle Emitters

     We have evaluated the risk to specific body  organs by applying the
 ICRP quality factor of 20 for  alpha  radiation to the risk estimates for
 low dose  rate low-LET radiation as described above.   For some organs
 this quality factor may be too conservative.   Several authors have noted
 that estimates of  leukemia based on  a  quality factor of 20 for bone
 •arrow Irradiation overpredicts the  observed incidence of leukemia In
 persons receiving  thorotrast (thorium  dioxide) and in the U.S.  radium
 dial painters  (Mo79,  Sp83).  Nevertheless,  In view of the paucity of
 applicable human data and  the  uncertainties  discussed above,  the  ICRP
 quality factor provides a  reasonable and prudent way of evaluating the
 risk due  to alpha  emitters deposited within body organs.

    All of EPA risk estimates for high-LET radiations  are  based on a
 linear dose response function.   For bone cancer and  leukemia we use the
 absolute risk  projection model described in the previous  section.  For
 other cancers  we use the arithmetic average of relative and absolute
 risk projections.

    Table 6.3-1 indicates the Agency's estimates of  the risk of fatal
cancer due to  a uniform organ dose in various organs from Internally
deposited alpha particles.  These estimates are for lifetime doses at a
constant dose  rate.  It was prepared by multiplying the average risk
 (based on the  linear model for a uniformly distributed whole-body dose
of low-LET radiation and a dose rate effectiveness factor of 2.5) by a
quality factor of 20 and then apportioning this risk by organ, as
indicated in Table 6.3-1.
                                   6-21

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   Table 6.3-1.  Estimated number of cancer fatalities from a lifetime
         exposure  to  internally deposited alpha particle emitters
                                   f           Fatalities per
 Site             Proportional  risk*-*'         106 person
Lung
Breast ^c^
Red marrow
-------
  Include  such factors  as  the  expected duration of risk expression and
  variations  in radlosensltlvlty as a function of age and demographic
  characteristics.  A major assumption is the shape and slope of the dose
  effects  response curve,  particularly at low doses where there is little
  or no epideaologlcal  data.   In 1971, the BEIR Committee based its
  estimates of cancer risk on  the assumption that effects at low doses are
  directly proportional to those observed at high doses, the so called
  linear-nonthreshold hypothesis.  As described above in Section 6.2, the
  BEIR-3 Committee considered  three dose response models and indicated a
  preference  for the  linear quadratic model.  The risk coefficients the
  BEIR-3 Committee derived for their linear quadratic model, and to a
  lesser extent their linear model, are subject to considerable
  uncertainty primarily because of two factors:  1) systematic errors in
  the estimated doses of the individual A-bomb survivors,  and
  2) statistical uncertainty because of the small number of cancers
 observed at various dose levels*

 6.4.1  Uncertainty of the Dose Response Models Due to Bias in the
        A-bomb Dosimetry

      Although the BEIR-3 Committee's  choice of a linear  quadratic
 response  has gained considerable  attention, it may not be  generally
 appreciated that the BEIR-3  Committee*s numerical evaluations of  dose
 response  functions for cancer due to low-LET radiation were based
 exclusively on the cancer mortality of  the A-bomb survivors.
 Unfortunately, the dosimetry  for  A-bomb survivors, on which the BEIR-3
 Committee relied,  has  since been  shown  to  have large systematic errors
 which serve to undermine  the  analyses made by  the  Committee.  As
 outlined  below,  the  mathematical  analyses  made by the Committee were
 "constrained" to meet  certain £ priori  assumptions.  These assumptions
 have  since  been shown  to  be doubtful.

     A careful state-of-the-art evaluation  of the dose to  A-bomb
 survivors was carried  out by  investigators  from Oak Ridge  National
 Laboratory  in the early 1960's (Au67, Au77).  The results  of these
 studies resulted in  a. "T65" dose being assigned to the dose (kerma) in
 free air  at  the location  of each survivor for both gamma rays and
 neutrons.  A major conclusion of the ORNL study was that the mix of
 gamma ray and neutron  radiations was quite different in the two cities
 where A-bomblng occurred.  These results indicated that at Hiroshima,
 the neutron  dose was more important than the gamma dose when the
 greater biological efficiency of the high-LET radiations produced by
 neutrons was taken into account.  Conversely, the neutron dose at
 Nagasaki was shown to be negligible compared to the gamma dose for that
 range of doses where there were a significant number of survivors*
 Therefore, the 1980 BEIR Committee evaluated the cancer risks to the
 survivors at Hiroshima on the assumption that the combined effects of
gamma rays and particularly neutrons caused the observed  cancer
response.
                                 6-23

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      Since the BEIR-3 report was published, it has become evident that
 the organ doses due to neutrons at Hiroshima were overestimated by
 about an order of magnitude at distances of 1000 to 1500 meters, where
 most of the irradiated persons survived bomb blast and yet received
 significant doses.  In fact, the neutron doses at Hiroshima are quite
 comparable to those previously assigned, at similar distances,  to
 Nagasaki survivors (Ke81a,b, RERF83,84).  Moreover, there are now
 grounds to believe the T65 estimates of gamma ray doses in both cities
 are also incorrect (RERF83.84).  While several factors need further
 evaluation, reduction of the gamma dose to individual survivors due to
 the local shielding provided by surrounding structures, is signifi-
 cant.  The important point,  however, is that the overestimate of the
 neutron dose to the Hiroshima survivors led to the BEIR-3 Committee
 attributing most of the risk to neutrons rather than gamma-rays.
 Hence,  they underestimated the risk for low-LET radiations by,  as yet,
 an unknown amount.

      For their analysis of the A-bomb survivor data,  the BEIR-3 Commit-
 tee expanded the equations for low-LET radiations listed in Section
 6.2, Table 6.2-1, to include a linear dose response function for
 neutrons:
          1)   P(d,D)  -  cjd + kAD                                 (6-2)

          2)   P(d,D)  -  c2d2  + k2D                               (6-3)

          3)   P(d,D)  -  cad + C4d2 + kaD                          (6-4)
 where  d  is  the  gamma dose and D  is  that part of dose due to high-LET
 radiations  from neutron Interactions.  Note that in equation  (6-4) the
 linear quadratic  (LQ) response,  has two linear terms, one for neutrons
 and one  for gamma radiation.  In analyzing approximately linear data in
 terms  of equation (6-4), the decision as to how much of the observed
 linearity should  be assigned to  the neutron or the gamma component, i.e,
 k3  and 03,  respectively, is crucial.  As shown below, the BEIR-3
 Committee attributed most of the observed radiogenic cancer to a linear
 response from neutron doses which did not occur.

     The BEIR-3 Committee's general plan was to examine the dose re-
 sponse for  leukemia and for solid cancer separately to find statisti-
 cally valid estimates of the coefficients CI-.-.G^ and k^....k3
 by means of regression analyses.  The regressions were made after the
 data were weighted in proportion to their statistical reliability; thus,
 Hiroshima results dominate the analysis.  The T65 neutron and gamma
 doses to individual survivors are highly correlated since both are
 strongly decreasing functions of distance.  This makes accurate deter-
 mination of the coefficients in  equation (6-4) by means of a regression
analysis extremely difficult.  In addition, there is considerable
 sampling variation in the A-bomb survivor data due to small sample size
which exacerbates the regression problem.  Herbert gives a rigorous
discussion of these problems for the case of the A-bomb survivors
 (He83).  Because of these and other problems,  agreement between the
                                   6-24

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observed response for solid cancers and that predicted by any of the
dose response functions examined by the BEIR-3 Committee is not
impressive.  For example, goodness of fit, based on Chi square, ranges
from 0.20 for equation (6-3) to 0.23 for equation (6-4), to 0.30 for
equation (6-2) (Table V-ll in NAS80).  For leukemia, the goodness of fit
between the observed data and that predicted by the regression analysis
is better, e.g., 0.49 for equations (6-2) and (6-3) (Table V8 in HAS80).

     The Committee analyzed the A-bomb survivor data in two separate
sets, i.e., first leukemia and then all cancer excluding leukemia (solid
cancers).  Their treatment of these two cases was not equivalent.
Unlike the analysis of solid cancers, the Committee's analysis of
leukemia considered the Nagasaki and Hiroshima data separately.  Their
approach (p. 342 in NAS80) appears to be based on an unpublished paper
by Charles Land and a published report by Ishlmaru et al. on estimating
the RBE of neutrons by comparing leukemia mortality in Hiroshima to that
in Nagasaki (Is79).  Unlike the case for solid cancers (see below), the
Committee's regression analysis of the leukemia mortality data did
provide stable values for all of the coefficients in equation (6-4), and
therefore an RBE for neutrons as a function of dose, as well as the
ratio of the linear to the dose-squared terms for leukemia induction due
to gamma rays, (03/04).

     Estimating the linear quadratic response coefficients for solid
cancers proved to be less straightforward.  When the BEIR-3 analysis
attempted to fit the A-bomb survivor data on solid cancers to a linear
quadratic dose response function, they found that the linear response
coefficient, 03 in equation (6-4), varied from zero to 5.6 depending
on the dose range considered.  Moreover, their best estimate of the
coefficient for the dose squared term in equation (6-4), i.e., 04, was
zero, i.e., the best fit yielded a linear response.  Therefore, it was
decided that the observations on solid cancers were "not strong enough
to provide stable estimates of low dose, low-LET cancer risk when
analyzed in this fashion," (NAS80, p. 186).

     As outlined in the BEIR-3 Report, the Committee decided to use a
constrained regression analysis, that Is, substitute some of the
parameters for equation (6-4) found in their analysis of leukemia deaths
to the regression analysis of the dose response for solid cancers.  That
is, both the neutron RBE at low dose (the ratio of the coefficient k3
to 03) and the ratio of 03 to 04, as estimated from the leukemia
data, were assumed to apply to the induction of fatal solid cancers.
Regression analyses that are constrained in this manner can yield much
higher estimates of precision than is warranted by the data, as
discussed by Land and Pierce (La83).  They can also be very misleading.
Herbert has discussed this point in detail as it applies to the BEIR-3
regression analysis (He83).  The BEIR-3 Committee's substitution of the
results of the leukemia regression for the data on solid cancers allowed
them to make stable estimates of 03, 04, and k3»  These estimates
became the basis for the "preferred" linear quadratic risk estimates for
solid cancers presented in NAS80, i.e., the LQ-L model, page 187.  (The
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 response models for solid  cancers  that  are  based on  the Conmittee's
 constrained regression analysis are  designated with  a bar in their 1980
 report,  e.g.,  LQ-L and L-L.)

      Given the information discussed above,  it is possible to see, at
 least qualitatively, how the high  bias  in the estimated T65 neutron dose
 to the Japanese survivors  affects  the 1980  BE1R Committee's "preferred"
 LQ estimates of the risk coefficients for leukemia.  The Committee's
 age-adjusted risk  coefficients for leukemia are listed in Table V-8
 (NAS80,  page 184).   For the linear quadratic response, kj, the neutron
 risk coefficient is 27.5.   Tables  A-ll  (HAS80, page  341) and V-6 (NAS80,
 page 152) provide  the  estimates of neutron  and gamma doses to the bone
 marrow of Hiroshima survivors that were used by the  Committee.
 Substituting these  doses in their  risk  equations (Table V-8) indicates
 that about 70  percent  of the leukemia deaths were ascribed to the
 neutron  dose component then thought  to  be present at Hiroshima.  As
 noted above, subsequent research Indicates  that the  high-LET dose due to
 neutrons was actually  much smaller.

      It  is not possible to accurately quantify what  effect the
 Committee's use of  these same coefficients  had on their analysis of the
 dose response  for solid cancers.   Equation  V-10 for  solid cancers, p.
 187 in NAS80,  indicates about 60 percent  of  the solid tumor response was
 attributed to  the T65  neutron dose;  but this is a minimum estimate that
 Ignores  the effect  of  the  assumed  neutron doses on the value of k3 and
 the ratio of 03 to  04.

      The BEIR-3 Committee's LQ-L model  assumes an RBE of 27.8 at low
 doses.   In the Committee's L-L linear response model, the assumed RBE is
 11.3.  Therefore, this linear model  is  considerably  less sensitive to
 the neutron dose component, assumed  by  the  Committee, than their LQ-L
 model.   For either  model, most of  the A-bomb survivors' radiogenic
 cancer was ascribed to the 165 neutron  doses at Hiroshima.

      There is  no simple way of adjusting  the 1980 BEIR risk estimates to
 account  for the risk they attributed to neutrons.  Adjustment of neutron
 doses  alone is  clearly inappropriate, since there is good reason to
 believe  that T65 estimates of the  dose  due to gamma rays are also
 subject  to considerable change.  Moreover, not all of the individuals in
 a given  T65 dose category will, necessarily, remain grouped together
 after new estimates of  neutron and gamma doses are obtained.  Both the
 numerator  and denominator in the ratio  of observed to expected cases are
 subject  to  change and  Indeed could change in opposite directions, a fact
not considered  in some  preliminary analyses (St81).  Nevertheless, it is
 reasonable  to conclude  that bias in  the estimated neutron doses at
 Hiroshima has led to considerable  uncertainty in the BEIR-3 risk
 estimates  and also  to a systematic underestimation of the risk due to
 low-LET  radiations.  For this reason we believe that estimates based on
 the more conservative linear dose  response should be given considerable
weight vis & vis those made using  the BEIR-3 linear quadratic models.
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6.4.2  Sampling Variation

     In addition to the systematic bias in the BEIR-3 risk estimates for
low-LET radiation outlined above,  the precision of the estimated  linear
and quadratic risk coefficients in the BEIR-3 report is poor due  to
statistical fluctuations due to sample size.   Recently, Land and  Pierce
have reevaluated the precision of the BEIR-3 linear quadratic risk
estimates to take into account, at least partially, the Committee's use
of a constrained regression analysis (La83).   This new analysis
indicates that for the BEIR-3 LQ-L model for leukemia, the standard
deviation of the linear term is nearly as large as the risk coefficient
itself (+0.9 compared to a risk coefficient of 1).  For the LQ-L  model,
solid cancer, the standard deviation is +1.5 compared to a risk
coefficient of 1.6.

     It is likely that at least part of the uncertainty attributed to
sampling variation in the BEIR-3 risk estimates is not due to sample
size and other random factors but rather due to the use of incorrect
dose estimates for the A-bomb survivors.  The correlation of neutron and
gamma-ray doses has been a major underlying cause of the uncertainty in
regression analysis using the T-65 doses.  Analyses of revised data with
much smaller neutron doses may result in better precision.  At present,
we have concluded that the BEIR-3 risk coefficients are uncertain by at
least a factor of two, see below, as well as being biased low by an
additional factor of two or more.

6.4.3  Uncertainties Arising from Model Selection

     In addition to a dose response model, a "transportation model" is
needed to apply the risks from an observed irradiated group to another
population having different demographic characteristics.  A typical
example is the application of the Japanese data for A-bomb survivors to
western people.  Seymore Jablon, (Director of the Medical Follow-up
Agency of the National Research Council, HAS) has called this the
transportation problem, a helpful designation because  it is often
confused with the risk projection problem described below.  However,
there is more than a geographic aspect to demographic  characteristics.
The transportation problem includes estimating the risks for one sex
based on data for another and a consideration of habits influencing
health status such as differences between smokers and  nonsmokers.

     The BEIR-3 Committee addressed this problem in their 1980 report
and concluded, based largely on the breast cancer evidence, that the
appropriate way to transport the Japanese risk to the  U.S. population
was to assume that the absolute risk  over a  given observation period was
transferable but that relative risk was not.  Therefore,  the Committee
calculated what the relative risk would be if the  same number of excess
cancer deaths were observed in a U.S. population having  the  same age
characteristics as the A-bomb  survivors.  The base  line  cancer rates in
the U.S. and Japan are quite different for some specific cancers  so this
is a reasonable approach.  However, it contains the assumption that
while the cancer initiation process is the same in the two countries,
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 the actual number of  radiogenic  cancers  that actually occur is the
 result  of  cancer  promotion, the  latter being a culturally dependent
 variable.

     An alternative approach to  solving  the transportation problem is
 that of the 1972  HAS  BEIR-1 Committee.   This Committee assumed relative
 risks would be  the same in the United States and Japan and transferred
 the observed percentage increase directly to the U.S. population.  We
 have compared estimates of the lifetime  risk for these two treatments of
 the transportation problem in order  to find out how sensitive the BEIR-3
 Committee  risk  estimates are to  their assumptions.  To do this, we
 calculated new  relative risk estimates for solid cancers based on the
 age-specific cancer mortality of the Japanese population rather than the
 U.S. data  used  by the BEIR-3 Committee.  We found that this alternative
 approach did not  have much effect on the estimated lifetime risk of
 solid radiogenic  cancer, i.e., a change  of 3 percent for males, and 17
 percent for females.  We have concluded  that the amount of uncertainty
 introduced by transporting cancer risks  observed in Japan to the U.S.
 population is small compared to  other sources of uncertainty in this
 risk assessment.   Base-line leukemia rates are about the same in the
 countries,  so we  believe these risks are also "transportable."

     The last of  the models needed to estimate risk is a risk projection
 model.  As outlined in Section 6.2,  such models are used to project what
 future  risks will be as an exposed population ages.  For leukemia and
 bone cancer, where the expression time is not for a full lifetime but
 rather  25  years,  absolute and relative risk projection models yield the
 same number of  radiogenic cancers, but would distribute them somewhat
 differently by  age.  For solid cancers,  other than bone, the BEIR-3
 Committee  assumed that radiogenic cancers would occur throughout the
 lifetime.   This makes the choice of  projection model more critical,
 because the relative risk projection yields estimated risks about three
 times larger than that obtained with an  absolute risk projection, as
 shown In Table  6.2-2.  Because we have used the average of these two
 projections for solid cancers, we believe this reduces the uncertainty
 from the choice of model to about a  factor of two or perhaps less,
 depending  on the  age distribution of fatal radiogenic cancer, as
 outlined in Section 6.2 above.

     Similarly, there is as yet insufficient information on radiosensitivity
 as a function of  the age at exposure.  The age-dependent risk coefficients
we have used are  those presented in  the BEIR-3 report.  As yet, there is
 little  information on the ultimate effects of exposure during childhood.
As the A-bomb survivors' population ages, more Information will become
available  on the  cancer mortality of persons irradiated when they were
 young.  Table 6.2-2 Indicates that the more conservative BEIR-1 estimates
 for the effect of childhood exposures would Increase BEIR-3 risk estimates
 by about 40  percent.  As this is probably an upper limit, the lack of more
precise Information is not a major source of uncertainty in estimates of
 the risk caused by lifetime exposure.  Similarly, the BEIR-3 Committee did
not calculate population risks for radiogenic cancer that included in utero
 radiation  because they felt the available data were unreliable.  We have
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deferred to their judgment In this regard.  The BEIR-1 report did include
in utero cancer risk.  These had little effect, 1 to 10 percent,  on the
lifetime risk of cancer from lifetime exposure.  An effect this small is
not significant relative to other sources of uncertainty in the risk
assessment.

6.4.4  Summary

     We can only seal-quantitatively estimate the overall uncertainty in
the risk per rad for low-LET radiations.  We expect that more quantitative
estimates of the uncertainty will be possible only after the A-bomb dose
reassessment is completed and the A-bomb survivor data reanalyzed on the
basis of the new dose estimates.  It should be noted, however, that even if
all systematic bias is removed from the new dose estimates, there will
still be considerable random error in the dose estimate for each survivor.
This random error biases the estimated slope of the dose response curve so
that it is smaller than the true dose response (Da72, Ma59).  The amount of
bias introduced depends on the size of the random error in the dose
estimates and their distribution which are unknown quantities at this stage
of the dose reassessment.

     The source of uncertainty in risk estimates for low-LET radiations can
be ranked as shown in Table 6.4-1.
        Table 6.4-1.  A ranking of causes of uncertainty in estimates
                            of the risk of cancer
         Source of uncertainty               Degree of uncertainty


     Choice of dose response model                  +250 percent^a^

     Slope of dose response resulting               +200 percent^
       from sampling variation

     Choice of an average  risk                      +100 percent^)
       projection  model                             ~*

     Choice of transportation model                  +20 percent (<*)

     A-bomb T-65  dosimetry                         Plus only,
                                                    amount unknown


      (*)por choices  limited  to  BEIR-3 linear and linear quadratic
         models, see  6.2.                                  	
      (^Estimate  of  2 standard  deviations for the BEIR-3 LQ model (La83).
      ((^Average of relative  and absolute projection as described above.
             the total of  all cancers,  not specific cancers.
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     The estimates of uncertainty In Table 6.4-1 are not wholly comparable
 and must be interpreted carefully.  However, they do have some Illustrative
 value, particularly when ordered in this way.  The uncertainty listed for
 the slope of dose response is a nominal value for the BEIR-3 linear
 quadratic LQ formulation (La83) in that it is only valid insofar as the
 Committee's assumptions are true.  It is based on a two standard deviation
 error so that the expectation that the error ie less than Indicated is 95
 percent.  We do not believe the uncertainty in the BEIR-3 linear estimate,
 1-L, is significantly smaller, c.f.  Tables V-9 and V-ll in NAS80.

     The other uncertainties listed in Table 6.4-1 are quite different,
 being more in the nature of informed judgments than the result of a
 statistical analysis.   It is doubtful that all radiogenic cancers have the
 same type  of response  functions.   However, if they were all linear,  as
 breast cancer and thyroid appear  to  be,  the BEIR-3 linear quadratic
 response model would underestimate the response by 250 percent.   If most
 cancers have a linear  quadratic response,  or equivalently,  a dose rate
 reduction  factor equal to the difference in slope  at low doses between the
 BEIR-3 linear and linear quadratic models,  the use of a linear model would
 overestimate the response  by a factor of 2.5.   We  believe that a factor of
 250  percent  is a conservative estimate of the uncertainty introduced by the
 lack of data at  low  dose rates.

     As  discussed above,  the  uncertainty due  to the choice of an absolute  or
 a relative risk  model  is about a factor  of  three.   The use  of the average
 risk for these  two models reduces  the  uncertainty  in risk projection by
 more than a  factor of  two, since it is known that  a relative risk
 projection is high for some kinds  of cancer  and  that an absolute risk
 projection is low for others.

     The uncertainties listed  in Table  6.4-1  are  largely independent  of each
 other and therefore  unlikely  to be correlated  in sign.   Their root mean
 square  sum is about  300  percent, indicating  the  expectation that calculated
 risks would  be within a factor of  three  or so  of the  true value.   This
 result  is overly optimistic because it does  not  Include  consideration of
 the  uncertainty  introduced by the  bias in the A-bomb  dosimetry or by the
 constrained  regression analysis used by  the BEIR-3  Committee.

 6.5  Other Radiation-Induced  Health Effects

    The earliest  report of radiation-induced health effects was In 1896,
 and it dealt with acute effects in skin  caused by x-ray exposures  (Mo67).
Within the six-year period following,  170 radiation-related skin damage
 cases had been reported.  Such injury, like many other acute effects, is
 the result of exposure to hundreds or  thousands of rads.  Under normal
 environmental exposure situations, however, such exposure conditions are
not possible and  therefore will not be considered in assessing the risk to
 the general population from radionuclide releases.

    Although radiation-induced carclnogenesis was the first delayed health
effect reported, radiation-induced genetic changes were reported early
too.   In 1927, H.J. Muller reported on x-ray induced mutations in animals
                                  6-30

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and  In 1928, L.J. Stadler reported a similar finding in plants (Ki62).
At about the same time, radiation effects on the developing embryo were
reported.  Case reports in 1929 showed a high rate of microcephaly
(small head size) and central nervous system disturbance and one case of
skeletal defects in children irradiated in utero (UNSCEAR69).   These
effects, at unrecorded but high exposures, appeared to produce central
nervous system and eye defects similar to those reported in rats as
early as 1922 (Ru50).

