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
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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
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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
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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
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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
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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).
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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.
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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).
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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.
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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.
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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).
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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.
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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.
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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.
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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
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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.
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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
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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.
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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
2-12
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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.
2-13
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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).
2-15
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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.
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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.
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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.
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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
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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
-------
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.
-------
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)
-------
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.
-------
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
-------
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
-------
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
-------
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.
-------
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
-------
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
-------
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
-------
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
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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
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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
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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
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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.
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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
-------
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
-------
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
-------
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
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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
-------
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