     For purposes of assessing the risks of environmental exposure from
radionuclide releases, the genetic effects and in utero developmental
effects are the only health hazards other than cancer that are addressed
in this BID.

6.5.1  Types of Genetic Harm and Duration of Expression

     Genetic harm or the genetic effects of radiation exposure are those
effects induced in the germ cells (eggs or sperm) of exposed
Individuals, which are transmitted to and expressed only in their
progeny and future generations.

     Of the possible consequences of radiation exposure, the genetic risk
is more subtle than the somatic risk.  Genetic risk is incurred by
fertile people when radiation damages the nucleus of the cells which
become their eggs or sperm.  The damage, in the form of a mutation or a
chromosome aberration, is transmitted to, and may be expressed in, a
child conceived after the radiation exposure and in subsequent
generations.  However, the damage may be expressed only after many
generations or, alternately, it may never be expressed because of
failure to reproduce.

    EPA treats genetic risk as independent of somatic risk because,
although somatic risk is expressed in the person exposed, genetic risk
is expressed only in progeny and, in general, over many subsequent
generations.  Moreover, the types of damage incurred often differ in
kind from cancer and cancer death.  Historically, research on genetic
effects and development of risk estimates has proceeded Independently of
the research on carcinogenesis.  Neither the dose response models nor
the risk estimates of genetic harm are derived from data on studies of
carcinogenesis.

    Although genetic effects may vary greatly in severity, the genetic
risks considered by the EPA when evaluating the hazard of radiation
exposure include only those "disorders and traits that cause a serious
handicap at sometime during lifetime" (NAS80).  Genetic risk may result
from one of several types of damage that ionizing radiation can cause in
the DMA within eggs and sperm.  The types of damage usually considered
are:  dominant and recessive mutations in autosomal chromosomes, muta-
tions in sex-linked (x-llnked) chromosomes, chromosome aberrations
(physical rearrangement or removal of part of the genetic message on the
                                   6-31

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chromosome or abnormal numbers of chromosomes), and irregularly Inherit-
ed disorders (genetic conditions with complex causes, constitutional and
degenerative diseases, etc.).

    Estimates of the genetic risk per generation are based on a 30-year
reproductive generation.  That is, the median parental age for produc-
tion of children is age 30 (one-half the children are produced by
persons less than age 30, the other half by persons over age 30).  Thus,
the radiation dose accumulated up to age 30 Is used to estimate the
genetic risks.  Using this accumulated dose and the number of live
births In the population along with the estimated genetic risk per unit
dose, it is possible to estimate the total number of genetic effects per
year, those in the first generation and the total across all time.  Most
genetic risk analyses have provided such data.  EPA assessment of risks
of genetic effects includes both first generation estimates and total
genetic burden estimates.

    Direct and Indirect Methods of Obtaining Risk Coefficients for
    Genetic Effects

    Genetic effects, as noted above, may occur in the offspring of the
exposed individuals or they may be spread across all succeeding
generations.  Two methods have been used to estimate the frequency of
mutations in the offspring of exposed persons, direct and indirect.  In
either case, the starting point is data from animal studies, not data
obtained from studies of human populations.

    For a direct estimate, the starting point is the frequency of a
mutation per unit exposure in some experimental animal study.  The 1982
UNSCEAR report gave an example of the direct method for estimating
induction of balanced reciprocal translocatlons (a type of chromosomal
aberration) in males per rad of low level, low-LET radiation (UNSCEAR82).
This method required the following six steps:

                                                    Induction rate/rad
     (1) Rate of Induction in rhesus monkey
         Spermatogonla:  cytogenetic data               0.86 x 10~4

     (2) Rate of induction that relates to
         recoverable translocations In the F^
         (1st filial generation) progeny [divide                   ,
         (1) by 4]                                      0.215 x ID"

     (3) Rate after low dose rate X-rays:
         based on mouse cytogenetic observations
         [divide (2) by 2]                              0.1075 x 10
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     (4) Rate after chronic gamma-irradiation:
         based on mouse cytogenetic observations
         [divide (2) by 10]                             0.022 x 10~4

     (5) Expected rate of unbalanced products:
         [multiply (3) and (4) by 2]     for (3)        0.215 x 10~4
                                         for (4)        0.043 x 10~4

     (6) Expected frequency of congenitally
         malformed children in the Fj., assuming
         that about 6 percent of unbalanced prod-
         ucts [item (5) above] contribute to this
                   for low dose rate X-rays             1.3 x 10~6
                   for chronic gamma radiation          «0.3 x 10~6


     For humans, UNSCEAR estimates that as a consequence of induction
of balanced reciprocal translocations in exposed fathers, an estimated
0.3 to 1.3 congenitally malformed children would occur in each 106
live births for every rad of parental radiation exposure.

     A complete direct estimate of genetic effects would include
estimates, derived in a manner similar to that shown above for each
type of genetic damage.  These direct estimates can be used to
calculate the risk of genetic effects in the first generation (Fi)
children of exposed parents.

     The indirect (or doubling dose) method of estimating genetic risk
also uses animal data but in a different way.  The 1980 BEIR-3 report
demonstrates how such estimates are obtained (NAS80).


     (1) Average radiation-induced mutation per
         gene for both sexes in mice [based on
         12 locus data in male mice]:  Induction
         rate per rad                                   0.25 x 10~7

     (2) Estimated human spontaneous mutation
         rate per gene                                  0>5 x 10-6 to
                                                        0.5 x 10~5

     (3) Relative mutation risk in humans
         [divide (1) by (2)]                            0.005 to 0.05

     (4) Doubling dose:  the exposure needed
         to double the human mutation rate              20 to 200 rad

     The doubling dose can then be used to estimate the equilibrium
genetic effects or the genetic burden in all future generations caused
by the exposure of parents.  Since the genetic component of congenital
defects occurring in the population can be estimated by epidemiological
surveys, and this component is considered to be maintained at an
                                   6-33

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 equilibrium level by mutations,  a  doubling dose of ionizing radiation
 would double these genetic  effects.   Dividing  the number of the various
 genetic effects  in 106  live-births by the doubling dose yields the
 estimate of genetic effects per  rad.   For example:

     (1)  Autosomal dominant and  x-linked            10,000 per 106
          diseases,  current  incidence                live births

     (2)  Estimated doubling dose                   20 to 200 rad

     (3)  Estimate of induced autosomal             SO to 500 per 106
          dominant and x-linked diseases             live births per rad of
                                                    parental exposure

      A doubling  dose estimate assumes that the total population of both
 sexes is equally irradiated, as  occurs from background radiation, and
 that the population exposed is large  enough so that all genetic damage
 can  be expressed In future  offspring.  Although it is basically an
 estimate of the  total genetic burden  across all future generations, it
 can  also provide an estimate of  effects that occur in the first
 generation.   Usually a  fraction  of the total genetic burden for each
 type of damage is assigned  to the first generation using population
 genetics data as a basis to determine the fraction.  For example, the
 BEIR-3 committee geneticists estimated that one-sixth of the total
 genetic burden of x-linked  nutations  would be expressed in the first
 generation,  five-sixths across all future generations.  EPA assessment
 of risks of  genetic effects includes  both first generation estimates
 and  total genetic burden estimates.

 6.5.2   Estimates of Genetic Harm Resulting from Low-LET Radiations

     One of  the  first estimates  of genetic risk was made in 1956 by the
 HAS  Committee on the Biological Effects of Atomic Radiation (BEAR
 Committee).   Based  on Drosophila (fruit fly) data and other
 considerations,  the BEAR Genetics Committee estimated that 10 Roentgens
 (10 R*)  per  generation continued indefinitely would lead to about 5,000
 new instances of  "tangible  Inherited  defects" per 106 births, and
 about  one-tenth  of  them would occur in the first generation after the
 Irradiation began (NAS72).  The UNSCEAR addressed genetic risk in their
 1958, 1962, and 1966 reports (UNSCEAR58,62,66).  During this period,
 they estimated one  rad of low-LET radiation would cause a 1 to 10
 percent increase in the spontaneous incidence of genetic effects.

     In 1972, both the HAS BEIR Committee and UNSCEAR reexamined the
question of genetic risks (MAS72, UNSCEAR72),  Although there were no
definitive human data, additional information was available on the
genetic effects of radiation on mammals and Insects.   In 1977, UNSCEAR
*R is the symbol for Roentgen, a unit of measurement of x-radiation,
equivalent to an absorbed dose in tissue of approximately 0.9 rad.
                                  6-34

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reevaluated the 1972 genetics estimates (UNSCEAR77).  Their new
estimates used recent information on the current incidence of various
genetic conditions, along with additional data on radiation exposure of
mice and marmosets and other considerations.

     In 1980, an ICRP Task Group (ICRPTG) summarized recommendations
that formed the basis for the genetic risk estimates published in ICRP
Report 26 (Of80).  These risk estimates provided in Table 6.5-1, are
based on data similar to that used by the BEIR and UNSCEAR Committees,
but with slightly different assumptions and effect categories.

Table 6.5-1.  ICRP task group estimate of number of cases of serious
              genetic ill health in liveborn from parents Irradiated with
              10' man-rent in a population of constant size^a'
              (Assumed doubling dose - 100 rad)


      Category of                    First
     genetic effect                generation        Equilibrium


 Unbalanced translocations:
 risk of malformed liveborn            23                  30

 Trlsomlcs and XO                      30                  30

 Simple dominants and sex-
 linked mutations                      20                 100

 Dominants of incomplete
 penetrance and multifactorial
 disease maintained by mutation        16                 160

 Multifactorial disease not
  maintained by mutation                0                  0

 Recessive disease                     —                 	

     Total                             89                 320


This is equivalent to effects per 106 liveborn following an average
   parental population exposure of 1 rem per 30-year generation, as used
   by BEIR and UNSCEAR.
Source:  (Of80).

    In 1980 the MAS BEIR-3 Committee revised the genetic risk estimates
(HAS80).  The revision considered much of the same material that was in
BEIR-1, the newer material considered by UNSCEAR in 1977, and some
additional data.  Estimates for the first generation are about a factor of
two smaller than reported in the BEIR-1 report.  For all generations,  the
new estimates are essentially the same as shown in Table 6.5-2.
                                    6-35

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       Table 6.5-2.  BEIR-3 estimates of genetic effects of an average
                     population exposure of 1 rem per 30-year generation


   Type of genetic        Current incidence     Effects per 106 liveborn
       disorder           per 10$ liveborn       per rem per generation

                                                           First generation
 Equilibrium

  Autosomal dominant
  and x-linked                 10,000           5-65            40-200

  Irregularly inherited        90,000      (not  estimated)       20-900

  Recessive                     1,100         Very few        Very  slow
                                                              increase

  Chromosomal aberrations        6,000       Fewer than 10     Increases
                                                              only
                                                              slightly

      Total                   107,100            5-75           60-1100


  Source:   (NAS80).
    The most  recent  genetic risk estimates are given in Table 6.5-3 and
Include some  new data on cells in culture and the results of genetic
experiments using primates rather than rodents (UNSCEAR82).

    Although  all of  the reports described above used somewhat different
sources of information, there is reasonable agreement in the estimates
presented in  Table 6.5-4.  Most of the difference is caused by the newer
information used in  each report.  Note that all estimates listed above
are based on  the extrapolation of animal data to humans.  Groups differ
in their interpretation of how genetic experiments in animals might be
expressed in humans.  While there are no comparable human data at
present, Information on hereditary defects among the children of A-bomb
survivors provides a degree of confidence that the animal data do not
lead to underestimates of the genetic risk following exposure to
humans.  (See "Observations on Human Populations" which follows.)

    It should be noted that the genetic risk estimates summarized in
Table 6.5-4 are for low-LET, low dose, and low dose rate irradiation.
Much of the data were obtained from high dose rate studies, and most
authors have used a sex-averaged factor of 0.3 to correct for the change
from high dose rate, low-LET to low dose rate, low-LET exposure
(NAS72,80, UNSCEAR72.77).   However,  factors of 0.5 to 0.1 have also been
used in estimates of specific types  of genetic damage (UNSCEAR72,77,82).
                                   6-36

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Table 6.5-3.
UNSCEAR 1982 estimated effect of 1 rad per generation of low
dose or low dose rate, low-LET radiation on a population of
10° liveborn according to the doubling dose method
(Assumed doubling dose " 100 rad)
    Disease
 classification
               Current
              incidence
Effect of 1 rad per generation
                                           First generation  Equilibrium
 Autosomal dominant and
  x-llnked diseases          10,000

 Recessive diseases           2,500

 Chromosomal diseases
    Structural                  400
    Numerical                 3,000
 Congenital anomalies,
 anomalies expressed later,
 constitutional and
 degenerative diseases       90,000
                                   15

                                 Slight
                                    2.4
                                 Probably very
                                 small
                     100

                  slow increase
     Total
              105,900
       4.5

      22
 45

149
 Source:  (UNSCEAR82).
6.5.3  Estimates of Genetic Harm for High-LET Radiations

    Although genetic risk estimates are made for low-LET radiation, some
radioactive elements, deposited in the ovary or testis can irradiate the
germ cells with alpha particles.  The ratio of the dose (rad) of low-LET
radiation to the dose of high-LET radiation producing the same endpoint
is called RBE and is a measure of the effectiveness of high-LET compared
to low-LET radiation in causing the same specific endpoint.

    Studies with the beta particle emitting isotopes carbon-14 and
tritium yielded RBE1a of 1.0 and 0.7 to about 2.0, respectively
(UNSCEAR82).  At the present time, the RBE for genetic endpoints due to
beta particles is taken .as one (UNSCEAR77.82).

    Studies of the RBE for alpha-emitting elements in germinal tissue
have used only plutonium-239.  Studies comparing cytogenetic endpoints
after chronic low dose rate gammma radiation exposure, or  Incorporation
of plutonium-239 in the mouse testis, have yielded RBE's of 23 to  50 for
                                   6-37

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  the  type of genetic injury  (reciprocal translocations) that might be
  transmitted to liveborn offspring (NAS80, UNSCEAR77,82).  However, an
  RBE  of four for plutonium-239 compared to chronic low-LET radiation was
  reported for specific locus mutations observed in neonate mice (NAS80).
  Neutron RBE, determined from cytogenetlc studies in mice, also ranges
  from about four to 50 (UNSCEAR82, Gr83a, Ga82).  Most reports use an RBE
  of 20 to convert risk estimates for low dose rate, low-LET radiation to
  risk estimates for high-LET radiation.


  Table 6.5-4.  Summary of genetic risk estimates per 106 liveborn for an
               average population exposure of 1 rad of low dose or low dose
               rate, low-LET radiation in a 30-year generation
                                    Serious hereditary effects
                            First generation          Equilibrium
       Source                                      (all generations)


  BEAR, 1956 (NAS72)              	                     500

  BEIR-I,  1972 (NAS72)          49 (12-200)(b>         300  (60-1500)

  UNSCEAR, 1972 (UNSCEAR72)      9 (6-15)              300

  UNSCEAR, 1977 (UNSCEAR77)     63                        185

  ICRPTG,  1980 (Of80)            89                        320

  BEIR-3,  1980 (NAS80)          19 (5-75)              257(a>  (60-1100)

  UNSCEAR, 1982 (UNSCEAR82)     22                        149


 ta'Geometric  mean is  calculated by taking the square root of the product
  of two  numbers for  which the  mean  is to be calculated.  The cube  root
  of three numbers, etc.   In general, it is the  Nth root  of  the product
  of N numbers for which the mean Is to be  calculated.

  ^'Numbers  in parentheses are the  range of estimates.


6.5.4  Uncertainty in Estimates of Radiogenetic  Harm

    Chromosomal damage and mutations have been demonstrated in cells in
culture,  in plants, in insects, and  in mammals (UNSCEAR72,77,82).
Chromosome studies in peripheral blood lymphocytes of persons exposed to
radiation have shown a dose-related increase in chromosome aberrations
(structural damage to chromosome) (UNSCEAR82).  In a study of nuclear
dockyard workers exposed to external x-radiation at rates of less than
five rads per year, Evans,  et al. found a significant increase in the
Incidence of chromosome aberrations (Ev79).   The Increase appeared to

                                   6-38

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have a linear dependence on cumulative dose.   In a study of people working
and living in a high natural background area  where there was both external
gamma-radiation and internal alpha-radiation, Pohl-Ruling et al.  reported
a complex dose response curve (Po78).   For mainly gamma-radiation exposure
(less than 10 percent alpha radiation), they  reported that the increase in
chromosome aberrations increased linearly from 100 to 200 mrads per  year
then plateaued from 300 mrads to 2 rads per year.  They concluded:

         "From these data, and data in the literature, it can be
         concluded that the initial part of the dose-effect curve
         for chromosome aberrations is not linear or sigmoid with
         a threshold at the lowest dose, but  rises sharply and
         passes into a complex upward form with a kind of plateau
         until it meets the linear curve of the high dose."

     Although chromosomal damage in peripheral blood lymphocytes cannot
be used for predicting genetic risk in progeny of exposed persons,  it
is believed by some to be a direct expression of the damage, analogous
to that induced in germ cells, resulting from the radiation exposure.
It is at least evidence that chromosome damage can occur in vivo in
humans.

     Since there is no quantitative human data on genetic risks
following radiation exposure, risk estimates are based on extrapo-
lations from animal data.  As genetic studies proceeded, emphasis has
shifted from Drosophila to mammalian species in attempts to find an
experimental system which would reasonably project what might happen in
humans.

     For example, Van Buul reported the slope (b) of the linear
regression, Y • a + bD, for induction of reciprocal translocations in
spermatogonia (one of the stages of sperm development) in various
species as follows (Va80):


           Specie                       b x  10* + sd x 104
Rhesus monkey
Mouse
Rabbit
Guinea pig
Marmoset
Human
0.86 + 0.04
1.29 + 0.02
1.A8 + 0.13
0.91 + 0.10
7.44 + 0.95
3.40 + 0.72

to 2.90 + 0.34




 These data indicate  that animal-based  estimates  for this type of genetic
 effect would  be within a factor  of  four of  the true human value.  In this
 case most of  the animal results  would  underestimate the  risk in humans.

     However, when risk estimates such as this are  used  in direct
 estimation of risk for the first generation,  the total uncertainty in the
 estimate becomes indeterminate.  Even  if studies have  been made In a
 species which can predict the dose  response and  risk coefficient for a
 specific radiation-induced genetic  damage,  there is no certainty that it

                                     6-39

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 species, are used to adjust the risk coefficient to what is expected for
 humans.  The uncertainty in these extrapolations has not been quantified.

     A rough estimate of the uncertainty can be obtained by comparing
 direct estimates of risk for the first generation with doubling-dose
 estimates In the 1977 UNSCEAR report (UNSCEAR77).  The estimates differ by
 a factor between two and six with the direct estimate usually smaller than
 the doubling dose estimate.

     A basic assumption in the doubling dose method of estimation Is that
 there is a proportionality between radiation-induced and spontaneous
 mutation rates.  Some of the uncertainty was removed in the 1982 UNSCEAR
 report with the observation that In two test systems (fruit files and
 bacteria), there Is a proportionality between spontaneous and induced
 mutation rates at a number of individual gene sites.  There is still some
 question as to whether the sites that have been examined are representa-
 tive of all sites and all gene loci or not.  The doubling dose estimated
 dose, however, seems better supported than the direct estimate.

     While there is still some uncertainty as to what should be doubled,
 future studies on genetic conditions and diseases can only increase the
 total number of such conditions.  Every report, from the 1972 HAS and
 UNSCEAR reports to the most recent, has listed an increased number of
 conditions and diseases which have a genetic component.

     Observations on Human Populations

     As noted earlier, the genetic risk estimates are based on interpreta-
 tion of animal experiments as applied to data on naturally-occurring
 hereditary diseases and defects in man.  A study of birth cohorts was
 Initiated in the Japanese A-bomb survivors in mid-1946.  This resulted in
 a detailed monograph by Neel and Schull which outlined the background of
 the first study and made a detailed analysis of the findings to January
 1954 when the study terminated (Ne56).  The authors concluded only that It
was improbable that human genes were so sensitive that exposures as low as
 3 R, or even 10 R, would double the mutation rate.  While this first study
addressed morphological endpoints, subsequent studies have addressed other
endpolnts.  The most recent reports on this birth cohort of 70,082 persons
have attempted only to estimate the minimum doubling dose for genetic
effects in humans (Sc81, Sa82).

     Data on four endpolnts have been recorded for this birth cohort.
Frequency of stillbirths, major congenital defects, prenatal death, and
frequency of death prior to age 17 have been examined in the entire
cohort.   Frequency of cytogenetic aberrations (sex chromosome aneuploidy)
and frequency of biochemical variants (a variant enzyme or protein
electrophoresls pattern) have been measured on large subsets of this
cohort.

     Although the updated data reported appear to suggest radiation
effects have occurred, the numbers are small and not statistically
significant.  Overall, the estimated doubling dose for low-LET radiation
                                    6-40

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at high doses and dose rates for human genetic effects  is about 156 rem
and 250 rem (Sc8l, Sa82).   As noted above,  animal studies indicate that
chronic exposures to low-LET radiation would be less hazardous by a factor
of three (NAS72,80).  This would increase the estimated doubling dose  to
468 rem to 750 rem, respectively.  These recent reports suggest the
minimum doubling dose for humans may be four to seven times higher than
those in Table 6.5-4 (based on animal data).  It would  be premature to
reach a firm decision on the exact amount since these reports are based on
the T65 dosimetry in Japan which is being revised.  However, we believe
EPA estimates of genetic risks will prove to be conservative even when the
dosimetry of A-bomb survivors is revised.

     EPA is using the geometric mean of the BEIR-3 range of doubling
doses, about 110 rads.  The minimum doubling dose reported  above  is four
to seven times greater.  It is unlikely that dose estimates for Japanese
survivors will change by this much (RERF83,84).  Therefore, EPA believes
the estimate of doubling of about 100 rads will continue to be a
conservative estimate.

     Ranges of Estimates Provided by Various Models

     EPA has continued to follow the recommendations of the 1980 BEIR-3
committee and uses a linear nonthreshold model for estimating genetic
effects.  Although, as pointed out by the 1982 DNSCEAR committee, there
are a number of models other than linear (Y - c + ad), e.g., linear
quadratic (Y " c + bD + eD2), quadratic  (Y " k + fD2), even power
function (Y - k + gDh).  However, there are strong data to  support the
hypothesis that mutations themselves are single track events.  That is,
the mutations follow a linear dose response function while  the observed
mutation rate shows the influence of other factors,  and may be nonlinear
(UNSCEAR82).  Y is yield of genetic effects; D is radiation dose; c,  C, k,
and K. are spontaneous incidence constants for genetic effects; and a, b,
e, f, g, and h are  the rate constants for radiation  induced genetic
effects.

     Most of the  arguments  for  a nonlinear dose  response have been based
on target theory  (Le62) or microdosimetric  site  theory  (Ke72).  However,
other theories based  on biology [e.g.,  enzyme  induction-saturation
(6o80,82), repair-misrepalr (To80)] could also provide models that fit the
observed data.  There is  still  much disagreement on which  dose response
model is appropriate  for  estimating genetic effects in humans.  Until
there is more consensus,  the linear nonthreshold model appears to be  a
prudent approach that will not  grossly underestimate the risks.

      The agreement in estimates made  on a linear nonthreshold model in
various reports  is reasonably good.   Even though the authors of  the
reports used different animal models,  interpreted them in  different ways,
and  had different estimates of  the level of human genetic  conditions  in
the  population,  the range of risk coefficients is about an order of
magnitude  (see  Table 6.5-4).  For the most recent, more comparable
estimates,  the  range is a factor of two to four (see ICRPTG, BEIR-3 and
UNSCEAR82  in Table 6.5-4).
                                    6-41

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 6.5.5  The EPA Genetic Risk Estimate

      There is no compelling evidence for preferring any one set of the
 genetic risk estimates listed in Table 6.5-4.  EPA has used the estimates
 from BEIR-3 (NAS80).  These "indirect" estimates are calculated using the
 normal prevalence of genetic defects and the dose that Is considered to
 double this risk.  The HAS estimates which EPA uses are based on a
 "doubling dose" range with a lower bound of 50 reins and an upper bound of
 250 rems.  We prefer these risk estimates to those made by the ICRP task
 group, which used a "direct" estimate because the ICRPTG tabulation
 combines "direct" estimates for some types of genetic damage with doubling
 dose estimates for others (Of80).   We also prefer the BEIR-3 risk
 estimates to the "direct" estimates of UNSCEAR82 which tabulates genetic
 risk separately by the direct method and by the doubling dose method.   The
 risk estimated by the direct method does not Include the same types of
 damage estimated by doubling doses and was not considered further.
 Moreover, the BEIR-3 genetic risk  estimates provide a better estimate of
 uncertainty than the UNSCEAR82 and ICRPTG estimates because the BEIR-3
 Committee assigned a range of uncertainty for multifactorlal diseases
 (>5 percent to <50 percent) which  reflects the uncertainty In the
 numbers better than the other estimates do (5 percent and 10 percent,
 respectively).

      In developing the average mutation rate for the two sexes used in the
 calculation of the relative mutation risk,  the BEIR-3 Committee postulated
 that the Induced mutation rate In  females was about 40 percent of that in
 males (NAS80).  Recent studies by  Dobson et al.  suggest that the
 assumption was Invalid and that human oocytes should have a risk
 equivalent to that of human spermatogonla.   This would Increase the risk
 estimate obtained from doubling dose methods by a factor of 1.43 (Do83a,
 Do83b,  Do84af  Do84b).

      We recognize,  however,  that the use of the doubling dose concept  does
 assume  that radiation-Induced genetic damage is  in some way proportional
 to  "spontaneous" damage.   As noted earlier,  the  recent evidence obtained
 in  insects (Drosophila) and  bacteria (E.  coll)  supports the hypothesis
 that, with the exception  of  "hot spots"  for mutation,  the radiation-
 induced mutation rate  is  proportional to the spontaneous  rate (UNSCEAR82).
 No  proof  that  this  is  also true in mammals  is available yet.

     The  BEIR-3  estimates give a considerable range.   To express the range
 as  a single estimate,  the geometric mean of  the  range is  used,  a method
 first recommended by UNSCEAR for purposes of calculating  genetic risk
 (UNSCEAR58).   The factor  of  three  increase  In risk  for high dose rate,
 low-LET radiation noted earlier is  also used.

     The  question of RBE  for high-LET radiation  is more  difficult.  As
noted above, estimated RBE'a for plutonium-239 alphas  versus  chronic gamma
 radiation for  reciprocal  translocations as determined  by  cytogenetic
analyses  are between 23 and  50  (NAS80, UNSCEAR82).  However,  the  observed
RBE for single locus mutations  in developing  offspring of male mice  given
plutonlua-239  compared to those  given X-ray  irradiation is  four  (NAS80).
                                    6-42

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The average of RBE'a for reciprocal translocations and for specific  locus
mutations Is 20.25.  Since reported neutron RBE's are similar to those
listed above for plutonium-239 alpha radiation,  we use an RBE of 20  to
estimate genetic risks for all high-LET radiations.   This Is  consistent
with the RBE for high-LET particles recommended  for estimated genetic
risks associated with space flight (Gr83b).

     Genetic risk estimates used by EPA for high- and low-LET radiations
are listed in Table 6.5-5.  As noted above, EPA  uses the dose received
before age 30 in assessing genetic risks.


   Table 6.5-5.  Estimated frequency of genetic  disorders in  a birth cohort
                 due to exposure of the parents  to 1 rad per  generation
     Radiation
                                      Serious Heritable Disorders
                                       (Cases per 106 liveborn)
                             First Generation.            All Generati
Low Dose Rate, Low-LET         20          30              260       370

High Dose Rate, Low-LET        60          90              780      1110

High-LET                      400         600             5200      7400
          sensitivity to induction of genetic effects is 40 percent as
   great as that of males.

-------
 our knowledge of medicine improves, that recessive hereditary defects will
 be carried on for many more generations than assumed by the BEIR Committee.

      The relative risk of high-LET radiation compared to low dose rate,
 low-LET radiation (ELBE) is also uncertain*  The data are sparse, and
 different studies often used different endpoints.  In addition, the
 microscopic dosimetry, I.e., the actual absorbed dose in the cells at
 risk, is poorly known.  However, the RfiE estimate used by EPA should be
 within a factor of five of the true RBE for hlgh-LET radiation.

 6.5.6  Teratogenic Effects

      Although human teratogenesis (congenital abnormalities or defects)
 associated with x-ray exposure has a long history, the early literature
 deals mostly with case reports.  Stettner reported a case In 1921 and
 Murphy and Goldstein studied a series of pregnancies in which 18 of the
 children born to 76 irradiated mothers were microcephallc (St21, Mu29,
 Go29).  However, the irradiation exposures were high.

      In 1930, Murphy exposed some rats to X-rays at doses of 200 R to
 1600 R.  Thirty-four of 120 exposed females had litters, and five of the
 litters had animals with developmental defects (Mu30).   He felt that this
 study confirmed his clinical observations and earlier reports of animal
 studies.  Although there were additional studies of radiation-induced
 mammalian teratogenesis before 1950,  the majority of the studies were done
 after that time (see Ru53 for a review), perhaps reflecting radiation
 hazards caused by the explosion of nuclear weapons in 1945 (Ja70).

      Much of  the work done after World War II was done  In mice and rats
 (Ru50,54,56,  W154,  H154).   Early studies,  at relatively high radiation
 exposures,  25 R and above,  established some dose response relationships.
 More  importantly,  they established the time table of sensitivity of the
 developing  rodent embryo  and fetus to  radiation effects  (Ru54,  H153,  Se69,
 H166).

     Rugh,  in his review  of  radiation  teratogenesis  listed  the reported
 mammalian anomalies and the  exposure causing them (Ru70).   The lowest
 reported exposure was 12.5 R for  structural defects  and  1 R for functional
 defects.  He  also  suggested  human exposure  between ovulation and about  7
 weeks gestations! age could  lead  to structural  defects and  from about 6
 weeks gestational age until  birth could lead to functional  defects.   In a
 later review, he  suggested structural  defects In the skeleton might be
 induced as late as the 10th  week  of gestation and functional defects  as
 early as the  4th week (Ru71).  It  should be noted that the  gestation
 period in nice is much shorter than that in humans and that weeks of
 gestation referred to above  are in terms of equivalent stages of
mouse-human development.  Estimates of  equivalent gestational age are not
very accurate.

     In the reports of animal studies  it appeared  as if teratologic
effects, other than perhaps growth retardation, had a threshold for
Induction of effects  (Ru54,53, W154).  However, Ohzu showed that doses as


                                   6-44

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low as 5 R to preimplantatlon mouse embryos caused increased resorption of
implanted embryos and structural abnormalities in survivors  (Oh65).  Then
in 1970, Jacobsen reported a study in which mice were exposed to 5,  20  or
100 R on the 8th day of pregnancy (Ja70).   He concluded that the dose
response function for induction of skeletal effects was linear,  or nearly
linear, with no observable threshold.  This appears consistent with  a
report by Russell, which suggested a threshold for some effects  whereas
others appeared linear (Ru57).

     Rugh suggested there may be no threshold for radiation-induced
congenital effects in the early human fetus (Ru71).  In the  case of
microcephaly and mental retardation, at least this may be the case.  For
other teratogenlc effects, the dose response in humans is unknown.  In
1978, Michel and Fritz-Niggli reported induction of a significant increase
in growth retardation, eye and nervous system abnormalities, and post
implantation losses in mice exposed to 1 R (Mi78).  The increase was still
greater if there was concurrent exposure to radiosensiticing chemicals
such as iodoacetimide or tetracycline (M178).

     One of the problems with the teratologic studies in animals is the
difficulty of determining how dose response data should be interpreted.

     Russell pointed out some aspects of the problem: 1) although
radiation is absorbed throughout the embryo, it causes selective damage
which is consistently dependent on the stage of embryonic development at
the time of irradiation, and 2) the damaged parts  respond, in a consistent
manner, within a narrow time range (Ru54).  However, while low dose
irradiation at a certain stage of development produces changes only in
components at their peak sensitivity, higher doses may Induce additional
abnormalities which have peak sensitivity  at other stages of development,
and may further modify expression  of  the changes  induced in  parts of the
embryo  at peak sensitivity during  the time of irradiation.   In  the  first
case, damage may be  to primordial  cells themselves, while in the  second,
the damage may lead  indirectly  to  the same or different endpoints.

      The embryo/fetus  starts as a  single fertilized egg and divides and
differentiates to  produce  the normal Infant at  term.  (The  embryonic
period, when  organs  develop,  is the  period from conception  to 7 weeks
gestational age.   The  fetal period,  a time of in utero growth,  is the
period from 8 weeks  gestational age to  birth.)   The different organ and
tissue primordia develop independently  and at different rates.   However,
they  are in contact through chemical induction or evaporation (Ar54).
These chemical messages between cells are  Important in bringing about
orderly development and the correct timing and fitting together of  parts
of organs  or organisms.   While radiation can disrupt this pattern,
interpretation of the response may be difficult.  Since the cells in the
 embryo/fetus differentiate, divide, and proliferate at different times
during gestation and at different rates,  gestational times  when cells of
 specific organs or tissues reach maximum sensitivity to radiation are
 different.   Each embryo/fetus has a different timetable.  In fact, each
 half (left/right) of an embryo/fetus may have a slightly different
 timetable.
                                     6-45

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       In addition,  there is a continuum of variation from the hypotheti-
  cal normal to the  extreme  deviant, which is  obviously  recognizable.
  There is no logical place  to draw a  line of  separation between normal
  and abnormal.   The distinction between minor variations  of normal and
  frank malformation,  therefore,  is an arbitrary one,  and  each investiga-
  tor must establish his  own criteria  and apply them  to  spontaneous and
  induced abnormalities alike (HWC73).   For example,  some  classify mental
  retardation based  on IQ (80 or  lower),  some  classify based on ability to
  converse or hold a job,  some on the  basis of the need  to be institution-
  alized.

       Because of the  problems  in Interpretation listed above, it appears
  a pragmatic approach is  useful.  The dose  response should be given as
  the  simplest function that  fits the data,  often linear or linear with a
  threshold.   No  attempt should be made to develop complex dose response
  models unless the  evidence  is unequivocal.

       The first  report of congenital abnormalities in children exposed
  in utero to  radiation from atomic bombs was that of Plummer (P152).
  Twelve children with microcephaly of which 10 also had mental
  retardation  had been identified in Hiroshima in the jLn utero exposed
  survivors.   They were found as part of a program started in 1950 to
  study children exposed in the first trimester of gestation.   In 1955 the
 program was expanded to include all survivors exposed in utero.

      Studies Initiated during the program have shown the  following
 radiation-related effects:   1) growth retardation;  2) increased
 microcephaly; 3) increased  mortality, especially infant mortality; 4)
 temporary suppression of antibody production against influenza;  and  5)
 Increased frequency of chromosomal aberrations  in peripheral  lymphocytes
 (Ka73).

      Although there have been a number  of studies of Japanese A-bomb
 survivors,  including one showing a dose and gestational age related
 Increase in postnatal mortality (Ka73),  only  incidence  of microcephaly
 and  mental retardation have been investigated to  any great extent.   In
 the  most recent  report,  Otake and  Schull showed that mental retardation
 was  associated with exposure between  8  and 15 weeks  of  gestation (10 to
 17 weeks  of gestation if counted from the last menstrual  period)
 (Ot83).   They further found a linear  dose response relationship for
 induction of mental retardation  that  had a slope  yielding a doubling
 dose  for  mental  retardation of about  2  rads,  fetal absorbed dose
 (Ot83).   Classification as mentally retarded  was  based  on "unable to
 perform simple calculations,  to  care  for himself  or herself, or if he or
 she was completely  unmanageable  or had been Institutionalized" (Ot83).

     Estimates of the risk of mental retardation for  a rad of
embryo/fetus  exposure in  the U.S. population  can be derived by three
methods.  The first and easiest method is to use the absolute risk
calculated by Otake and Schull for the Japanese survivors (Ot84).  A
second method is to use the doubling dose calculated by Otake and Schull
times the incidence of mental retardation per 103 live births (Ot83).
                                   6-46

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Unfortunately, a number of assumptions oust be made to establish the
Incidence of mental retardation per 10^ live births.  Mental
retardation nay be classified as «ild (IQ 50-70),  moderate (IQ 35-49),
severe (IQ 20-34) and profound (IQ <20) (WH075).  However, some
investigators use only mild mental retardation (IQ 50-70) and severe
mental retardation (IQ <50) as classes (Ha8l, St84).  Mental
retardation is not usually diagnosed at birth but at some later time,
often at school age.  Since the mental retardation may have been caused
before or during gestation, at the time of birth,  or at some time after
birth, that fraction caused before or during gestation must be
estimated.  In like manner since mental retardation caused before birth
may be due to genetic conditions, infections, physiologic conditions,
etc., the fraction related to unknown causes during gestation must be
estimated.  This is the fraction that might possibly be doubled by
radiation exposure.

     A third method to estimate the risk is indirectly using the
relationship of microcephaly and mental retardation reported in the
Japanese survivors (Wo65, Ot83).  If head size is assumed to be normally
distributed, then the fraction of the population with a head size 2 or 3
standard deviations smaller than average can be obtained from
statistical tables.  The fraction of 103 llveborn with microcephaly
multiplied by the proportion of mental retardation associated with that
head size yields an estimate of the incidence of mental retardation per
10J live births, which can then be used with the doubling dose to
estimate the risk as described above.

     Risk estimates for mental retardation are derived below for
comparison purposes using each of the three methods described above.

    Estimate of Incidence Per Rad Based on Direct Application of the
    Slope of the Japanese Data

     Otake and Schull gave an estimate of "The Relationship of Mental
Retardation to Absorbed Fetal Exposure in the  'Sensitive* Period When
All 'Controls* are Combined" (Ot84).  The estimate of 0.416 cases of
mental retardation per 100 rad could be directly applicable to a U.S.
population.  In this case the risk estimate would be about 4 cases of
mental retardation per rad per 1000 live births.

    Estimate of Incidence Per Rad Based on the Doubling Dose

     The Otake and Schull report suggested the doubling dose for mental
retardation was about 2 rads fetal absorbed dose or about a 50 percent
increase in mental retardation per rad (Ot83).  It would seem reasonable
that this doubling dose would apply only to ideopathic cases of mental
retardation caused during gestation, that is,  those which have no known
genetic, viral, bacterial, etc., cause.

     Data from studies of the prevalence of mental retardation in  school
age populations in developed countries suggest a prevalence  of
2.8 cases/1000 (Uppsala County, Sweden) to 7.4 cases/1000 (Amsterdam,


                                   6-47

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 Holland) of severe mental retardation, with a mean of about 4.3 + 1.3
 cases/1000 (St84).  Where data are available for males and females
 separately, the male rate is about 30 percent higher than the female
 rate (St84).  Historically, the prevalence of mild mental retardation
 has been 6 to 10 times greater than that of severe mental retardation.
 But in recent Swedish studies, the rates of prevalence of mild and
 severe mental retardation have been similar (St84).  This was suggested
 to be due to a decline in the "cultural-familial syndrome".  That is,
 improved nutrition, decline in Infection and diseases of childhood,
 increased social and Intellectual stimulation, etc., combined to reduce
 the proportion of nonorganlc mental retardation and, therefore, the
 prevalence of mild mental retardation (St84).

      In studies of the causes of mental retardation, 23 to 42 percent of
 the mental retardation has no identified cause (Gu77, Ha8l, StS4).  It
 ±8 this portion of the mental retardation which may be susceptible to
 Increase from radiation exposure of the embryo/fetus.  In that case, the
 prevalence of ideopathic mental retardation would be 0.6 to 3.1 cases
 per 1000 of severe mental retardation and perhaps an equal number of
 cases of mild mental retardation.

      For purposes of estimating the effects of radiation exposure of the
 embryo/fetus,  a risk of spontaneous ideopathic mental retardation of 1
 to 6 per 1000 will be used.   If this spontaneous Ideopathic mental
 retardation can be Increased by radiation the estimate would be:

         (1 to 6 cases per 1000 live births)(0.5 Increase per rad)

 or about 0.5  to 3 cases  of mental  retardation per rad per 1000  live
 births.

      This  estimate may be biased low because mental  retardation induced
 during gestation is often associated vith high childhood  death  rate
 (St84).   If this is generally  true for  ideopathic  causes  of mental
 retardation,  it would  cause  an underestimation of  the risk.

    Estimate  of  Incidence Per  Rad  Based on Incidence  of Microcephaly

     1)  Of live  born children, 2.275 percent will have a head  circum-
 ference  2 standard  deviations  or more smaller than average, 0.621 per-
 cent will have a head circumference  2.5 standard deviations or more
 smaller  than average, and 0.135 percent will have  a head  circumference  3
 standard deviations or more  smaller  than average (statistical estimate
 based on a normal distribution).

     2)  There is evidence in a nonselected group  of  9,379 children  that
mental retardation can be estimated using incidence of microcephaly,
even though head circumference in  the absence of other supporting data,
e.g., height or proportion,  is an uncertain indicator of mental retarda-
tion.  Based on a study of 9,379 children, Nelson and Deutschberger  con-
cluded that about half of the children with a head circumference 2.5
standard deviations or more  smaller than average had IQ's of 79 or lover
                                   6-48

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(Ne70).  Since 0.67 percent of those studied were in this  group,  the
observed number is about what would be expected based on the normal
distribution of head size in a population,  0.62 percent.   The  estimated
incidence of mental retardation per live birth in a population would
be:


       6.7 cases of microcephaly    0.5 cases of mental retardation
           1000 live births             case of microcephaly

or about 3.4 cases of mental retardation per 1000 live births.

     3)  A first approximation of risk of mental retardation might then  be:

               3.4 cases of mental retardation   0.5 increase
                      1000 live births         *      rad

or about 2 cases of mental retardation per 1000 live births per rad.

     Both microcephaly and mental retardation were increased in Japanese
survivors (Wo65,66).  About half of those with head sizes  2 or more
standard deviations smaller than average had mental retardation (RERF78),
a result similar to that observed by Nelson and Deutschberger  (Ne70).
Therefore, the above estimate based on the incidence of microcephaly in  a
population should be a reasonable estimate of the risk from radiation.

     Summary of the Calculated Risk of Mental Retardation

     The risk of increased mental retardation per rad of embryo/fetus
exposure during the 8- to 15-week gestational period estimated above
ranges from about 5 x 10~* to 4 x 10"3 cases per live birth,  the
larger being a direct estimate.  The geometric mean of these estimates is
1.4 x 10"J; the arithmetic mean is 2.4 x 10~3 cases per live  birth.

     All the estimates derived above by any of the three methods are in
the same range as an earlier UNSCEAR estimate of an increase of 1 x 10~3
cases of mental retardation per rad per live birth (UNSCEAR77).  The
UNSCEAR estimate, however, did not consider gestational age at the time of
exposure.  The Otake and Schull report did address gestational age and
estimated a higher risk, but a narrower window of susceptibility (Ot83).

     If the estimates are applicable, the 15 mrads of low-LET background
radiation delivered during the 8- to 15-week gestational age-sensitive
period could induce a risk of 6 x 10~5 to 7.5 x 10"6 cases of mental
retardation per live birth.  This can be compared to an estimate of a
spontaneous occurrence of 1.5 x 10"2 to 3.4 x 10~3 cases of mental
retardation per live birth.

     Japanese A-bomb survivors exposed in utero also showed a number of
structural abnormalities and, particularly in those who were  microcepha-
lic, retarded growth (Wo65).  No estimate has been made of the radiation-
related Incidence or dose response relationships  for these abnormalities.
                                    6-49

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 However, UNSCEAR made a very tentative estimate based on animal studies
 that the Increased incidence of structural abnormalities In animals may
 be 5 x 10~3 cases per R per live birth, but stated that projection to
 humans was unwarranted (UNSCEAR77).  In any event, the available human
 data cannot show whether the risk estimates derived from high dose
 animal data overestimates the risk In humans.

     It should be noted that all of the above estimates are based on high
 dose rate low-LET exposure.  UNSCEAR in 1977 also investigated the dose
 rate question and stated:

          "In conclusion,  the majority of the data available for
          most species indicate a decrease of the cellular and
          malformature effects by lowering the dose rate or by
          fractionating the dose.  However, deviations from this
          trend have been  well documented in a few instances and
          are not inconsistent with the knowledge about
          mechanisms of the teratogenic effects.   It is therefore
          Impossible to assume that dose rate and fractionation
          factors have the same influence on all  teratological
          effects."  (UNSCEAR77).

      From this analysis,  EPA has concluded that  a range of risk is
 4  x 10~3 to 5 x 10~* cases of mental  retardation per live birth per
 rad of low-LET radiation  delivered between weeks 8 and 15 of gestation
 with no threshold identified at  this  time.

      No attempt can be made now  to estimate total teratogenic effects.
 However, it should  be noted that the  1977  UNSCEAR estimate from animals
 was 5 x 10~3 cases  of structural abnormalities per R per live birth
 (about the same number per rad of low-LET).  This estimate must be viewed
 as a minimum one since it  Is based, to  a large extent,  on observation of
 grossly visible malformations.   Differences  in criteria for identifying
 malformations have  compounded the problem, and questions of threshold and
 species differences have made risk projection to humans unwarranted.

 6.5.7  Nonstochastic  Effects

      Nonstochastic  effects,  those effects  that increase in severity with
 increasing  dose  and may have a threshold, have been reviewed  in the 1982
 UNSCEAR report  (UNSCEAR82).   In  general, acute doses  of 10 rads of
 low-LET radiation and higher are  required  to induce these  effects.  It  is
 possible that some of the observed effects of in utero  exposure are
 nonstochastlc, e.g.,  the risk of  embryonic loss, estimated  to  be 10~2
 per R (UNSCEAR77), following radiation exposure  soon after  fertilization.
 However, there are no data to address the question.  Usually, no
nonstochastic effects of radiation are expected at  environmental levels
of radiation exposure.
                                   6-50

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6.6  Radiation Risk - A Perspective

     To provide a perspective on the risk of fatal radiogenic cancers and
the hereditary damage due to radiation,  we have calculated  the risk from
background radiation to the U.S. population using the risk  coefficients
presented in this chapter and the computer codes described  in Appendix A.
The risk resulting from background radiation is a useful perspective for
the risks caused by releases of radionuclides.   Unlike cigarette  smoking,
auto accidents, and other measures of common risks, the risks resulting
from background radiation are neither voluntary nor the result of alcohol
abuse.  The risk caused by background radiation is very largely
unavoidable; therefore, it is a good benchmark for judging  the estimated
risks from radionuclide releases.  Moreover, to the degree  that  the
estimated risk of radionuclides is biased, the same bias is present in
the risk estimates for background radiation.

     Low-LET background radiation has three major components:  cosmic
radiation, which averages to about 28 mrads per year in the U.S.;
terrestrial sources, such as radium in soil, which contributes an average
of 26 mrads per year (NCRP75); and the low-LET dose resulting from
internal emitters.  The last differs between organs, to some extent,  but
for soft tissues is about 24 mrads per year (NCRP75).  Fallout  from
nuclear weapons tests, naturally occurring radioactive materials in
buildings, etc., contribute about another 10 mrems for a total  low-LET
whole-body dose of about 90 mrads per year.  The lung and bone  receive
somewhat larger doses due to high-LET radiations; see below. Although
extremes do occur, the distribution of this background annual doae to the
U.S. population is relatively narrow.  A population weighted analysis
indicates that 80 percent of the U.S. population would receive annual
doses that are between 75 mrads per year and 115 mrads per year  (EPA81).

     As outlined in Section 6.2, the BEIR-3 linear models yield, for
lifetime exposure to low-LET radiation, an average lifetime risk of fatal
radiogenic cancer of 280 per 10e person rad.  Note that  this average is
for a group having the age- and sex-specific mortality rates of  the 1970
U.S. population.  We can use this datum to calculate  the average lifetime
risk due to low-LET background radiation as follows.  The average
duration of exposure in this group is 70.7 years  and  at  9 x 10~" rad
per year, the average lifetime dose  is 6.36 rads.  The risk of fatal
cancer per person in this group is:

                  280 fatalities x 6.36 rads - 1.78 x 10~3
                 106 person rad

or about 0.18 percent of all deaths.  The vital statistics we use in our
radiation risk analyses indicate  that the probability of dying  from
cancer in the United States from  all causes is about 0.16,  i.e.,
16 percent.  Thus, the 0.18 percent  result for the BEIR-3  linear dose
response model indicates  that  about  1 percent  of all U.S.  cancer is due
to low-LET background radiation.   The BEIR-3 linear quadratic model
indicates that about 0.07 percent of all deaths are due to low-LET
background radiation or about  0.4 percent of all cancer deaths.


                                   6-51

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    Table 6.3-1 indicates a risk of 460 fatalities per 106 organ rad for
alpha emitters In lung tissue.  The lifetime cancer from this exposure is:
              460 fatalities x Ml^ad x ?0>7 years . ^ x lQ-3

              10  organ rad     year

This is twice the risk due to low-LET background radiation calculated by
means of the BEIR-3 linear quadratic model and more than half of the risk
calculated by means of the BEIR-3 linear model.

     The 1982 UNSCEAR report indicates that the average annual dose to the
endosteal surfaces of bone due to naturally-occurring hlgh-LET alpha
radiation is about 6 mrads per year or, for a quality factor 20, 120 mrems
per year (UNSCEAR82).  Table 6.3-1 indicates that the lifetime risk of
fatal bone cancer due to this portion of the naturally occurring radiation
background is:


               20 cases
           10  person rad     year

     The spontaneous incidence of serious congenital and genetic
abnormalities has been estimated to be about 105,000 per 10° live
births, about 10.5 percent of live births (NAS80, UNSCEAR82).  The low-LET
background radiation dose of about 90 mrads/year in soft tissue results in
a genetically significant dose of 2.7 rads during the 30-year reproductive
generation.  Since this dose would have occurred in a large number of
generations, the genetic effects of the radiation exposure are thought to
be an equilibrium level of expression.  Since genetic risk estimates vary
by a factor of 20 or more, EPA uses a log mean of this range to obtain an
average value for estimating genetic risk.  Based on this average value,
the background radiation causes 700 to 1000 genetic effects per 10° live
births, depending on whether or not the oocyte is as sensitive to
radiation as the spermatogonia.  This result indicates that about 0.67
percent to 0.95 percent of the current spontaneous incidence of serious
congenital and genetic abnormalities may be due to the low-LET background
radiation.
                                  6-52

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                                REFERENCES
Ar54       Arey L. B., Developmental Anatomy,  6th ed., W.B.  Saunders,
           Philadelphia, 1954.

Ar8l       Archer V. E., Health Concerns In Uranium Mining and Milling,
           J. Occup. Med. 23, 502-505, 1981.

Au67       Auxier J. A., Cheka J. S., Haywood F. F., Jones T.  D.,  and
           Thorngate J. H., Free-Field Radiation Dose Distributions from
           the Hiroshima and Nagasaki Bombings, Health Phys. 12, 425-429,
           1967.

Au77       Auxier J.A., Ichlban - Radiation Doslmetry for the Survivors
           of the Bombings of Hiroshima and Nagasaki, T1D 27080,
           Technical Information Center, Energy Research and Development
           Administration, National Technical Information Service,
           Springfield, Virginia, 1977.

Ba81       Baverstock K. F., Papworth D., and Vennart J., Risks of
           Radiation at Low Dose Rates, Lancet, 430-433, Feb. 21, 1981.

Be78       Beebe 6. W., Kato H., and Land C. E., Studies of the Mortality
           of A-bomb Survivors, 6:  Mortality and Radiation Dose,
           1950-74, Rad. Res., 75, 138-201  (RERF TR 1-77, Life Study
           Report 8), 1978.

Bo82       Bond V.  P. and  Thiessen J. W., Reevaluations of Dosimetric
           Factors, Hiroshima and Nagasaki, DOE Symposium Series  55,
           CONF-810928,  Technical Information Center, U.S. Department of
           Energy,  Washington, D.C.,  1982.

Bu81       Bunger B.,  Cook J. R., and Barrick M. K., Life Table
           Methodology for Evaluating Radiation Risk:  An Application
           Based  on Occupational Exposure,  Health  Phys. 40, 439-455, 1981.

Ch81       Chameaud J.,  Perraud  R.,  Chretien J., Masse R.,  and Lafuma J.,
           Contribution of Animal Experimentation to the  Interpretation
           of  Human Epidemiological Data,  in:   Proc.  Int.  Conf. on
           Hazards  in Mining:  Control, Measurement,  and Medical  Aspects,
           October  4-9, 1981, Golden, Colorado, pp. 228-235,  edited by
           Manual Gomez, Society of Mining Engineers, New York, 1981.
                                     6-53

-------
Ch83       Charles M. E., Lindop P. J., and Mill A. J., A Pragmatic
           Evaluation of the Repercussions for Radiological Protection of
           the Recent Revisions in Japanese A-bomb Dosimetry, IAEA
           SM-266/52, Proceedings Intejrnational Symposium on the
           Biological Effects of Low-Level Radiation with Special Regard
           to Stochastic and Non-stochastic Effects, Venice, IAEA,
           Vienna, April 11-15, 1983.

Co78       Cook J. R., Bunger B. M., and Barrick M. K., A Computer Code
           for Cohort Analysis of Increased Risks of Death (CAIRD), ORP
           Technical Report 520/4-78-012, U.S. Environmental Protection
           Agency, Washington, D.C., 1978.

Da72       Davie R., Butler N., and Goldstein H., From Birth to Seven,
           Longmans, London, 1972.  Cited in St84.

Do83a      Dobson R. L. and Felton J. S., Female Germ Cell Loss from
           Radiation and Chemical Exposures, Amer. J. Ind. Med,, 4^
           175-190, 1983.

Do83b      Dobson R. L., Straume J., Felton J. S., and Kwan T. C.,
           Mechanism of Radiation and Chemical Oocyte Killing in Mice and
           Possible Implications for Genetic Risk Estimation [abstract},
           Environ. Mutagen., _5, 498-499, 1983.

Do84a      Dobson R. L., and Straume T., Mutagenesis in Primordial Mouse
           Oocytes Could Be Masked by Cell Killing:  Monte Carlo
           Analysis, Environ. Mutagen. 6, 393, (1984) [Abstract].

Do84b      Dobson R. L., Kwan T. C., and Straume T., Tritium Effects on
           Germ Cells and Fertility, pp. 285-298, in Radiation Protection
           European Seminar on the Risks from Tritium Exposure,
           EUR9065en, Commission of the European Communities, 1984.

E177       Ellett W. H. and Richardson A. C. B., Estimates of the Cancer
           Risk Due to Nuclear Electric Power Generation, pp. 511-527, in
           Origins of Human Cancer, Book A., H. H. Hlatt et al., eds.,
           Cold Spring Harbor Laboratory, 1977.

EPA78      Environmental Protection Agency, Response to Comments:
           Guidance on Dose Limits for Persons Exposed to Transuranium
           Elements in the General Environment, EPA 520/4-78-010, Office
           of Radiation Programs, U.S. EPA, Washington,  D.C., 1978.

EPA81      Environmental Protection Agency, Population Exposure to
           External Natural Radiation Background in the United States,
           Technical Note ORP/SEPD-80-12, Office of Radiation Programs,
           U.S. EPA, Washington, D.C., 1981.
                                  6-54

-------
EPA84      Environmental Protection Agency,  Report  on the  review of
           Proposed Environmental Standards  for the Management and
           Disposal of Spent Nuclear Fuel, High-level and  Transuranic
           Radioactive Waste (40CFR191),  High-Level Radioactive Waste
           Disposal Subcommittee, Science Advisory  Board,  U.S.
           Environmental Protection Agency,  Washington,  D.C., 1984.

Ev79       Evans H. J., Buckton K. E., Hamilton G.  E., et  al.,
           Radiation-induced Chromosome Aberrations in Nuclear Dockyard
           Workers, Nature, 277. 531-534, 1979.

Ga82       Garriott M. L. and Grahn D., Neutron and y Ray  Effects
           Measured by the Micronucleus Test, Mut.  Res. Let., 105,
           157-162, 1982.

Go29       Goldstein L. and Murphy D. P., Etiology of Ill-health of
           Children Born After Maternal Pelvic Irradiation:  11,
           Defective Children Born After Post Conception Pelvic
           Irradiation, Amer. J. Roentgenol. Rad. Ther., 22, 322-331,
           1929.

Go80       Goodhead D. T., Models of Radiation Interaction and
           Mutagenesis, pp. 231-247, in Radiation Biology in Cancer
           Research, R.E. Meyn and H. R.  Withers, eds., Raven,  New York,
           1980.

Go82       Goodhead D. I., An Assessment of the Role  of Microdosimetry in
           Radiobiology, Rad. Res., 91, 45-76, 1982.

Gr83a      Grahn D., et al., Interpretation of Cytogenetic Damage Induced
           in  the  Germ Line of Male Mice Exposed for  Over 1 Year to
           Z39Pu Alpha Particles, Fission Neutrons, or 60Co Gamma
           Rays, Rad.  Res., 95,  566-583, 1983.

Gr83b      Grahn D., Genetic Risks Associated with Radiation Exposures
           During  Space Flight,  Adv.  Space Res., 3(8), 161-170, 1983.

Gu77       Gustavson K-H,  Hagberg B.,  Hagberg  G.,  and Sars K., Severe
           Mental  Retardation  in a  Swedish  County, I, Epidemiology,
           Gestational Age,  Birth Weight and Associated CNS Handicaps in
           Children Born 1959-70, Acta Paediatr. Scand.,  66, 373-379,
           1977.                                          —

Ha81       Hagberg B., Hagberg G.,  Lewerth  A.,  and Lindberg U., Mild
           Mental  Retardation  in Swedish School Children, I, Prevalence,
           Acta Paediatr. Scand., 70, 441-444,  1981.

He83       Herbert D.  E., Model or Metaphor?  More Comments on the BE1R
           III Report, pp. 357-390, in Epidemiology  Applied to  Health
           Pays.,  CONF-—830101, DE-83014383, NTIS, Springfield,  Virginia,
           1983.
                                  6-55

-------
H153       Hicks S. P., Developmental Malformations Produced by
           Radiation.  A Timetable of Their Development, Amer. J.
           Roentgenol. Radiat. Thera., 69, 272-293, 1953.

H154       Hicks S. P., The Effects of Ionizing Radiation, Certain
           Hormones, and Radiomimetic Drugs on the Developing Nervous
           System, J. Cell. Comp. Physiol., 43 (Suppl. 1). 151-178, 1954.

H166       Hicks S. P. and D'Amato C. J., Effects of Ionizing Radiations
           on Mammalian Development, Adv. Teratol., 1, 195-266, 1966.

Ho8l       Hornung R. W. and Samuels S., Survivorship Models for Lung
           Cancer Mortality in Uranium Miners - Is Cumulative Dose an
           Appropriate Measure of Exposure?, in:  Proc. Int. Conf. on
           Hazards in Mining:  Control, Measurement, and Medical Aspects,
           October 4-9, 1981, Golden, Colorado, 363-368, edited by Manuel
           Gomez, Society of Mining Engineers, New York, 1981.

HWC73      Health and Welfare Canada, The Testing of Chemicals for
           Carclnogenicity, Mutagenlcity and Teratogenicity, Health
           Protection Branch, HWC. Ottawa, 1973.

ICRP77     International. Commission on Radiological Protection,
           Recommendations of the International Commission on
           Radiological Protection, ICRP Publ. 26, Ann. ICRP, 1, (3),
           Pergamon Press, 1977.

ICRP79     International Commission on Radiological Protection, Limits
           for Intakes of Radionuclides by Workers, ICRP Publication 30,
           Part 1, Ann. ICRP, 2 (3/4), Pergamon Press, Mew York, 1979.

ICRP80     International Commission on Radiological Protection,
           Biological Effects of Inhaled Radionuclides, ICRP Publication
           31, Ann. ICRP, 4 (1/2), Pergamon Press, New York, I960.

Is79       Ishimaru T., Otake M., and Ishimaru M., Dose-Response
           Relationship of Neutrons and Gamma-rays to Leukemia Incidence
           of Cancer in 1950-1971, Based on the Tumor Registry, Nagasaki,
           Radiat. Res., 77, 377-394, 1979.

Ja70       Jacobsen L., Radiation Induced Fetal Damage, Adv. Teratol., £,
           95-124, 1970.

Ka73       Kato H., Late Effects in Children Exposed to the Atomic Bomb
           While In Utero, Technical Report 18-73, Atomic Bomb Casualty
           Commission, Hiroshima, 1973.

Ka82       Kato H. and Schull W. J., Studies of the Mortality of A-bomb
           Survivors, 7. Mortality, 1950-1978:  Part I, Cancer Mortality,
           Rad. Research 90, 395-432, 1982.  (Also published by the
           Radiation Effect Research Foundation as:  RERF TR 12-80, Life
           Span Study Report 9, Part 1.)
                                  6-56

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Ke72       Kellerer A. M. and Rossi H.  H., The Theory of  Dual Radiation
           Action, Curr. Topics Rad., Res. Quart.,  &, 85-158, 1972.

Ke81a      Kerr G.D., Review of Doslmetry for the Atonic  Bomb Survivors,
           la Proceedings of the Fourth Symposium on Neutron Doslmetry,
           Geasellschaft fur Strahlen- und Umweltforschung,
           Munlch-Neuherberg, Federal Republic of Germany, June 1-5,  1,
           501, Office for Official Publications of the European     ~
           Communities, Luxemburg, 1981.

KeSlb      Kerr G. D., Findings of a Recent ORNL Review of Dosinetry for
           the Japanese Atomic Bomb Survivors, ORNL/TM-8078, Oak Ridge
           National Laboratory, Oak Ridge, Tennessee, 1981.

K162       King R. C., Genetics, Oxford University Press, New York, 1962.

La78       Land C. E. and Norman J. E., Latent Periods of Radiogenic
           Cancers Occurring Among Japanese A-bomb Survivors, in:  Late
           Biological Effects of Ionizing Radiation, I, 29-47, IAEA,
           Vienna, 1978.                             "~

La80       Land C. E., Boice J. 0., Shore R. E., Norman J. E., and
           Tokunaga M.,  et al., Breast Cancer Risk from Low-Dose
           Exposures to  Ionizing Radiation:  Results of Parallel Analysis
           of Three Exposed Populations of Women, J. Natl. Cane. Inst.,
           .65, 353-376,  1980.

La83       Land C. E. and Pierce D. A., Some Statistical Considerations
           Related to the Estimation of Cancer Risk  Following Exposure to
           Ionizing Radiation,  pp. 67-89, in Epidemiology Applied  to
           Health Phys., CONF-830101,  DE83014383, NTIS,  Springfield,
           Virginia, 1983.

Le62       Lea D. E., Actions  of Radiations on Living Cells, 2nd edition,
           Cambridge University Press, 1962.

Lo81       Loewe  W.  E.  and Mendelsohn  E., Revised  Dose Estimates at
           Hiroshima and Nagasaki,  Health Phys.t 41, 663-666, 1981.

Ha59       Mandansky A., The Fitting of Straight Lines When Both
           Variables Are Subject  to Error,  J.  Aner.  Statin.  Assoc.,  54,
           173-205,  1959.                                            ~

Ma83       Mays C.  W.  and  Spleas  H., Epidemlological Studies in German
           Patients Injected with Ra-224, pp.  159-266, in Epidemiology
           Applied  to Health Physics,  CONF-830101,  DE-83014383, NTIS,
           Springfield,  Virginia,  1983.
                                   6-57

-------
 M178       Michel C.  and Fritz-Niggli H.,  Radiation-Induced Developmental
            Anomalies  in Mammalian Embryos  by Low Doses and Interaction
            with Drugs,  Stress and Genetic  Factors,  pp.  399-408,  in Late
            Biological Effects of Ionizing  Radiation,  Vol.  11,  IAEA.,
            Vienna, 1978.

 Mo67       Morgan K.  Z.  and Turner J. E.,  Principles  of Radiation
            Protection,  John Wiley and Sons,  Inc., New York,  1967.

 Mo79       Mole R. H.,  Carclnogenesls by Thorotrast and Other  Sources of
            Irradiation,  Especially Other a-Emltters,  Environ.  Res., 18,
            192-215, 1979.

 Mu29       Murphy D.  P., The Outcome  of  625  Pregnancies in Women Subject
            to Pelvic Radium or Roentgen  Irradiation,  Amer.  J.  Obstet.
            Gyn.,  18,  179-187,  1929.

 Mu30       Murphy D. P.  and DeRenyi M.,  Postconception Pelvic  Irradiation
            of the Albino Rat (Mus  Norvegieus):   Its Effects  Upon the
            Offspring, Surg.  Gynecol.  Obstet., 50, 861-863, 1930.

 NAS72       National Academy of Sciences  -  National  Research  Council, The
            Effects of Populations  of  Exposures to Low Levels of  Ionizing
            Radiation, Report of  the Committee on the  Biological  Effects
            of Ionizing Radiations  (BEIR  Report), Washington, D.C., 1972.

 NAS80       National Academy of Sciences  -  National  Research  Council, The
            Effects of Populations  of  Exposure to Low  Levels  of Ionizing
            Radiation, Committee  on the Biological Effects  of Ionizing
            Radiation, Washington,  D.C.,  1980.

 NCHS73      National Center  for Health Statistics, Public Use Tape, Vital
            Statistics - Mortality  Cause  of Death Summary - 1970,
            PB80-133333, NTIS,  Washington,  D.C.,  1973.

 NCHS75      National Center for Health Statistics, U.S. Decennial Life
            Tables  for 1969-71, .1(1),  DHEW  Publication No. (HRA) 75-1150,
            U.S. Public Health  Services,  NCHS, Rockville, Maryland, 1975.

 NCRP75      National Council  on Radiation Protection and Measurements,
            Natural Background  Radiation  in the United States,  NCR? Report
            No. 45, NCRPM, Washington, D.C., 1975.

 NCRP77      National Council  on Radiation Protection and Measurements,
            Protection of the Thyroid  Gland in the Event of Releases of
            Radioiodine, NCRP Report No.  55, NCRPM, Washington, D.C., 1977.

NCRP80     National Council  on Radiation Protection and Measurements,
            Influence of Dose and Its  Distribution in  Time on
           Dose-Response Relationships for Low-LET Radiation, NCRP Report
           No. 64, NCRPM, Washington, D.C., 1980.
                                  6-58

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Ne56       Neel J. V. and Schull W.  J.,  The Effect  of  Exposure to the
           Atomic Bombs on Pregnancy Termination in Hiroshima and
           Nagasaki, National Academy of Sciences,  Publ.  461, Washington,
           D.C., 1956.

Ne70       Nelson K. B. and Deutschberger J.,  Head  Size at One Year  as a
           Predictor of Four-Year I.Q.,  Develop. Med.  Child Neurol., 12,
           487-495, 1970.

Of80       Oftedal P. and Searle A.  G.,  An Overall  Genetic Risk
           Assessment for Radiological Protection Purposes, J. Med.
           Genetics, 17, 15-20, 1980.

Oh65       Ohzu E., Effects of Low-Dose X-Irradiation on Early Mouse
           Embryos, Rad. Res. 26, 107-1JL3, 1965.

ORNL84     Oak Ridge National Laboratory, Age Dependent Estimation of
           Radiation Dose,  [in press], 1984.

Ot83       Otake.M. and  Schull W. H., Mental Retardation in Children
           Exposed  In Utero to the Atomic  Bombs:  A Reassessment,
           Technical Report RERFTR 1-83, Radiation  Effects Research
           Foundation, Hiroshima, 1983.

Ot84       Otake M.  and  Schull W. J., In Utero  Exposure  to A-bomb
           Radiation and Mental  Retardation:  A Reassessment, Brit. J.
           Radiol.,  57,  409-414, 1984.

 P152       Plummer G.  W.,  Anomalies  Occurring in Children Exposed in
           Utero to the  Atomic  Bomb in  Hiroshima,  Pediat.,  10,  687-692,
           1952.                                           ~~

 Po78       Pohl-Ruling J., Fischer  P.,  and Pohl E., The Low-Level Shape
           of Dose Response for Chromosome Aberration, pp.  315-326  in
           Late Biological Effects  of Ionizing  Radiation, Volume II,
            International Atomic Energy Agency,  Vienna, 1978.

 RERF78    Radiation Effects Research Foundation,  1 April 1975  - 31 March
            1978.  RERF Report 75-78, Radiation Effects Research
            Foundation, Hiroshima,  1978.

 RERF83     Radiation Effects Research Foundation,  Reassessment of Atomic
            Bomb Radiation Dosimetry in Hiroshima and Nagasaki,  Proc. of
            the U.S.-Japan Joint Workshop, Nagasaki, Japan, Feb. 16-17,
            1982, Radiation Effects Research Foundation, Hiroshima,  730,
            Japan, 1983.

 RERF84     Radiation Effects Research Foundation,  Second U.S.-Japan Joint
            Workshop for Reassessment of Atomic Bomb Radiation Dosimetry
            in Hiroshima and Nagasaki, Radiation Effects Research
            Foundation,  Hiroshima, 730,  Japan,  1984.
                                   6-59

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Ro78       Rowland R. E., Stehney A. P., and Lucas H. F., Dose Response
           Relationships for Female Radium Dial Workers, Rad. Res. 76,
           368-383, 1978.

Ru50       Russell L. B., X-ray Induced Developmental Abnormalities in
           the Mouse and Their Use in the Analysis of Embryologlcal
           Patterns, I.  External and Gross Visceral Changes, J. Ezper.
           Zool., 114. 545-602, 1950.

Ru53       Rugh R., Vertebrate Radioblology:  Embryology, Ann. Rev. Nucl.
           Sci., 3, 271-302, 1953.

Ru54       Russell L. B. and Russell W. L., An Analysis of the Changing
           Radiation Response of the Developing Mouse Embryo, J. Cell.
           Comp. Physiol., 43 (Suppl. 1), 103-149, 1954.

Ru56       Russell L. B., X-Ray Induced Developmental Abnormalities in
           the Mouse and Their Use in the Analysis of Embryological
           Patterns, II.  Abnormalities of the Veretebral Column and
           Thorax, J. Exper. Zool., 131, 329-390, 1956.

Ru57       Russell L. B., Effects of Low Doses of X-rays on Embryonic
           Development In the Mous, Proc. Soc. Exptl. Biol. Med., 95,
           174-178, 1957.

Ru70       Rugh R., The Effects of Ionizing Radiation on the Developing
           Embryo and Fetus, Seminar Paper No. 007, Bureau of
           Radiological Health Seminar Program, Public Health Service,
           Washington, D.C., 1970.

Ru71       Rugh R., X-ray Induced Teratogenesis in the Mouse and Its
           Possible Significance to Man, Radiol., 99, 433-443, 1971.

Sa82       Satoh C., et al., Genetic Effects of Atomic Bombs, in:  Human
           Genetics, Part A:  The Unfolding Genome, A. R. Lias, Inc., New
           York, 267-276, 1982.

Sc81       Schull W. J., Otake M., and Neel J. V., Genetic Effects of the
           Atomic Bombs:  A Reappraisal, Science, 213, 1220-1227, 1981.

Se69       Senyszyn J. J. and  Rugh R.,  Hydrocephaly Following Fetal
           X-Irradiation, Radiol., 93,  625-634, 1969.

Sm78       Smith P. G. and Doll R., Radiation-Induced Cancers in Patients
           with Ankylosing Spondylitis Following a Single Course of X-ray
           Treatment,  in:  Proc.  of the IAEA Symposium, Late Biological
           Effects of Ionizing Radiation, 1,  205-214, IAEA, Vienna, March
           1978.
                                  6-60

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Sp83       Spiers F.  W.,  Lucas  H.  F.,  Rundo J., and Anaat G. A., Leukemia
           Incidence  In the U.S. Dial  Workers, in:  Conference Proc. on
           Radioblology of Radium  and  the Actlnldes In Man,
           October 11-16, 1981, Health Phys., 44  Suppl. 1. 65-72, 1983.

St21       Stettner E., Eln welterer Fall elner Schadlngung elner
           menschlchen Frucht durch Roentgen Bestrahlung., Jb.
           Klnderhellk.  Phys.  Erzleh., 95, 43-51, 1921.

St8l       Straume T. and R. L. Dobson, Implications  of New Hiroshima and
           Nagasaki Dose Estimates: Cancer Risks and Neutron RBE, Health
           Phys. 41,  666-671, 1981.

St84       Stein Z. A.  and Susser  M. W., The Epidemiology of Mental
           Retardation, in Epidemiology of Pediatric  Neurology,  B.
           Schoenberg,  editor,  Marcel  Dekker,  Inc., New York,  [in press],
           1984.

To80       Tobias C.  A., et al., The Repalr-Mlsrepalr Model, pp. 195-230,
           in R. E. Meyn and H. R. Withers,  eds., Raven,  New York, 1980.

To84       Tokunaga M., Land C. E., Yamamoto T.,  Asano  M., Takioka S.,
           Ezaki E.,  and Nlshimari I., Incidences of  Female  Breast Cancer
           Among Atomic Bomb Survivors, Hiroshima and Nagasaki,
           1950-1980, RERF TR 15-84, Radiation Effects  Research
           Foundation, Hiroshima,  1984.

U182       Ullrich R. L., Lung Tumor Induction in Mice:  Neutron RBE at
           Low Doses, DE 82009642.  National Technical Information
           Service, Springfield,  Virginia,  1982.

UNSCEAR58  United Nations, Report of the United Nations Scientific
           Committee on the Effects of Atomic Radiation, Official
           Records:  Thirteenth Session, Supplement No. 17 (A/3838),
           United Nations, New York, 1958.

UNSCEAR62  United Nations, Report of the United Nations Scientific
           Committee on the Effects of Atomic Radiation, Official
           Records:  Seventeenth Session, Supplement No. 16 (A/5216),
           United Nations, New York, 1962.

UNSCEAR66  United Nations, Report of  the United Nations Scientific
           Committee on the Effects of Atomic Radiation, Official
           Records:  Twenty-First Session, Supplement No. 14 (A/6314),
           United Nations, New York,  1966.

UNSCEAR69  United Nations, Report of  the United Nations Scientific
           Committee on  the Effects of Atomic Radiation, Supplement
           No.  13  (A/7613), United Nations, New York, 1969.
                                   6-61

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 UNSCEAR72  United Nations Scientific Committee on the Effects  of Atomic
            Radiation, Ionizing Radiation:   Levels and Effects,
            volume II:  Effects, Report to  the General Assembly.  Sales
            No. E. 72.  IX.18., United Nations, New York,  1972.

 UNSCEAR77  United Nations Scientific Committee on the Effects  of Atomic
            Radiation, Sources and Effects  of Ionizing Radiation, Report
            to the General Assembly,  with Annexes,  Sales No. E.77 IX.1.,
            United Nations, New York, 1977.

 UNSCEAR82  United Nations Scientific Committee on the Effects  of Atomic
            Radiation, Ionizing Radiation:   Sources and Biological
            Effects,  1982  Report to the General Assembly,  Sales No. E.82.
            IX.8,  United Nations,  New York,  1982.

 Va80       Van Buul  P.  P.  W.,  Dose-response Relationship  for X-ray
            Induced Reciprocal Translocations in Stem  Cell Spermatogonla
            of the Rhesus  Monkey (Macaca mulatta),  Mutat.  Res., 73,
            363-375,1980.  (Cited in UNSCEAR82.)

 Wa83       Hakabayashl  T., Kato H.,  Ikeda T.,  and  Schull  W. J., Studies
            of the Mortality of A-bomb Survivors, Report 7, Part III,
            Incidence of Cancer in 1959-78 Based on the Tumor Registry
            Nagasaki,  Radiat. Res., 93,  112-142, 1983.

 Wh83       Whlttemore A.  S. and McMillan A., A Lung Cancer Mortality
            Among  U.S. Uranium Miners:  A Reappraisal, Technical Report
            No.  68, SIAM Inst. Math.  Soc., Stanford University, Stanford,
            1983.

 WH075       World  Health Organization,  International Statistical
            Classification of Diseases, Injuries, and Causes of Death, 9th
            Revision, WHO, Geneva, 1975.

 W154        Wilson J. G., Differentiation and the Reaction of Rat Embryos
            to Radiation, J. Cell. Comp. Physiol., 43 (Suppl. 1). 11-37,
            1954.

 W665       Wood J. W.,  Johnson K. G., and Omari Y., In Utero Exposure to
            the Hiroshima Atomic Bomb:  Follow-up at Twenty Years,
           Technical Report 9-65, Atomic Bomb Casualty Commission,
           Hiroshima, 1965.

Wo66       Wood J. W*, Johnson K. G., Omari Y., Kawamoto S., and Keehn
           R. J., Mental Retardation in Children Exposed in Utero to the
           Atonic Bomb—Hiroshima and Nagasaki, Technical Report 10-66,
           Atomic Bomb Casualty Commission, Hiroshima, 1966.
                                6-62

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   Chapter 7:   MOVEMENT AKD HEALTH RISKS OF RADIONUCLIDE RELEASES TO THE
                          ACCESSIBLE ENVIRONMENT
7.1  Introduction

     This chapter describes analyses used to assess the health risks
caused by the environmental transport of radionucliUes once they are
released from a repository to the accessible environment.   As part of its
program to develop 40 CFR Part 191, the Agency estimated population
health risks for a 10,000-year period following disposal in mined geo-
logic repositories (see Chapter 8). These estimates were used in selec-
ting the containment requirements in the disposal standards.  The objec-
tive of this chapter is:  1) to describe the environmental pathways that
were considered when calculating the environmental risk commitments
(ERC's: fatal cancers and serious genetic defects) that could occur as a
result of releases of radionuclides from disposal systems; and 2) to
summarize the results of these calculations.  A complete description of
the analysis is described in Sm85.  This chapter also describes how the
release limits of the containment requirements were derived from the
results of these calculations.

     In performing these long-term assessments of population health ef-
fects, the Agency recognizes that it is pointless to try to make precise
projections of the actual risks due to radionuclide releases from reposi-
tories.  Population distributions, food chains, living habits, and tech-
nological capabilities will undoubtedly change in major ways over 10*000
years.  Unlike geological processes, they can be realistically predicted
only for relatively short times.  Accordingly, very general models of
environmental pathways were formulated as opposed to the detailed analyt-
ical techniques that would be appropriate for near-term environmental
assessments of specific facilities.  Population characteristics similar
to those of today were assumed.

     The models discussed in this chapter consider risks to populations,
as opposed to risks to individuals.  Therefore, individual risks caused
by potential releases from a repository cannot be determined from these
analyses.  Analyses that assess individual risks are described in Chap-
ter 8.                                                               *
                                   7-1

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7.2  Methodology

     Radionuclides can be released from geologic repositories and move
through the environment through four pathways: 1) to surface water (e.g.,
a river) through ground water, 2) to an ocean through surface water, 3)
to a land surface directly, or 4) to multiple pathways after the very
unlikely possibility of disruption by a volcano or a meteorite.  For each
of these four release modes, radionuclide movement through the geosphere
and the biosphere to the population was modeled, and an estimate was made
of the intake by or exposure to the population through each of these
environmental pathways.  The environmental pathways included for each of
the release modes are described in Table 7.2-1.

     Risk conversion factors per unit intake or per unit external exposure
were applied to the radionuclide concentrations output by the model to
estimate fatal cancers and serious genetic effects to all generations per
curie of each different radionuclide released to the accessible environ-
ment.  The results were used to specify the release limits in Table 1 of
40 CFR Part 191, based on a consideration of only excess fatal cancers.
The genetic effects to all generations were lower than the estimated
fatal cancers by a factor of two or more for all radionuclides and were
not used to establish the release limits.  The risk conversion factors
used to estimate fatal cancers are listed in Table 7.2-2 (Sm85).

     Health effects were calculated for the entire population exposed to
the releases from a repository; calculations were not terminated at some
arbitrary distance from the repository.  A time integration was performed
to obtain the sum of the health effects from the time the repository is
sealed ("disposal") until a specified time in the future (usually 10,000
years after disposal).  The radioactivity intakes and exposures were then
converted to population EEC's by multiplying by the appropriate risk
conversion factors.  The following sections summarize the factors consi-
dered in the calculation of the population intake of radioactivity for
the internal pathways—or the integrated population exposure for the
external pathways—for each of the four release modes.

7.3  Releases to Surface Water

     In the surface water release pathway, the repository containment is
assumed to be breached—after some initial period—and ground water cir-
culates through the repository into the surrounding geologic media and
eventually to an aquifer.  The aquifer then flows underground until it
intersects a river.  To determine the total release to the surface water
(river), the release rate was integrated over the time period of inter-
est.  The Integrated release rate, in equation form, was then used to
compute surface water concentrations for use with several environmental
pathways.  These are discussed in the following subsections.
                                   7-2

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         Table 7.2-1.   Release modes and environmental pathways
Release mode
Pathways Included in this release mode
Releases to river
Releases to ocean
Releases directly
  to land surface
Releases due to volcano/
  meteorite interaction

  Releases
    directly to land
  Releases to air
    over land
  Releases to air
    over ocean
Drinking water ingestion
Freshwater fish ingestion
Food crops ingestion
Milk ingestion
Beef Ingestion
Inhalation of resuspended material
External dose, ground contamination
External dose, air submersion

Ocean fish ingestion
Ocean shellfish ingestion

Food crops ingestion
Milk ingestion
Beef ingestion
Inhalation of resuspended material
External dose, ground contamination
External dose, air submersion
Food crops ingestion
Milk ingestion
Beef ingestion
Inhalation of resuspended material
External dose, ground contamination
External dose, air submersion

Food crops ingestion
Milk ingestion
Beef Ingestion
Inhalation of dispersed  and
  resuspended material
External dose, ground contamination
External dose, air submersion

Ocean  fish ingestion
Ocean  shellfish  ingestion
                                   7-3

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                             Table 7.2-2.   Fatal cancer risk conversion factors
                                                                               (a)

Fatal
cancers
Inhalation
Radionuclide
C-14
M-59
Sr-90
Zr-93
Tc-99
Sn-126
1-129
Cs-135
Cs-137
Sm-151
Pb-210
Ra-226
1
3.05E-3
4.76E-1
4.52E+2
2.72E+1
6.12E+0
5.72E-H
1.61E+1
1.27E+0
8.49E+0
5.27E+0
2.27E+4
4.38E+4
2
3.05E-3
4.76E-1
5.19E-H
6.60E+0
6.12E+0
5.72E+1
1.61E+1
1.27E+0
8.49E+0
5.27E+0
2.99E+3
5.33E-1-3
per Ci intake

Ingestion
1 2
4.32E-1
3.76E-2
2.29E+0
1.27E-1
5.37E-1
2.04E+0
2.41E+1
1.82E+0
1.24E+1
3.46E-2
4.13E+2
4.91E+2
4.32E-1
3.76E-2
2.85E+1
1.27E-1
5.37E-1
2.04E+0
2.41E+1
1.82E+0
1.24E+1
3.46E-2
4.13E+2
4.91E+2
Fatal cancers
Air
submersion
per
Ci-y/m3
0
4.10E-2
0
1 . 23E-1
5.97E-4
2.54E+3
7.57E+0
0
7.18E+2
8.15E-4
1.34E+0
2.35E+3
from external doses
Ground
contamination
per
Ci-y/m2
0
8.87E-3
0
1.89E-2
1.41E-5
5.11E+1
3.98E-1
0
1.43E+1
9.43E-5
6.09E-2
4.20E+1
(continued)

-------
                   Table 7.2-2.  Fatal cancer risk conversion factors     (continued)
Ul
Fatal cancers per Ci intake
Inhalation (b)
Radionuclide
Ra-228
Ac-227
Th-229
Th-230
Th-232
Pa-231
U-233
U-234
U-235
U-236
U-238
Np-237
1
7.98E+4
6.97E+4
6.45E+4
6.89E+4
1.05E+5
1.03E+5
2.42E+4
2.07E+4
2.03E+4
1.96E+4
1.86E+4
2.89E+4
2
1.58E+4
3.74E+4
2.82E+4
2.05E+4
2.94E+4
6.19E+4
3.70E+3
2.26E+3
2.72E+3
2.14E+3
2.04E+3
2.46E+4
Ingestion J
1 2
9.71E+1
2.85E+2
8.55E+1
5.13E+2
1.17E+2
4.67E+2
5.21E+0
9.38E-1
5.86E+0
8.86E-1
1.99E+0
1.86E+2
9.71E+1
2.85E+2
8.55E+1
5.13E+2
1.17E+2
4.67E+2
5.07E+1
4.61E+1
5.02E+1
4.35E+1
4.84E+1
1.86E+2
Fatal cancers from external doses
Air
submersion
per
Ci-y/m9
3.43E+3
4.81E+2
3.30E+2
2.35E+3
3.43E+3
5.17E+2
1.68E-H
1.63E-1
2.01+2
1.27E-1
2.36E+1
2.83E+2
Ground
contamination
per
Ci-y/m2
5.88E+1
1.05E+1
7.46E+0
4.20E+1
5.88E-H
1.14E+1
3.84E-1
1.63E-2
4.60E+0
1.48E-2
5.17E-1
6.39E+0
    (continued)

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             Table 7.2-2.  Fatal  cancer risk conversion  factors     (continued)
Fatal cancers per Ci intake Fatal cancers from external doses
Inhalation 
-------
7.3.1  Drinking Water

     It was assumed that the population receives 65 percent of its drink-
ing water from surface waters with no reduction in radionuclide concen-
trations due to water treatment (Mu77).  The annual intake rate of drink-
ing water and water-based drinks by an individual is 600 liters (ICRP75);
thus, 390 liters was assumed to be supplied by surface water.  The aver-
age ratio of the population drinking water to the river flow rate is 3.3
x 10   person-year/liter based upon an assumed world population of 10
billion persons and an annual flow rate for the world's rivers of 3 x
10lb liters (UNSCEAR77).  This ratio is within the range of similar
values associated with various river basins in the United States (Sm85).
The average fraction of a river's flow that is used for drinking water is
obtained by combining the fraction of drinking water which is surface
water, the drinking water rate for an individual, and the average ratio
of population drinking water to the river flow rate.  This is numerically
the same as the total intake of a radionuclide by the population per
curie of radionuclide released to the surface water.

7.3.2  Ingestion of Fish

     Fish caught in the river are assumed to contain radionuclides due to
uptake from the water.  The amount of radionuclides accumulated in the
fish (in terms of Ci per kg of fish body weight) is a direct function of
the radionuclide concentration (Ci per liter) in the surface water.  The
fraction of the radionuclides released to the surface water that is
ingested by the population through fish consumption is obtained by calcu-
lating the quantity of radionuclides in the fish through the use of
bioaccumulation factors, and by determining an average ratio of the
population's fish ingestion rate to the river flow rate (3.3 x 10"7
man-kilogram/liter*).  The bioaccumlation factors for fish are given in
Table 7.3-1.

7.3.3  Ingestion of Food Raised on Irrigated Land

     Surface water containing radionuclides released from the repository
may be used to spray or irrigate farm land, leading to direct deposition
of radionuclides onto the crops and the land surface below the crops.
The average fraction of the river flow used for irrigation was assumed to
be 0.1, based on the United States average of 0.07  (Sm85, Mu77).  In
addition, Irrigated plants that had incorporated radionuclides through
their leaves and root systems are consumed by humans as food, or are
consumed by either dairy or beef cattle that transfer radionuclides to
milk and meat.  The amounts of radioactivity consumed through these
pathways was determined by using a radionuclide-specific intake factor
for each pathway (food crops, milk, and beef) as given in Table 7.3-2,
the fraction of the river flow used for irrigation, and the average
number of people that can be fed per unit area of land by each of the
pathways as given in Table 7.3-3 (Sm85).  The average consumption of


*This ratio is determined by multiplying the person-year/liter (discussed
 in Section 7.3.1) by the assumed annual individual fish consumption of
 1.0 kg/year (UNSCEAR77).

                                   7-7

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      Table 7.3-1.   Bloaccumulation factors for freshwater fish
Radionuclide
C-14
Ni-59
Sr-90
Zr-93
Tc-99
Sn-126
1-129
Cs-135
Cs-137
Sm-151
Pb-210
Ra-226
Ra-228
Ac-227
Th-229
Th-230
Th-232
Pa-231
U-233
U-234
U-235
U-236
U-238
Np-237
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Am-241
Am-243
Cm-245
Cm-246
Bloaccumulation factor
(Ci/kg per Ci/llter)
NA(a>
l.OOE+2
1.10E+1 (Ho79)
3.33E+0
4.30E+1 (B182)
3.00E+3
3.30E+1 (Ho79)
1.30E+3 (Ho79)
1.30E+3 (Ho 7 9)
2.50E-H
l.OOE+2
5.00E+1
5.00E+1
2.50E+1
3.00E+1
3.00E+1
3.00E+1
1.10E+1
l.OOE+1
l.OOE+1
l.OOE+1
l.OOE-H
l.OOE+1
5.00E+2 (Sc83)
8.00E+0 (R183)
8.00E+0 (R183)
8.00E+0 (R183)
8.00E+0 (R183)
8.00E+0 (R183)
8.10E+1 (R183)
8.10E+1 (Ri83)
2.50E+1
2.50E+1
(a)
   NA - Not  Applicable.




   Source:   Th72  (unless otherwise noted)
                              7-8

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          Table 7.3-2.   Radionuclide intake factors  for  farm
     products raised in areas  using contaminated irrigation water
(a)
   NA - Not Applicable.

   Source:  Sm85
                                    Radionuclide intake factor
                                 (Ci intake per Ci/m  deposited)
Radionuclide
C-14
Ni-59
Sr-90
Zr-93
Tc-99
Sn-126
1-129
Cs-135
Cs-137
Sm-151
Pb-210
Ra-226
Ra-228
Ac-227
Th-229
Th-230
Th-232
Pa- 231
U-233
U-234
U-235
U-236
U-238
Np-237
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Am- 241
Am-243
Cm-245
Cm-246
Food Crops
HA«
4.38E+0
2.57E+0
4.21E+0
1.57E+0
1 . 10E+0
1.17E+1
1.40E+1
8.51E-1
5.47E-1
4.98E-1
6.62E-1
3.95E-1
3.95E-1
7.33E-1
2.77E+0
6.73E+0
6.92E-1
1.19E+0
1.19E+0
1.19E+0
1.19E+0
1.19E+0
5.42E-1
3.92E-1
4.77E-1
4.53E-1
3.90E-1
4.89E-1
4.35E-1
4.87E-1
4.10E-1
4.08E-1
Milk
NA
3.22E-1
1.07E+0
8.18E-2
4.00E+0
3.04E-1
1.03E+1
8.04E+0
1.74E+0
4.54E-3
5.75E-2
1 . 26E-1
9.81E-2
4.36E-3
1.49E-3
3v87E-3
8.51E-3
1.43E-3
1.57E-1
1.57E-1
1.57E-1
1.57E-1
1.57E-1
2.52E-3
2. 17E-5
2.37E-5
2.32E-5
2.17E-5
2.40E-5
9.45E-5
1.03E-4
4.67E-3
4.63E-3
Meat
NA
2.48E-1
8.20E-2
2.10E-H
1.31E+0
9.36E+0
2.78E+0
8.84E+0
1.91E+0
4.37E-1
2.66E-2
6.26E-2
4.53E-2
2.10E-3
6.87E-4
1.79E-3
3.93E-3
1.10E-3
2.01E-2
2.01E-2
2.01E-2
2.01E-2
2.01E-2
1.94E-2
1.67E-4
1.83E-4
1.79E-4
1.67E-4
1.85E-4
3.18E-4
3.48E-4
3.14E-4
3.12E-4
                                   7-9

-------
 these various  food  products was  then  determined.  Combining  the consump-
 tion of  food products with the radionuclide  content of  the products yields
 an estimate of the  fraction of the  radionuclides  released to the  surface
 water that are transferred to crops by  irrigation and ultimately  consumed
 by populations.


     Table 7.3-3.   Values for persons fed per unit area of land
                                                    o
                  Food                 Person  fed/m
Vegetative Food Crops
Milk
Meat
4.79 E-3
1.56 E-3
7.85 E-5
          Source:  Sm85
 7.3.4   Inhalation of Resuspended Material

     Some of the radionuclides deposited on the soil by Irrigation are
 resuspended into the air.  The air concentration of resuspended radio-
 nuclides corresponding to the fraction of radionuclides released to the
 surface water that wind up in water used for irrigation is calculated
 using a resuspension factor of 10~9/m and the integrated soil surface
 concentration (Be76, Ne78).  The population intake of these radionuclides
 is then calculated using an annual lnhalation_rate of 8400 cubic meters
 and an  average population density of 6.7 x 10~5 persons per square meter
 (ICRP75, UNSCEAR77, Wo79).

 7.3.5   External Exposure from Air Submersion

     The radionuclides resuspended into the air can also cause submersion
 exposures to the population.  These exposures are also based on the
 integrated air concentration to which the population is exposed and are
 calculated from the integrated air concentration, the average population
 density, and a shielding and occupancy factor 0.33 (UNSCEAR77, Wo79).

 7.3.6   External Exposure from Ground Contamination

     Finally, the radionuclides deposited on the ground during irrigation
 can also cause external exposures to persons in the area.  Throughout the
 irrigation period, radionuclides continue to build up on the ground until
either irrigation stops or equilibrium is reached with losses through the
soil.   The methods for estimating these exposures are similar to those
applied for air submersion.
                                   7-10

-------
7.4  Releases to an Ocean

     Releases to a surface water system are assumed to subsequently dis-
charge into an ocean.  Since radionuclide decay during travel in the
river or depletion of the radionuclide inventory due to river water use
and sedimentation is not considered, the radionuclide releases to an
ocean are equal to the releases to a surface water.  The ocean pathway
model has two compartments consisting of a shallow upper layer in which
it is assumed that all edible seafood is grown, and a lower layer that
includes the remainder of the ocean.  Differential equations were devel-
oped whose solutions describe the quantities of radionuclides in these
two compartments over time.  The equation for the upper compartment
inventory was divided by the volume of the compartment to determine the
time-dependent concentation of radionuclides in the upper layer.  This
concentration was then used to estimate the fraction of the radionuclides
released to the river that is consumed by the population due to bioaccumu-
lation of radionuclides in ocean fish and shellfish.

7.5  Releases Directly to Land Surface

     For the land surface pathway models, some of the radioactive waste
from the repository is assumed to be brought to the surface after an
event such as inadvertent intrusion while drilling for resources.  Such
releases to the surface are assumed to be over a small area and a short
period of time; as such, they can be modeled as instantaneous point
sources.  The mechanisms distributing the material to humans are resus-
pension and subsequent dispersion in the atmosphere.  After the initial
release to the land surface is determined, a time-dependent release rate
to the air is estimated using a simple exponential model that depletes
the land surface source to account for resuspension and radioactive
decay.  This release rate is applied in conjunction with an atmospheric
dispersion equation to predict air concentrations as a function of time
and distance from the source; these air concentrations are then used to
estimate ground surface concentrations as a function of time and dis-
tance.  Once ground surface concentrations are determined, the techniques
used to calculate population intake are similar to those described for
the surface water release mode.  The pathways considered for releases to
land surface are ingestion of food raised on land contaminated with
radionuclides, including food crops, milk, and meat; inhalation of resus-
pended radionuclides; external exposure due to air submersion; and exter-
nal exposure due to ground contamination.

7-6  Releases Due to a Volcanic Eruption or Meteorite Impact

     Releases to the land surface and directly to the air can be caused
by the extremely unlikely events of disruption by volcanoes or meteor-
ites.  The methodology described for the land surface release mode is
used for the material released to the land surface.  For the material
released to the air, it is assumed that the radionuclides would be quickly
dispersed in such a manner that they would eventually be distributed uni-
formly within the troposphere.  The airborne material is divided into the
                                   7-11

-------
 fraction over land and the fraction over water using the ratio of earth
 land surface and earth water surface.  Compartment models, with their
 systems of coupled differential equations, were used to estimate the
 quantity of radlonuclides reaching the land surface or ocean.  Finally,
 the amount of radionuclides or radiation exposure reaching people was
 estimated through the same pathways described for the land surface or the
 ocean, respectively.

 7.7  Special Considerations for Carbon-14 Environmental Risk Commitment

      Unlike the other radionuclides considered in these analyses, stable
 carbon constitutes a significant fraction of the elemental composition of
 the human body and man's diet.   Thus, transport processes through the
 different environmental pathways and within plants, animals,  and man that
 apply to trace quantities of other radionuclides do not necessarily apply
 to radionuclides such as carbon-14 (C-14), where the corresponding stable
 elements are present in such quantities that saturation effects are
 significant (Mo79).

      Atmospheric releases of C-14 as carbon dioxide can be evaluated
 using a diffusion-type model of the carbon cycle developed by Killough
 (K177).  It seems clear that this model is the correct  calculational
 procedure to use for releases for the volcano/meteorite release mode
 where it is assumed  that  high temperatures would cause  carbon releases to
 be oxidized to carbon dioxide.   Models are not available to explicitly
 treat the ERC calculations for  C-14 released to water,  land surfaces,  or
 air in a chemical form other than carbon dioxide.   A review of the litera-
 ture indicated that  the chemical form of C-14 released  in the water and
 land surface release modes is not well known.   Also,  the rate of oxida-
 tion to carbon dioxide of other chemical forms of  C-14  over the extensive
 Integration period is not known for these release  modes.   Considering  all
 these uncertainties,  it was  concluded  that the most  prudent course was to
 use the Killough carbon dioxide model  for all  four release  modes,  real-
 izing that  this  probably  leads  to conservative estimates  of the ERC for
 the water and  land release modes.

      The environmental risk commitment  for C-14  is obtained by calcula-
 ting the total body  environmental dose  commitment  (EDC) and multiplying
 by  a fatal  cancer risk conversion  factor.   Values  of  the  total body
 environmental dose commitment per  curie of C-14  released  to the  atmos-
 phere have been  calculated by Fowler using the Killough model  (Fo79,
 K177).   It  is estimated that  the  ingestion pathway contributes  99 percent
 of  the  carbon-14 environmental  dose commitment  (Fo76); however,  it  is
 assumed that the ingestion pathways contribute  100 percent  for purposes
 of  computational convenience.   For estimating  the  environmental  dose
 commitment, Fowler's curve of worldwide EDC to the total body per curie
 release versus time after release was used.

     The environmental risk commitment is obtained by multiplying the
total body environmental dose commitment by the fatal cancer risk factor
                                   7-12

-------
of  1.46 x  10  **  fatal cancers per total body man-rem as given by Fowler
 (Fo79).  For  C-14, this is the total environmental risk commitment for
all the pathways within each release mode; the methodology is not applied
separately for  each pathway.

7.8  Fatal Cancers per Curie Released to the Accessible Environment

     This section presents the results of all analyses in terms of the
premature fatal cancers induced (over 10,000 years) for each curie of the
various radionuclides that may be released to the accessible environment.
These fatal cancer estimates have been used to develop the radionuclide
release limits  in Table 1 of 40 CFR Part 191 of the final rule.  The
fatal cancer  estimates for releases to surface water (the sum of releases
to a river and  releases to the ocean), to land surfaces, and to the
atmosphere are  tabulated in Table 7.8-1.  Table 7.8-2 shows how the
various environmental pathways contribute to the fatal cancer per curie
released estimate for releases to surface water.  As can be seen from
Table 7.8-2,  the dominant pathway for each radionuclide is usually inges-
tion of surface crops irrigated with contaminated water.

7.8.1  Development of Release Limits for 40 CFR Part 191

     The analyses described in this chapter were used to develop radio-
nuclide release limits that correspond to the level of protection chosen
for the containment requirements of the final rule (Section 191.13).
Since releases  to surface water through ground water are usually the most
important release mode for mined repositories, and since the health
effects per curie released are usually the highest for this release mode,
the release limits in 40 CFR Part 191 were based solely on the surface
water release mode.

     To develop the release limits, the appropriate population risk level
must first be chosen.  The Agency has chosen to base the containment
requirements on a population risk level of no more than 1,000 premature
cancer deaths over 10,000 years from disposal of 100,000 metric tons of
heavy metal (MTHM) contained in spent fuel (or from disposal of the
high-level radioactive wastes produced by this much spent fuel).  For
convenience, the release limits in 40 CFR Part 191 are stated in terms of
1,000 MTHM and  can be adjusted to reflect the actual amount of waste in a
disposal system.  Therefore, the release limits in 40 CFR Part 191 are to
be the amount of each radionuclide that would cause 10 health effects
over 10,000 years.

     Table 7.8-3 summarizes the procedure used to arrive at the release
limits in 40 CFR Part 191, Table 1 of the final rule.  First, the number
of fatal cancers caused per curie released to surface water for each
radionuclide  (the first column of Tables 7.8-1 and 7.8-2) was divided
 This C-14 fatal cancer risk factor is less than that used for most other
 radionuclides because a large percentage of the total body dose from
 C-14 is to adipose tissue and is not effective in producing cancer
 (Fo79).

                                   7-13

-------
 Table 7.8-1.
Fatal cancers per curie released to the accessible
 environment for different release nodes
    Radionuclide
     Releases to
    surface water
Releases to
land surface
Releases due
 to violent
interactions
                                                                 (a)
C-14
Ni-59
Sr-90
Zr-93
Tc-99
Stx-126
1-129
Cs-135
Cs-137
Sm-151
Pb-210
Ra-226
Ra-228
Ac-227
Th-229
Th-230
Th-232
Pa-231
U-233
U-234
U-235
U-236
U-238
Np-237
Pu-238
Pu-239
Pu-240
Pu-241
Fu-242
Am- 241
An-243
Cm-245
Cn-246
5.83E-02
4.78E-05
2.26E-02
1.59E-04
3.68E-04
1.25E-02
8.09E-02
7.76E-03
1.07E-02
9.78E-06
1.25E-01
1.68E-01
2.42E-02
6. 8 7 E- 02
6.20E-02
7.25E-01
3.83E-01
1.50E-01
2.18E-02
1.98E-02
2.19E-02
1.87E-02
2.08E-02
8.66E-02
4.27E-02
5.20E-02
5.03E-02
2. 18E-03
5.01E-02
5.80E-02
6.81E-02
1.24E-01
6.00E-02
5.83E-02
6.79E-07
3.76E-05
2.26E-05
5.65E-08
1.38E-03
3.96E-03
5.75E-04
2.19E-05
6.71E-08
1.52E-04
5.62E-03
1.57E-05
1.24E-04
1.90E-02
3.86E-01
3.76E-01
2.36E-02
7.51E-04
6.54E-04
8.42E-04
6.18E-04
6.90E-04
1.21E-04
3.10E-04
6.23E-03
5.22E-03
2.50E-06
6.34E-03
1.05E-03
2.45E-03
8.08E-03
3.54E-03
5.83E-02
2.89E-05
1.16E-03
1.22E-04
1.99E-04
2.73E-02
5.57E-02
4.91E-03
3.39E-03
4.72E-06
4.31E-02
7.20E-02
2.78E-02
3.82E-02
5.06E-02
1.26E+00
3.73E-01
1.28E-01
7.75E-03
5.94E-03
8.27E-03
5.62E-03
5.67E-03
2.83E-02
2.07E-02
1.20E-02
1.15E-02
9.36E-04
1.09E-02
2.54E-02
3.40E-02
6.09E-02
2.89E-02
(a)
   Interactions of a metorite or a volcanic eruption with a repository.
                               7-14

-------
                         Table 7.8-2.   Fatal  cancers per curie released  to the accessible
                                      environment for releases  to surface water
                                                      Injeetlon
                                                                                   Inhalation
                                                                                                 external doc*
Cn
Radio- Drinking Freahwater Surface Ocean Ocean Reauapended Ground Air
nucllde TOTAL vater flah cropa Milk Beef flah ahellflah material contamination anbaanlon
C - 14 5.838-02 H/A H/A H/A R/A H/A H/A R/A H/A R/A H/A
Ml- 59 4.788-05
Sr- 90 2.268-02
Zr- 93 1.398-04
Te- 99
Sn-126
t -129
Ce-135
Ca-137
SB-15 1
Pb-210
Ra-226
Ra-228
Ac-227
Th-229
Th-230
Th-232
Pa-231
U -233
U -234
0 -235
0 -236
U -238
Hp-237
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Am-241
Av-243
Ck-245
Ca-246
.688-04
.258-02
.098-02
.768-03
.078-02
.788-06
.258-01
.686-01
.428-02
.878-02
.208-02
.258-01
.838-01
.308-01
.188-02
.988-02
.198-02
.878-02
.088-02
.668-02
.278-02
.208-02
.038-02
.188-03
.918-06 1.258-06 3.948-05 4.728-07
.728-03 1.048-04 . 1.738-02 1.198-03
.668-03 1.418-07 1.288-04 4.058-07
.028-05 7.708-06 2.028-04 8.388-05
.678-04 2.048-03 5.378-04 2.428-05
.158-03 2.658-04 6.758-02 .68E-03
.388-04 7.898-04 6.108-03 .718-04
.628-03
.528-06
.408-02
.418-02
.278-02
.728-02
. 128-02
.708-02
.538-02
.108-02
.628-03
.028-03
.568-03
.688-03
.328-03
.438-02
.438-02
.618-02
.608-02
.258-03
.018-02 2.488-02
.808-02 2.708-02
.818-02 2.698-02
.248-01 5.508-02
.008-02 2.748-02
.378-03 2.338-03 .428-04
.888-07 4.538-06 .138-09
.388-02 4.938-02 .268-04
.188-03 7.788-02 .418-03
.628-03 9.198-03 3.71E-04
.378-03 2.708-02 4.858-05
.538-04 1.308-02 4.978-06
.138-03 3.408-01 7.748-05
.178-03 1.898-01 3.888-05
.718-03 7.748-02 2.608-05
.698-04 1.448-02 3.108-04
.548-04 1.318-02 2.828-04
.678-04 1.438-02 3.078-04
.458-04 1.248-02 2.668-04
.618-04 1.388-02 2.968-04
.108-02 2.418-02 1.838-05
.968-04 1.758-02 1.578-07
.338-04 2.288-02 1.858-07
.318-04 2.168-02 1.808-07
.358-05 8.948-04 8.108-09
.078-04 2.238-02 1.788-07
.598-03 2.168-02 7.638-07
.368-03 2.408-02 8.288-07
.838-08 1.208-06 5.008-07 3.238-10 3.178-10 1.118-15
.398-06 2.308-06 3.848-06 4.058-09 0.008+00 0.008+00
.238-06 8.058-06 5.378-07 6.388-08 1.458-07 4.868-14
.388-06 1.738-06 1.448-06 4.678-11 0.008+00 1.808-19
.758-05 1.968-03 1.09E-04 6.478-08 7.55E-03 1.148-10
.318-04 7.82E-05 6.15E-05 3.688-08 S.41E-06 6.86E-13
.I6E-05 2.36E-05 2.46E-06 5.388-09 O.OOE+00 O.OOE+00
.658-05 2.07E-06 2.1SE-06 1.338-09 3.198-04 4.458-12
.978-08 5.23E-08 3.49E-07 2.148-09 0.008+00 1.318-17
.168-05 4.42E-03 2.46E-03 3.458-07
.038-05 4.05E-03 1.358-03 8.918-06
.63E-06 8.02E-05 2.67E-OS 5.618-07
.178-06 2.53E-04 1.69E-03 4.298-06
.158-07 2.03E-02 6. 76E-03 4.858-04
.808-06
.028-07
.018-06
.008-06
.828-06
.98E-06
.728-06
.918-06
.088-06
.108-08
.188-08
.998-08
.148-09
.908-08
.298-07
.418-07
.408-01 4.678-02 4.298-04
.248-02 1.088-02 6.278-04
.438-03 2.388-04 5.338-04
.638-04 2.71E-OS 7.418-06
.488-04 2.478-03 4.538-06
.628-04 2.70E-OS 5.468-06
.418-0* 2.348-05 4.298-06
.378-04 2.618-03 4.098-06
.038-03 1.008-03 3.408-06
.358-05 3.73E-04 1.148-05
.818-04 2.018-03 3.148-04
.578-04 1.758-03 2.758-04
.638-07 8.48E-06 8.738-08
.808-04 2.00E-03 3.138-04
.88-04 3.25E-03 3.858-05
.428-03 9.44E-03 7.928-05
.518-03 4.138-02 7.678-05 2.598-07 2.988-03 1.998-02 3.858-04
.758-03 2.038-02 3.798-05 1.298-07 1.318-03 8.758-0 1.758-04
.60E-08 6.138-15
.OOE-02 1.56E-10
.238-04 4.83E-12
.07E-04 2. 18E-12
.39E-03 2.25E-10
.25E-01 1.958-09
.348-01 2.908-09
.588-03 1.76E-10
.438-05 1.338-12
.638-07 1.298-14
.OOE-04 1.608-11
.418-09 1.018-14
.658-05 1.888-12
.838-05 1.558-12
.748-09 1.60E-15
.218-08 4.26E-14
.978-08 3.558-14
.468-09 1.68E-15
.958-08 3.628-14
.228-06 1.10E-12
.088-04 2.938-11
.49E-04 2.798-11
.118-08 1.708-14

-------
   Table 7.8-3.
              Development  of Release Limits presented  in Table  1
                        of 40 CFR Part 191
Radio-
nuclide
C-14 ,.,
Ni-59*0'
Sr-90,,»
Zr-93W;
Tc-99
Sn-126
1-129
Cs-135
Cs-137,,.
Sm-151,(d>
Pb-210VC;
Ra-226, v
Ra-228(e)
Ac-227je>
Th-229*e;
Th-230
Th-232
Pa-231Ve'
U-233
U-234
U-235
U-236
U-238
Np-237
Pu-238
Pu-239
Pu-240,..
Pu-241(d)
Pu-242
Am- 241
Am-243, ,
Cm-245:. v
Cm-246
Fatal cancers
per curie
released to , .
surface water
5.83E-02
4.78E-05
2.26E-02
1.59E-04
3.68E-04
1.25E-02
8.09E-02
7.76E-03
1.07E-02
9.78E-06
1.25E-01
1.68E-01
2.42E-02
6.87E-02
6.20E-02
7.25E-01
3.83E-01
1.50E-01
2.18E-02
1.98E-02
2.19E-02
1.87E-02
2.08E-02
8.66E-02
4.27E-02
5.20E-02
5.03E-02
2.18E-03
5.01E-02
5.80E-02
6.81E-02
1.24E-01
6.00E-02
Curies required
to cause 10,, .
fatal cancers '
172
209,000
442
62,900
27,200
800
124
1.290
935
1,020,000
80
60
413
145
161
14
26
67
459
505
457
535
480
115
234
192
199
4580
200
172
147
81
167
Release limit
per 1000 MTHM
or other unit
of waste*0'
(curies)
100,,.
l,000Cd)
1,000,,,.
l,000Cd)
10,000
1,000
100
1,000
1,000,.
i,ooo*d>
100 '
A W
100, v
100^e).
100«
100* '
10
* w
10, x
100(e)
100
* w
100
100
100
L \f W
100
100
100
A W
100
±\P\J
100 (.
100
x w
100
A W
100
iV/U * v
100
-------
 into 10 health effects  to  determine  the number of curies of that radio-
 nuclide that would cause  10 health effects  (shown in the third column of
 Table 7.8-3).   Then,  these estimates were rounded to the nearest order of
 magnitude  to reflect  the approximate nature of all of these calculations.
 (For example*  if a number  were between 100 and 1,000, it would generally
 be  rounded to  100 If  it were less than 316, the logarithmic midpoint, and
 to  1,000 if it were more than 316).  Several judgments were made in this
 rounding process.   First,  radlonuclldes present only in small quantities,
 or  which appear to be insignificant  to the overall risk were combined
 Into an "any other radionuclide" category.  There are two of these "any
 other radionuclide" categories, one  for alpha-emitting radlonuclides and
 one for non-alpha-emitting radionuclides.  Radlonuclides included in
 either of  these categories are Identified in Table 7.8-3.  Each category
 was assigned a release  limit.  Also, uncertainties in the long-term risk
 estimates  were considered  when rounding values up or down.  For example,
 the projected  curies  of strontium-90 and the various isotopes of uranium
 needed to  cause 10 health  effects are all about the same and are all near
 the midpoint of the rounding range.  However, the release limit for the
 relatively short-lived  strontium-90 was rounded up to 1,000 curies, while
 the release limits for  the very long-lived uranium isotopes (for which
 ultimate environmental  pathways will be more uncertain) were conserva-
 tively rounded down to  100 curies.

 7.9  Uncertainty Analysis

      Environmental pathway doslmetry and risk models generally employ an
 environmental  transport methodology consisting of multiplicative chain
 algorithms incorporating several variables. When regulatory analyses are
 performed, the tendency is to choose conservative values for these vari-
 ables due  to the inherent  uncertainty In the parameters.  The multi-
 plicative  nature of the models means that conservatism In chosing values
 for individual parameters  can lead to larger conservatism in the result.
 The  problem with this approach is that widespread conservatism can lead
 to  extremely conservative  and sometimes unrealistic results.

      The consideration  of  uncertainty in Individual parameter values used
 in  environmental pathway models has been a subject of discussion in the
 technical  community for more than a decade (Ba79a,b, Ho79, M179, Ru79,
 Sh79).   However, the  comprehensive consideration of overall uncertainty
 in  environmental pathway,  dosimetry, and health Impact analyses has begun
 to  be addressed only  recently (R183, Ru83).

      When  considering the  uncertainty in the input parameters associated
with  environmental  pathway calculations,  the most common procedure has
been  to  qualitatively consider the range of reported parameter values and
to use judgement to select  the "best" value to use for a particular
application.  More  recently, attempts have been made to statistically
analyze  the distribution of  data for Individual parameters and to choose
a mean or median as the "best" value for regulatory purposes.

      It appears  that the most systematic  mechanism for considering uncer-
tainty in multiplicative chain models would be to include a probability
distribution representing current uncertainty about parameter values in

                                   7-17

-------
the input data and to run a sufficient number of cases (with parameter
values for each case chosen by a suitable sampling procedure) such that
the distribution of results can be evaluated.  The results of this type
of analysis could be considered in choosing an appropriate aet of single-
valued parameters to apply for regulatory calculations.   Alternately, a
decision might be made to perform the regulatory calculations probabi-
listically, then choose limits for a standard (or to perform calculations
to see if a limit is met) at a specified confidence level.  We believe
the subject needs additional study to determine the most appropriate use
of uncertainty analysis for standards setting applications; however, it
is clear to us that a quantitative analysis of the uncertainties is most
useful for focusing on important uncertainties for more  intensive con-
•idsration.

     Host of the technical analyses discussed in this chapter were per-
formed prior to the increase in emphasis on uncertainty  in risk assess-
ment calculations.  In most cases, point valuea for each parameter were
chosen after review of the range of values reported in the literature and
were chosen to be nesr the mean or median value to avoid obtaining un-
realistlcally conservative results.  Baes has published  a very complete
review and analysis of parameters used to predict the transport of radlo-
nuclides through agricultural pathways (Ba84).  For most radtonuclides we
considered, the surface water release mode was dominant  and the food
pathways either dominated or were major pathways in determining the fatal
cancers psr curie released to surface water.  Default values from the
Baes report were ueed for many critical food pathway parameters, and Baes
states that these default values were chosen to be realistic rather than
highly conservative.  Since the default values used were based on Baes
recent extensive review of the literature, the analysis  is more realistic
than would be the case if other sources of data were used.

     In response to one of the recommendations of the SAB Subcommittee
that reviewed the technical basis for 40 CFR Part 191, the Agency asked
Che Invlrosphere Company to perform an uncertainty analytic Of the calcu-
lations that produced the estimates for fatal cancers per curie of radio-
activity released to surface water (Sm85).  Envirosphere reviewed the
surface water pathway models and identified the key parameters.  The
uncertainties in these parameters were characterized by probability
distributions that were propagated through the models using a simulation
technique, which produced uncertainty distributions in the estimates of
fatal cancers per unit of radionucllde release to surface water.

     An example of the results of the Envirosphere uncertainty analysis
is presented in Figure 7.9-1, which shows uncertainty distributions for
the fatal cancers risk for Am-243 releases to surface water.  Figure
7.9-1 shows a probability plot of the fatal cancers per 10.000 years per
curie of An-243 released to surface water.  This figure indicates s 75
percent probability, considering parameter uncertainties, that EPA's
release 11*it of  100 Ci for An-243 will result in 10 or less deaths over
10,000 years per 1000 MTHW.  Similar calculations were performed for
several other radionuclides listed in Table 7.8-3 (Sm85. EMV85).

                                  7-18

-------
*-•
VO
                                                10"       10 V     10"'        1

                                        FATAL CANCERS OVER 10,000 YEARS PER CURIE
10
                          Figure 7.9-1.   Probability distribution of  population  risks
                                per curie of Am-243 released to surface water.

-------
                                 REFERENCES
Ba79a     Baes C. F. Ill, The Soil Loss Constants  Due to Leaching from
          Soils, in  A Statistical Analysis of Selected Parameters for
          Predicting Food Chain Transport and Internal Doses of Radio-
          nuclides, F. 0. Hoffman and C. F. Baes III editors, Nuclear
          Regulatory Commission ORNL/NUREG/TM-282, pp. 85-92, 1979.

Ba79b     Baes C. F. Ill, Productivity of Agricultural Crops and Forage,
          Y , in  A Statistical Analysis of Selected Parameters for
          Predicting Food Chain Transport and Internal Doses of Radio-
          nuclides, F. 0. Hoffman and C. F. Baes III editors, Nuclear
          Regulatory Commission, ORNL/NUREG/TM-282, pp. 15-29 1979.

Ba84      Baes C. F. Ill, Sharp R. D., Sjoreen A. L., and Shor R. W.  A
          Review and Analysis of Parameters for Assessing Transport of
          Environmentally Released Radionuclides through Agriculture, Oak
          Ridge National Laboratory, ORNL-5786, 1984,

Be76      Bennett B. G., Transuranic Element Pathways to Man in Trans-
          uranic Nuclides in the Environment, IAEA, Vienna, Austria
          IAEA-SM-199/40, pp. 367-381, 1976.

B182      Blaylock B. G.» Frank M. L., and DeAngelis D. L., Bioaccumu-
          lation of Tc-95m in Fish and Snails, Health Physics, Vol  42
          Ko. 3, pp. 257-266, March 1982,

ENV85     Envirosphere Company, Revised Uncertainty Analysis of the EPA
          River Mode Pathways Model Used as the Basis for 40 CFR  191
          Release Limits, 1985.

EPA85     U.S. Environmental Protection Agency, Environmental
          Radiation Protection Standards for Management and Disposal of
          Spent Nuclear Fuel, High-Level and Transuranic Radioactive
          Wastes, 40 CFR Part  191, Federal Register,  to be published.

Fo76      Fowler T. W., Clark R. L., Gruhlke J. M., and Russell J. L.
          Public Health Considerations of Carbon-14 Discharges from the
          Light-Water-Cooled Nuclear Power Reactor Industry, USEPA T
          nical Note ORP/TAD-76-3, Washington, DC, 1976.

Fo79      Fowler T. W. and Nelson C. B., Health Impact Assessment  of
          Carbon-14 Emissions from Normal Operations  of Uranium Fuel
          Cycle Facilities, EPA 520/5-80-004, Montgomery, AL  1979
                                    7-20

-------
Ho79     Hoffman F. 0., Bioaccumulatlon Factors for Freshwater Fish, B. ,
         in  A Statistical Analysis of Selected Parameters for Predicting
         Food Chain Transport and Internal Doses of Radionuclides, F. 0.
         Hoffman and C. F. Baes III editors, Nuclear Regulatory Commis-
         sion, ORNL/NUREG/TM-282, pp. 96-108, 1979

ICRP75   International Commission on Radiological Protection, ICRP
         Publication 23:  Report of the Task Group on Reference Man,
         Elmsford, NY: Pergamon Press, 1975.

Ki77     Killough G. G., A Diffusion-Type Model of the Global
         Carbon Cycle for the Estimation of Dose to the World Population
         from Releases of Carbon-14 to the Atmosphere, ERDA, ORNL-5269,
         Oak Ridge National Laboratory, 1977.

Mi79     Miller C. W., The Interception Fraction, in  A Statistical
         Analysis of Selected Parameters for Predicting Food Chain Trans-
         port and  Internal Doses of Radionuclides, F. 0.  Hoffman and
         C. F. Baes III editors,  Nuclear Regulatory Commission, ORNL/
         NUREG/TM-282, pp. 31-42,  1979.

Mo79     Moore R.  E., Baes C. F. Ill, McDowell-Boyer L. M., Watson A. P.,
         Hoffman F. 0., Pleasant J. C., and Miller C. W., AIRDOS - EPA:
         A Computerized Methodology for Estimating Environmental Concen-
         trations  and Doses  to Man from Airborne  Releases of Radionuclides,
         Oak Ridge National  Laboratory, EPA 520/1-79-009, 1979.

Mu77     Murray C. R.  and Reeves E. B., Estimated Use of  Water  in  the
         United States  in 1975, U.S.  Department of  Interior USGS Circular
          765, Arlington,  VA, 1977.

Ne78     Nelson C. B.,  Davis R., and  Fowler T.  W., A Model to Assess
          Population  Inhalation Exposure  from a Transuranium Element  Con-
          taminated Land Area, in Selected Topics:   Transuranium Elements
          in the  General Environment,  EPA Technical  Note ORP/CSD-78-1, pp.
          213-280,  Washington, 1978.

 Ri83     Rish W.  R.,  Schaffer S.  A.,  and Mauro J. J.,  Uncertainty and
          Sensitivity Analysis of the Exposure Pathways Model Used as the
          Basis for Draft 40 CFR191,  Supplementary Report to National
          Waste Terminal Storage Technical Support Team - USDOE, New York,
          November 1983.

 Ru79     Rupp E.  M., Annual Dietary Intake and Respiration Rates, U  , in
          A Statistical Analysis of Selected Parameters for Predicting
          Food Chain Transport and Internal Doses of Radionuclides, F. 0.
          Hoffman and C. F. Baes III editors, Nuclear Regulatory Commis-
          sion ORNL/NUREG/TM-282, pp. 109-132,  1979.

 Ru83     Runkle G. E., Calculation of Health Effects per Curie Release
          for Comparison with the EPA Standard, in  Technical Assistance
          for Regulatory Development:  Review and Evaluation of the  Draft
          EPA Standard 40CFR191 for Disposal of High-Level Waste, Nuclear
          Regulatory Commission NUREG/CR-3235, Vol. 6, 1983.
                                     7-21

-------
Sc83       Schaffer S, A., Envirosphere Company Personal Communication to
           J. M. Smith - EPA, September 21,  1983.

Sh79       Shot R, W. and Fields D. E., Animal Feed Consumption Rate, Q
           in  A Statistical Analysis of Selected Parameters for Predicling
           Food Chain Transport and Internal Doses of Radionuclides, F  0
           Hoffman and C. F. Baes  III editors, Nuclear Regulatory Commis-
           sion ORNL/NUREG/TM-282, pp. 51-58, 1979.

Sm85       Smith J. M., Fowler T.  W., and Goldin A. S., Environmental
           Pathway Models for Estimating Population Health Effects from
           Disposal of High-Level  Radioactive Waste in Geologic Reposi-
           tories, EPA 520/5-85-026, Montgomery, Alabama, 1985.

Th72       Thompson S. E., Burton  C. A., Quinn D. J.t and Ng Y. C.
           Concentration Factors of Chemical Elements in Edible Aquatic
           Organisms, Lawrence Livermore Laboratory, U.S. Atomic Enerav
           Commission UCRL-50564/Rev. 1, 1972.                       8y

UNSCEAR77  United Nations Scientific Committee on the Effects of Atomic
           Radiation, Sources and  Effects of Ionizing Radiation:  UNSCEAR
           1977 Report, United Nations Publication Sales No. E 77 IX 1
           1977.                                                    ' '

Wo79       Newspaper Enterprise Association, Inc., The World Almanac and
           Book of Facts 1979, New York, 1978.
                                    7-22

-------
              APPENDIX A

A DESCRIPTION OF THE RADRISK AND CAIRD
 COMPUTER CODES USED BY EPA TO ASSESS
DOSES AND RISKS FROM RADIATION EXPOSURE
                A-l

-------
A. 1  Introduction
     This appendix provides a brief overview of the RADRISK (Du80) and
CAIRD (Co78) computer codes used by the Environmental Protection Agency
to assess the health risk from radiation exposures.  It describes how the
basic dose calculations are performed and describes the mechanics of the
life table implementation of the risk estimates derived in Chapter 6.

A. 2  Overview of the EPA Analysis

     RADRISK, the computer code used to calculate dose and risk, calcu-
lates the radiation dose and risk resulting from an annual unit intake of
a given radionuclide or the risk resulting from external exp"olure to a
unit concentration of radionuclide in air or on ground surface (Du80 84
Su81).  Since both dose and risk models are linear, the unit dose and  '
risk results can then be scaled to reflect the exposure associated with a
specific source.

     As outlined in Chapter 5, estimates of the annual dose rate to
organs and tissues of interest are calculated by using, primarily, models
recommended by the International Commission on Radiological Protection
(ICRP79,80).  Because EPA usually considers lifetime exposures to a
general population, these dose rates are used In conjunction with a life
table analysis of the increased risk of cancer resulting from radiation
(Co78).  This analysis, described below, takes account of competing risks
and the age of the population at risk.                      «p«ing TISKS

A. 3  Dose Rates from Internal Exposures

     Internal exposures occur when radioactive material is inhaled or
ingested.  The RADRISK code implements contemporary dosimetric models to
estimate the dose rates at various times to specified reference oreans in
the body from inhaled or ingested radionuclides.  The dosimetric methods
in RADRISK are adapted from those of the INREM-II code which is based on
models recommended by the International Commission on Radioloeir*! «,.«*..
tion  (K178, ICRP79).  The principal qualitative difference is that
comutes dose rates to specified organs searatel          -
          ,        .                quaave  ifference is that RAD
computes dose rates to specified organs separately for high- and low
linear energy transfer (LET) radiations, while INREM-II calculates the
committed dose equivalent to specified organs.  The time-deoendeni- rfnB»
rates are used in the life table calculations of RADRISK.

     In RADRISK, the direct intake of each radionuclide is treated *ena
rately.  For decay chains, the ingrowth and dynamics of decay DrorWr*
(daughters) in the body after intake of a parent radfcmuoliZ are consid
ered explicitly in the calculation of dose rate.  The decay product
contributions to the dose rate are included in the dose calculating
based on the metabolic properties of the element and the organ in which
they occur.
     The dose rate D  (X;t) to target organ X at time t due to rad-i^nri
  (l
-------
deposited annually in a given mass of tissue as a result of radioactive
decay, and is computed as:
                            .         m  ,
                            D,(X;t)-2  D,(X<-Y.;t)          (A-l)
where


                                                             
-------
proceed through the small intestine, upper large intestine, and lower
large intestine; radionuclides may be absorbed by the bloodstream from
any of these four segments, although only absorption from the small
intestine is considered in this study.
     The activity,  A4ir^t^» of radionuclide i in organ k may be divided
among several "pools" or "compartments," denoted here by the subscript
Each differential equation describing the rate of change of activity
within a compartment is a special case of the equation:


                                  1-1      Ljk
                                                            (A-3)

where
- activity of radionuclide i in compartment i of organ k,

  number of exponential terms in the retei
  radionuclide j (j«l to i-1) in organ k,
     L .  • number of exponential terms in the retention function for
     B..  « branching ratio of radionuclide j to radionuclide i,


     XR                           -1
      i  -  rate coefficient  (time  ) for radiological decay of
            radionuclide i,
      B                           — 1
     X... » rate coefficient  (time  ) for biological removal of
            radionuclide i from compartment £ of organ k,

     c  ,. - fractional coefficient for radionuclide i in the £,-th compart-
            ment of  organ k,

     P .  - inflow rate of radionuclide  i into organ k.


     If  the inflow rate P „ remains constant, the  equations may be  solved
explicitly for Aik
-------
In this model, shown in Chapter 5, there are four major regions:  the
nasopharyngeal, tracheobronchial, pulmonary, and lymphatic tissues.  A
fraction of the inhaled radioactive material is initially deposited in
each of the nasopharyngeal, tracheobronchial, and pulmonary regions.  The
material is then cleared (removed) from the lung to the blood and the
gastrointestinal tract, also as shown in Chapter 5.  Deposition and
clearance of inspired particulates in the lung are controlled by the
particle size and solubility classification.

     The size distribution of the particles is specified by the activity
median aerodynamic diameter (AMAD); where no AMAD is known, a value of
1.0 micron is assumed.  The model employs three solubility classes, based
on the chemical properties of the radionuclide; classes D, W, and Y
correspond to rapid (days), intermediate (weeks), and slow (years) clear-
ance, respectively, of material deposited in the respiratory passages.
Inhaled nonreactive, i.e., noble, gases are handled as a special case.

     Movement of activity through the gastrointestinal (GI) tract is
simulated with a catenary model, consisting of four segments;  stomach,
small intestine, upper large intestine, and lower large intestine.
Exponential outflow of activity from each segment into the next or out of
the system is assumed.  Outflow rate constants are calculated from the
transit times of Eve (Ev66).  Although absorption may occur from any
combination of the four segments, only activity absorbed into the blood
from the small intestine is normally considered; the fractional absorption
from the small intestine into the blood is traditionally denoted f..

     Activity absorbed by the blood from the GI or respiratory tract is
assumed to be distributed immediately to systemic organs.  The distribu-
tion of activity to these organs is specified by fractional uptake coeffi-
cients.  The list of organs in which activity is explicitly distributed
(termed source organs) is element-dependent, and may include such organs
as bone or liver where sufficient metabolic data are available.  This
list is complemented by an additional source region denoted as OTHER,
which accounts for that systemic activity not distributed among the
explicit source organs; uniform distribution of this remaining activity
within OTHER is assumed.

     Radioactive material that enters an organ may be removed by both
radioactive decay and biological removal processes.  For each source
organ, the fraction of the initial activity remaining at any time after
intake is described by a retention function consisting of one or more
exponentially decaying terms.

     The metabolic models and parameters employed in the present study
have been described by Sullivan et al. (Su81),   In most cases, the models
are similar or identical to those recently recommended by the ICRP
(ICRP79,80,81).   However, some differences in model parameters do exist
for some radionuclides (Su81).  In particular,  parameter values that are
thought to be more representative of metabolism following low-level
environmental exposures, rather than occupational exposures,  have been
used in this analysis [e.g., fi-0.2 for uranium in the environment (ICRP79,


                                   A-5

-------
NAS83)].  For transuranic Isotopes, metabolic parameters from EPA77,
related comments from EPA78 and from the National Radiological Protection
Board (Ha82), have been used rather than those from ICtPSO.  These param-
eters are listed in Table A.3-1.

     The EPA values were recommended by O.S. experts on transuranic
element matabollsn at Battelle Pacific Northwest Laboratory (EPA78).  The
recently-adopted National Radiation Protection Board fi values for trans-
uranics In the general environment are closer to the values proposed by
EPA In 1977 than those currently advocated by ICRP for occupational
exposures.  The larger f\ values will increase the estimated dose and
risk from ingestlon of transuranic materials but have little effect on
doses following inhalation.

A.4  Dose Rates from External Exposures

     Because of the penetrating nature of photons, radioactivity need not
be taken into the body to deliver a dose to body organs.  Energy absorbed
from photons emitted by radionuclides In the air or on the ground surface
may also contribute to the overall risk.  Natural background radiation is
an example of an important external exposure, ordinarily contributing the
largest component of dose to people.

     Organ doae rates to an Individual immersed in contaminated air or
standing on a contaminated ground surface are computed by Keener*s
DOSFACTOR computer code (Ko81).  These calculations assume that the
radlonuclide concentration is uniform throughout an infinite volume of
air or area of ground surface, and that the exposed Individual ia stand-
ing on the ground surface.  Only photons penetrate the body sufficiently
to deliver a significant dose to internal organs, and only doses from
photon radiation are considered in this analysis.  Beta radiation la far
lass penetrating and delivers a dose only to the body surface; because
skin is not a target tissue of concern in this analysis, no consideration
of beta contributions to dose is required.  Alpha particles have even
less penetration ability, and are also excluded from consideration here.
                                 •y
     The photon dose rate factor Dj(X) for a given target organ, X, of
an individual immersed in contaminated air at any time may be expressed
as:
                                    fY


C*        
-------
    Table A.3-1.  Small Intestine to Blood Transfer Fractions, fj, for
                         Transuranic Elements
Element
Isotope
Plutoniua-238 and 241
Oxide form
Nonoxide form
Bio. inc.(a)
Plutonium-239 and 240
Oxide form
Nonoxide form
Bio. inc.
Americiua
Oxide fora
Nonoxide fora
Bio. inc.
Curium
Oxide fora
Nonoxide fora
Bio. inc.
Neptunium
EPA
Child
0-12 mo

io-2
io-2
5xlO"2

io-3
12-2
5xlO*2

io-2
io-2
5xlO"2

lO'2
io-2
5xlO"2
-
Adult
>12 mo

io-3
io-3
5xlO"3

io-4
I--3
5xlO"3

ID'3
io-3
5xlO"3

io-3
io-3
5xlO'3
io-3
Adult

to-5(b)
5xlQ-4
SxlO"4

io-5(b>
5xlO"4
5xlO"4

SxlO"4
5xlO"4
5xlO"4

5x10"*
SxlO*4
5x10"*
io-3
HRPB
Child
0-12 mo

5,1010-4(b)
5xlO~3
5xlO'4

5,10-4(b)
5x1 0~3
SxlO"3

5xlO"3
5xlO"3
SxlO"3

5xlO"3
5xlO"3
5xlO"3
5xlO"3


0-3 mo

io-3(b)
10-2
io-2

io-3(b)
1C'2
io-2

io-2
1C'2
io-2

to'2
io-2
io-2
io-2
;*» Biologically incorporated for*.
"' Hydroxide fora,

Source:  EPA77, EPA78, Ha82.
NRPB:    National Radiological Protection Board,
                                   A-7

-------
where
      p   - density of air,

      pm  - 0.5, the particle-medium correction factor,


     f    « intensity of n   discrete photon  (number/disintegration),

      V                th
     ET   » energy of n   photon,


     ji/p  - photon mass energy absorption coefficient, with subscripts
            "t" and "a" denoting tissue and air, respectively, for
            photons of energy E ,

     G    » ratio of absorbed dose in organ X to absorbed dose at the
            body surface.

     c    - unit conversion proportionality contrast.

     The terms p/p and G  are functions of photon energy, E\

                                  •y
     The photon dose rate factor  D' (X) to organ X of an individual at a
distance z above a unit concentration contaminated ground surface may be
computed as:


                                      B  I fj E^[(p/p)t]n
                                 x /  1/r exp(-u  r)dr
                                    z           flu


                                 -tCan/(Dan" ^^
where
     K    "1.0, the particle-material correction factor,
      pm

     yan  • mass attenuation coefficient for the nC  discrete proton,

     z    « height of reference position above ground surface (taken to
            be 1 meter in this study).

     c    • unit conversion proportionality constant.
                                   A-8

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     The coefficients C   and D   are functions of the photon energy.
For detailed discussionaof the tirivation of these equations and a tabu-
lation of dose rate factors for various radionuclides, see Kocher (Ko79,
Ko81).

     In the analysis here, the dose rate factors described by these
equations are scaled to achieve a continuous exposure of 1 pCi/cm3 for
air immersion and 1 pCi/cm2 for ground surface exposure.  Risk estimates
for these exposure pathways are based on continuous lifetime exposure to
these levels.

A.5  Life Table Analysis to Estimate the Risk of Excess Cancer

     Radiation effects can be classified as stochastic or nonstochastic
(NAS80, ICRP77).  For stochastic effects, the probability of occurrence
of the effect, as opposed to the severity, is a function of dose; induc-
tion of cancer, for example, is considered a stochastic effect.  Nonsto-
chastic effects are those health effects for which the severity of the
effect is a function of dose; examples of nonstochastic effects include
cell killing, suppression of cell division, cataracts, and nonmalignant
skin damage.

     At the low levels of radiation exposure attributed to radionuclides
in the environment, the principal health detriment is the induction of
cancers (solid tumors and leukemia), and the expression, in later genera-
tions, of genetic effects.  In order to estimate these effects, instanta-
neous dose rates for each organ at specified times are sent to a subrou-
tine adaptation of CAIRD contained in the RADRISK code.  This subroutine
uses annual doses derived from the transmitted dose rates to estimate the
number of incremental fatalities in the cohort due to radiation-induced
cancer in the reference organ.  The calculation of incremental fatalities
is based on estimated annual incremental risks, computed from annual
doses to the organ, together with radiation risk factors such as  those
given in the 1980 NAS report BEIR-3 (NAS80).  Derivation of the risk
factors in current use is discussed in Chapter 6.

     An important feature of this methodology is the  use of actuarial
life tables to account for the time dependence of the radiation  insult
and to allow for competing risks of death in the estimation of risk due
to radiation exposure.  A life table consists of data describing  age-spe-
cific mortality rates from all causes of death for a  given population.
This information is derived from data obtained on actual mortality rates
in a real population; mortality data for the U.S. population during the
years 1969-1971 are used throughout this study  (HEW75).

     The use of life tables in studies of risk due to low-level radiation
exposure is important because of the time delay inherent in radiation
risk.  After a radiation dose is received,  there is a minimum induction
period of several years  (latency period) before a cancer is clinically
observed.  Following the  latency period, the probability of occurrence of
a cancer during a given year  is assumed  to  be constant  for a specified
                                    A-9

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period, called a plateau period.  The length of both the latency and
plateau periods depends upon the type of cancer.

     During or after radiation exposure, a potential cancer victim may
experience years of life in which he is continually exposed to risk of
death from causes other than incremental radiation exposure.  Hence, some
individuals in the population will die from competing causes of death,
and are not potential victims of incremental radiation-induced cancer.

     Each member of the hypothetical cohort is assumed to be exposed to a
specified activity of a given radionuclide.  In this analysis each member
of the cohort annually inhales or ingests 1 pCi of the radionuclide, or
is exposed to a constant external concentration of 1 pCi/cm3 in air or
1 pCi/cm2 on ground surfaces.  Since the models used in RADRISK are
linear, these results may be scaled to evaluate other exposure conditions.
The cohort consists of an initial population of 100,000 persons, all of
whom are simultaneously liveborn.  In the scenario employed here, the
radiation exposure is assumed to begin at birth and continue throughout
the entire lifetime of each individual.

     No member of the cohort lives more than 110 years.  The span from
0 to 110 years is divided into nine age intervals, and dose rates to
specified organs at the midpoints of the age intervals are used as esti-
mates of the annual dose during the age interval.  For a given organ, the
incremental probability of death due to radiation-induced cancer is
estimated for each year using radiation risk factors and the calculated
doses during that year and relevant preceding years.  The Incremental
probabilities of death are used in conjunction with the actuarial life
tables to estimate the incremental number of radiation induced deaths
each year.

     The estimation of the number of premature deaths proceeds in the
following manner.  At the beginning of each year, m, there is a probabil-
ity P  of dying during that year from nonradiological causes, as calcu-
Igted from the life table data, and an estimated incremental probability
P  of dying during that year due to radiation-induced cancer of the given
organ.  In general, for the m-th year, the calculations are:

     M(m)   - total number of deaths in cohort during year m,

            • [PN(m) + PR(m)J x N(m)

     Q(m)   • incremental number of deaths during year m due to
              radiation-induced cancer of a given organ,

            - PR(m) x N(m)

     N(m+l) • number of survivors at the beginning of year m + 1

            - N(m) - M(m)
                                   A-10

-------
 R                                     N
P  is assumed to be small relative to P , an assumption which is reason-
able only for low-level exposures, such as those considered here (Bu81).
The total number of incremental deaths for the cohort is then obtained by
summing Q(m) over all organs for 110 years.

     In addition to providing an estimate of the incremental number of
deaths, the life table methodology can be used to estimate the total
number of years of life lost to those dying of radiation-induced cancer,
the average number of years of life lost per incremental mortality* and
the decrease in the population's life expectancy.  The total number of
years of life lost to those dying of radiation-induced cancer is computed
as the difference between the total number of years of life lived by the
cohort assuming no incremental radiation risk, and the total number of
years of life lived by the same cohort assuming the incremental risk from
radiation.  The decrease in the population's life expectancy can be
calculated as the total years of life lost divided by the original cohort
size (N(l)-lOO.OOO).

     Either absolute or relative risk factors can be used.  Absolute risk
factors, given in terms of deaths per unit dose, are based on the assump-
tion that there is some absolute number of deaths in a population exposed
at a given age per unit of dose.  Relative risk factors, the percentage
increase in the ambient cancer death rate per unit dose, are based on the
assumption that the annual rate of radiation-induced excess cancer deaths,
due to a specific type of cancer, is proportional to the ambient rate of
occurrence of fatal cancers of that type.  Either the absolute or the
relative risk factor is assumed to apply uniformly during a plateau
period, beginning at the end of the latent period.

     The estimates of incremental deaths in the cohort from chronic
exposure are Identically those which are obtained if a corresponding
stationary population (i.e., a population in which equal numbers of
persons are born and die in each year) is subjected to an acute radiation
dose of the same magnitude.  Since the total person-years lived by the
cohort in this study is approximately 7.07 million, the estimates of
incremental mortality in the cohort from chronic irradiation  also apply
to a one-year dose of the same magnitude to a population of this size,
age distribution, and age-specific mortality rates.  More precise  life
table estimates for a specific population can be obtained by  altering  the
structure of the cohort to reflect the age distribution of a  particular
population at risk.

A. 6  Risk Analysis Methodology

     Risk estimates in current use at EPA are based on the  1980 report
(BEIR-3) of the National Academy  of Sciences Advisory Committee on the
Biological Effects of Ionizing Radiation (HAS80).  The form of these  risk
estimates is, to some extent, dictated by practical considerations, e.g.,
a desire to limit the number of cases which must be processed for  each
environmental analysis and a need to conform to  limitations of  the comput-
er codes in use.  For example, rather than analyze male and female popula-
tions separately, the risk estimates have been merged for use with the
                                    A-ll

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 general population;  rather than perform both an absolute and a relative
 risk calculation,  average values have been used.

      The derivation of the risk estimates from the BEIR-3 report is
 presented in Chapter 6.  A brief outline of the general procedure is
 Bummarized below.   Tables referenced from Chapter V of NAS80 are desig-
 nated by a V prefix.

      (1)  The total number of premature cancer fatalities from lifetime
 exposure to 1 rad  per year of low-LET radiation is constrained to be
 equal to the arithmetic average (280 per million person-rad) of the
 absolute and relative risk values (158 and 403 per million person-rad)
 given in Table V-25  of the BEIR-3 report for the L-L and L-L models for
 leukemia and solid cancers, respectively (NAS80).

      (2)  For cancers other than leukemia and bone cancer, the age and
 sex-specific incidence estimates given in Table V-14 were multiplied by
 the mortality/incidence ratios of Table V-15 and processed through the
 life table code at constant,  lifetime dose rates of 1 rad per year.  The
 resulting deaths-are  averaged, using the male/female birth ratio, and
 proportioned for deaths due to cancer in a specific organ as described  in
 Chapter 6.  These  proportional risks are then used to allocate the organ
 risks among the 235.5 deaths  per million person-rad remaining after the
 44.5 leukemia and  bone cancer fatalities (Table V-17) are subtracted from
 the arithmetic average of 280 given  in Table V-25.

      (3)  The RADR1SK code calculates dose rates for high- and low-LET
 radiations independently.   A  quality factor of 20 has been applied to all
 alpha doses to obtain the organ dose equivalent rates in rem per year
 (ICRP77).   For high-LET radiation risk estimates,  the risk from alpha
 particles is considered to be eight  times that for low-LET radiation to
 the same tissue except for bone cancer,  for which the risk coefficient  is
 twenty times the low-LET value.   Additional discussion was included in
 Chapter 6.

      A typical environmental  analysis requires that  a large number of
 radionuclides and  multiple exposure  models be considered.   The RADRISK
 code has been used to obtain  estimates of cancer risk for unit intakes  of
 approximately 200  radionuclides  and  unit  external  exposures by approxi-
 mately 500 radionuclides.   The calculated dose rates and mortality coeffi-
 cients described in the preceding sections are processed through the life
 table subroutine of the RADRISK code to obtain lifetime risk estimates.
 At  the low levels  of  contamination normally encountered in the environ-
 ment,  the  life table  population  is not appreciably perturbed by the
 excess radiation deaths  calculated and,  since both  the  dose and risk
 models  are linear, the  unit exposure results  may be  scaled to reflect
 excess  cancers  due to  the  radionuclide concentrations predicted  in the
 analysis of  a  specific  source.

     As noted  in the discussion  of the life table analysis,  risk estimates
 for chronic  irradiation  of  the cohort  may  also be applied  to  a stationary
population havingT:he same age-specific mortality rates as the 1970  U.S.
                                   A-12

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population.  This is, since the stationary population is formed by super-
position of all age groups in the cohort* each age group corresponds to a
segment of the stationary population with the total population equal to
the sum of all the age groups.  Therefore, the number of excess fatal
cancers calculated for lifetime exposure of the cohort at a constant dose
rate would be numerically equal to that calculated for the stationary
population exposed to an annual dose of the same magnitude.  Thus, the
risk estimates may be reported as a lifetime risk (the cohort Interpreta-
tion) or as the risk ensuing from an annual exposure to the stationary
population.  This equivalence is particularly useful in analyzing acute
population exposures.  For example, estimates for a stationary population
exposed to annual doses which vary from year to year may be obtained by
summing the results of a series of cohort calculations at various annual
dose rates.
                                   A-13

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                               REFERENCES
Bu81      Bunger B, M., Cook J. R., and K. K. Barrick, Life Table Method-
          ology for Evaluating Radiation Risk:  An Application Based on
          Occupational Exposures, Health Fhys. 40, 439-455.

Co78      Cook J. R., Bunger B., and H. K. Barrick, A Computer Code for
          Cohort Analysis of Increased Risks of Death (CAIRO), EPA
          520/4-78-012, 1978.

Du80      Dunning D. E. Jr., Leggett R. W., and M. G. Yalcintas, A Coin-
          Dined Methodology for Estimating Dose Rates and Health Effects
          from Exposure to Radioactive Pollutants, ORNL/TM-71D5, 1980.

Du84      Dunning D. E. Jr., Leggett R. V., and R. E. Sullivan, An Assess-
          ment of Health Risk from Radiation Exposures, Health Physics*
          46. (5), May 1984.

EPA77     U.S. Environmental Protection Agency, Proposed Guidance on Dose
          Limits for Persons Exposed to Transuranium Elements in the
          General Environment, EPA 520/4-77-016, 1977.

EPA78     U.S. Environmental Protection Agency, Response to Comments:
          Guidance on Dose Limits for Persons Exposed to Transuranium
          Elements in the General Environment, EPA 520/4-78-010, 1978.

Ev66      Eve 1. S., A Review of the Physiology of the Gastrointestinal
          Tract in Relation to Radiation Doses from Radioactive Materials,
          Health Physics, \2t 131-162, 1966.

Ha82      Harrison, J. D., Gut Uptake Factors for Plutonium, Americium
          and Curium, NRPB-R129, National Radiological Protection Board,
          NRPB-R129, HMSO, P. 0. Box 569, London, January 1982.

HEW75     U.S. Department of Health, Education and Welfare, 1975, U.S.
          Decennial Life Tables for 1969-1971, Vol. 1, No.  1, DHEW Publi-
          cation No. (HRA) 75-1150,  Public Health Service,  Health Resources
          Administration, National Center for Health Statistics, Rockville,
          Maryland.

ICRP72    International Commission on Radiological Protection, The
          Metabolism of Compounds of Plutonium and Other Actinides, ICRP
          Publication 19, Pergamon Press, 1972.
                                   A-14

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1CRP77    International Commission on Radiological Protection,  1977,
          Recommendations of the International Commission on Radiological
          Protection, Ann. ICRP, Vol. 1,  No.  1, Pergamon Press,  1977.

ICRP79    International Commission on Radiological Protection,  Limits  for
          Intakes of Radionuclides by Workers, ICRP Publication 30,  .
          Part 1, Annals of the ICRP, 2_ (3/4), Pergamon Press,  1979.

ICRP80    International Commission on Radiological Protection,  Limits  for
          Intakes of Radionuclides by Workers, ICRP Publication 30,
          Part 2, Annals of the ICRP, £ (3/4), Pergamon Press,  1980.

ICRP81    International Commission on Radiological Protection,  Limits  for
          Intakes of Radionuclides by Workers, ICRP Publication 30,
          Part 3, Annals of the ICRP, £ (2/3), Pergamon Press,  1981.

K178      Killough 6. G., Dunning D. E. Jr.,  and Pleasant J. C., INREM-II:
          A Computer Implementation of Recent Models for Estimating the
          Dose Equivalent to Organs of Man from an Inhaled or Ingested
          Radlonucllde, ORNL/NUREG/TM-84, 1978.

Ko79      Kocher D. C., Dose-Rate Conversion Factors for External Exposure
          to Photon and Electron Radiation from Radionuclides Occurring
          in Routine Releases from Nuclear Fuel-Cycle Facilities, ORNL/
          NUREG/TM-283, 1979.

Ko81      Kocher D. C., Dose-Rate Conversion Factors for External Exposure
          to Photon and Electron Radiation from Radionuclides Occurring
          in Routine Releases from Nuclear Fuel-Cycle Facilities, Health
          Physics, 38, 543-621, 1981.

Mo66      Morrow P. E., Bates D. V., Fish B. R., Hatch I. F., and Mercer
          T. T., Deposition and Retention Models for Internal Dosimetry
          of the Human Respiratory Tract, Health Physics, 12,  173-207,
          1966.                                           "~

HAS72     National Academy of Sciences - National Research Council* The
          Effects on Populations of  Exposures  to Low Levels of  Ionising
          Radiation, Report of  the Committee on the Biological  Effect* of
          Ionizing Radiations  (BEIR  Report), Washington, D.C.,  1972.

NAS80     National Academy of Sciences - National Research Council) The
          Effects on Populations of  Exposures  to Low Levels of  Ionising
          Radiation, Committee  on the Biological Effects of Ionising
          Radiations  (BEIR Report),  Washington, D.C.,  1980.

NAS83     National Academy of Sciences - National Research Council,
          Drinking Water  and Health, Vol. 5, Safe Drinking Water Committee,
          Washington, D.C.,  1983.
                                   A-15

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Sn74      Snyder W. 8., Fort M. R., Warner G. G., and Watson S. B., A
          Tabulation of Dose Equivalent per Microcurie-Day for Source and
          Target Organs of an Adult for Various Radionuclides, ORNL-5000,
          1974.

Su81      Sullivan R. E.t Nelson N. S., Ellett W. H., Dunning D. E. Jr.,
          Leggett R. W., Talcintas M. G., and Eckennan K. F.. Estimates
          of Health Risk form Exposure to Radioactive Pollutants, ORNL/TM-
          7745, 1981.
                                  A-16

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     APPENDIX B




GLOSSARY AMD ACRONYMS
          B-l

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                                 GLOSSARY
 actinide:
 alpha  particle:
 beta particle:
 contact-handled
 TRU wastes:
 critical organ:
 Ci:
The series of elements beginning with actinium,
Element No. 89, and continuing through lawrenclum,
Element No. 103.

Positively charged particle emitted by certain radio-
active materials.  It is made up of two neutrons and
two protons, Identical to the nucleus of a helium
atom.  It is the least penetrating type of radiation.

An elementary particle emitted from a nucleus during
radioactive decay, with a single electrical charge
and a mass equal to 1/1837 that of a proton.  A
negatively-charged beta particle is identical to an
electron.  A positively-charged beta particle is
called a positron.
TRU wastes that can be handled with Just the
shielding that is provided by the waste package
Itself.

Specific organ being most susceptible to the effects
of a specific type of radiation.

Curie - the unit rate of radioactive decay; the
quantity of any radionuclide which undergoes 3.7 x
1010 disintegrations/second.  Several fractions of
the curie are in common usage:

     Nanocurie (nCi) - one-billionth of a curie; 3.7
     x 101 disintegrations/second
                         Microcurie
                         curie
          curie Cud) - one-millionth of a
          ; 3.7 x 10* disintegrations/second
daughter:
     Millicurie (nCi) - one-thousandth of a curie;
     3.7 x 107 disintegrations/second

     Picocurie (pCi) - one-millionth of a microcurie;
     3.7 x 10~2 disintegrations/second or 2.22 disintegra-
     tions /minute

Synonym for decay product.
                                   B-2

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decay product:
dose:
dose equivalent:
doslmetry:
effective half
life  (t):
 fissile:
 fission:
 fission products:
 fuel cycle:
 gamma ray:
A nuclide resulting from the radioactive disintegra-
tion of a radionuclide, being formed either directly
or as the result of successive transformations in a
radioactive series.  Also called a daughter.  Decay
products may be stable or radioactive.

The amount of energy absorbed per gram of absorbing
tissue as a result of the exposure.

A term used to express the amount of effective
radiation when modifying factors have been con-
sidered; the product of absorbed dose multiplied by a
quality factor multiplied by a distribution factor.
It is expressed numerically in rents.

Quantification of energy absorbed by the population
from decaying radionuclides.
The time required for one-half of a radioactive
material originally present in the body to be removed
by biological clearance or radioactive decay.

Any material fissionable by neutrons of all  energies*
including  thermal  (slow) neutrons as well as fast
neutrons.

The splitting of a heavy nucleus into approximately
equal parts  (which are nuclei of lighter elements)*
accompanied by  the release of a  relatively  large
amount of  energy.  Fission can occur spontaneously*
but usually is  caused by nuclear absorption of  gamma
rays* neutrons, or other particles.

The nuclei formed by the fission of heavy elements,
plus  the nuclides  formed by  the  fission fragments'
radioactive decay.

The series of  steps  involved in supplying fuel for
nuclear power  reactors.   It  includes mining, refining,
 the original fabrication of  fuel elements,  their use
 in a  reactor,  chemical processing to  recover the
 fissionable material remaining in the  spent fuel,
 re-enrichment  of the fuel material,  and refabrication
 into  new fuel elements.

 High-energy,  short-wavelength electromagnetic radia-
 tion.  Gamma radiation frequently accompanies alpha
 and beta emissions and always accompanies fission.
 Gamma rays are very penetrating, and are best stopped
 by dense materials.
                                    B-3

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general
environment:
 geometric mean:
 geometric standard
 deviation:



 geosphere:


 GW:

 heavy metal:
high-level radio-
active waste:
high-temperature
gas-cooled reactor:
The total terrestrial* atmospheric, and aquatic
environments outside sites within which any activity,
operation, or process associated with the management
and storage of spent nuclear fuel* high-level* or
transuranlc radioactive wastes is conducted.

The mean of a set of numbers* calculated by taking
the Nth root of the product of N numbers or by find-
ing the arithmetic mean of the logarithms of the
individual numbers.
The standard deviation of a set of numbers obtained
when calculating the arithmetic mean of the logarithm
of the individual numbers.  Also see geometric mean.

The solid portion of the earth, synonomous with the
lithosphere.

Gigawatts - one billion (1C9) watts.

All uranium* plutonium, or thorium placed into a
nuclear reactor.
Waste whose radioactivity is predominantly character-
ized by high-energy radiation; consists of the by-
products of nuclear reactors and wastes generated by
spent fuel processing operations of the nuclear fuel
cycle.  These are highly radioactive materials result-
ing from the reprocessing of spent nuclear fuel*
including liquid waste produced directly in reprocess-
ing and any solid material derived from such liquid
waste.
Nuclear reactor using uranium and thorium as a fuel
whose core is designed for high fuel utilization
efficiency.  The heat removal system is based upon
helium as a coolant.
ionizing radiation: Any electromagnetic or particulate radiation capable
                    of producing ions, directly or indirectly* in its
                    passage through matter.

Irregularly In-
herited disorders:  Genetic conditions with complex causes*
                    constitutional and degenerative diseases, etc.
                                   B-4

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isotope:
kg:
light-water
reactor (LWR):
linear energy
transfer (LET):
lognormal distri-
bution:
m9:
management and
storage:
member of the
public:
metric ton  (t)


mR/h:

nanocurie:
One of two or more atoms with the same atomic number
(the same chemical element) but with different atomic
weights.  Isotopes usually have very nearly the same
chemical properties, but some have somewhat different
physical properties.

Kilogram - the SI unit of mass, approximately equal
to 2.2 pounds.
A nuclear reactor whose heat removal system is based
on the use of ordinary water as the moderator and
reactor coolant.
The rate at which charged particles transfer their
energy to the atoms in a medium; expressed as energy
lost per distance traveled in the medium.
A distribution of the frequency of a value plotted on
a linear scale versus the value plotted on a logarith-
mic scale, which results in a bell-shaped curve.

Cubic meter - the SI unit of volume, approximately
equal to 35.3 cubic feet.
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  disposal of
these wastes.
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.

The  SI unit of weight equal to 1000 kilograms or 2205
pounds.

See  Roentgen.

See  curie.
neutron activation: The process of making a material radioactive by
                    bombardment with neutrons.
                                   B-5

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 neutron capture:
 neutron:
 noble gas:
 nonstochastic
 effect:
 rad  (radiation
 adsorbed  dose):
 radioactive decay:



 radioactivity:




 radionuelide:

 RBE:
 The process in which an atomic nucleus  absorbs or
 captures  a neutron.   The probability that a  given
 material  will capture neutrons is  dependent  on the
 energy of the neutrons and  on the  nature of  the
 material.

 An uncharged elementary particle with a mass slightly
 greater than that  of a proton, and found in  the nu-
 cleus  of  every atom  heavier than hydrogen.   Neutrons
 sustain the fission  chain reaction in a nuclear
 reactor.

 Any of a  group of  rare gases that  include helium,
 neon,  argon,  krypton,  xenon, and sometimes radon and
 exhibit great stability and extremely low reaction
 rates.
 Those health  effects  that increase in severity with
 increasing dose and usually have a threshold.
A measure of the energy imparted to matter by
radiation; defined as  100 ergs per gram.

A process whereby the nucleus of an atom emits excess
energy.  The emission of this energy is referred to
as radioactivity.

The property of certain nuclides of spontaneously
emitting particles or gamma radiation or of emitting
X-radiatlon following orbital electron capture or of
undergoing spontaneous fission.

A radioactive nuclide.

The ratio of the dose  (rad) of low-LET radiation to
the does of high-LET radiation producing the same
endpoint.  It is a measure of the effectiveness of
high-LET compared to low-LET radiation in causing the
same specific endpoint.
rem (roentgen
equivalent man):
remotely-handled
TRU waste:
                    Millirad  (mrad) - one thousandth of a rad.
A measure of equivalence for the relative biological
effect of radiations of different types and energies
on nan.


Those types of TRU wastes that must be handled by
robotics.
                                   B-6

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risk projection:     Absolute - risk projection based on the assumption
                    that there is some absolute number of  deaths  in a
                    population exposed at  a given age per  unit  of dose.

                    Relative - risk projection based on the assumption
                    that the annual rate of radiation-induced excess
                    cancer deaths is proportional to the ambient  rate of
                    occurrence of fatal cancer.

roentgen:           R is the symbol for roentgen, a unit of measurement
                    of X-radiation, equivalent to an absorbed dose in
                    tissue of approximately 0.9 rad.

                    Milliroentgen (mR/h) - one-thousandth  of a  roentgen.

spent nuclear fuel: Any nuclear fuel removed from a nuclear reactor after
                    it has been irradiated and whose constituent  elements
                    have not been separated by reprocessing.
standards:
stochastic effect:
storage:
target:
target theory
(Hit theory):
teratogenesis:
transuranic waste:
The "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.

Those health effects for which the probability of
occurrence is a function of the dose received.

Placement of radioactive wastes with planned capa-
bility to readily retrieve such materials.

Material subjected to particle bombardment or irradi-
ation in order to induce a nuclear reaction.
A theory explaining some biological effects of
radiation on the basis that ionization occurring in a
discreet volume (the target) within the cell, directly
causes a lesion which subsequently results in a
physiological response to the damage at that loca-
tion.  One, two, or more "hits" (ionizing events
within the target) may be necessary to elicit the
response.

Congenital abnormalities or defects.

Waste containing more than 100 nanocuries of alpha-
emitting transuranic isotopes, with half-lives greater
than 20 years, per gram of waste.
                                   B-7

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X-ray:              Penetrating electromagnetic radiation whose wave
                    lengths are shorter than those of visible light.
                    They are usually produced by bombarding a metallic
                    target with fast electrons in a high vacuum.  In
                    nuclear reactions, it is customary to refer to pho-
                    tons originating in the nucleus as gamma rays* and
                    those originating in the extranuclear part of the
                    atom as X-rays.  These rays are sometimes called
                    roentgen rays.

Zircaloy:           A zirconium alloy used as fuel cladding in some types
                    of nuclear reactors.
                                  B-8

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                                ACRONYMS





AEC       U.S. Atomic Energy Commission



ALAP      As low as practicable



ALARA     As low as reasonably achievable



AMAD      Activity median aerodynamic diameter



ANL       Argonne National laboratory



BEAR      Biological Effects of Atonic Radiation



BEIR      Biological Effects of Ionizing Radiation



BID       Background Information Document



CFR       Code of Federal Regulations



CH        Contact-handled



CRUM      Civilian Radioactive Waste Management



DEIS      Draft Environmental Impact Statement



DOD       U.S. Department of Defense



DOE       U.S. Department of Energy



DOT       U.S. Department of Transportation



DREF      Dose rate  effectiveness  factor



DWPF      Defense Waste  Processing Facility



ERC       President's  Federal  Energy Resources  Council



ERDA      Energy Research and  Development Administration



EPA      U.S. Environmental Protection Agency



FFTF      Fast Flux  Test Facility



FRC      Federal  Radiation Council
                                    B-9

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GI        Gastrointestinal



GW(e)     Glgawatts of electric power



HANF      Hanford, Washington



HEW      U.S. Department of Health, Education, and Welfare



HLW      High-level radioactive waste



HTGR      High-temperature gas-cooled reactor



ICFP      Idaho Chemical Processing Plant



ICRP      International Commission on Radiological Protection



ICRPTG    International Commission on Radiological Protection Task Group



INEL      Idaho National Engineering Laboratory



IRG      Interagency Review Group



LAKL      LOB Alamos National Laboratory



LET      Linear energy transfer



LLI      Lower large intestine



LMFBR     Liquid metal fast breeder reactor



LQ        Linear quadratic



LWR      Light-water reactor



MFRP      Midwest Fuel Recovery Plant



MIRD      Medical internal radiation dose



MRS       Monitored retrievable storage



MTHM      Metric tons of heavy metal



NCHS      National Center for Health Statistics



NCRP      National Council on Radiation Protection and Measurements



NFS       Nuclear Fuel Services



N-P       Nasophoryngial



NRC       U.S. Nuclear Regulatory Commission
                                   B-10

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NRFB      National Radiological Protection Board

NWPA      Nuclear Waste Policy Act of  1982

NWTS      National Waste Terminal Storage

OMB       Office of Management and Budget

ORNL      Oak Ridge National Laboratory

P         Pulmonary

R         Roentgen

RBE       Relative biological  effectiveness

RFP       Rocky Flats Plant

RH        Remote-handled

RIA       Regulatory  Impact Analysis

S         Stomach

SAB       Science Advisory Board

SI        Small intestine

SRP       Savannah  River Plant

T-B       Tracheobronchial

TRU       Transuranic

UL1       Upper  large intestine

UNSCEAR  United  Nations  Scientific Committee on the Effects  of  Atomic
          Radiation

WIPP      Waste  Isolation Pilot Plant


&U.3. GOVERNMENT PRINTING OFFICE 1995  529 866 31099
                                    B-ll

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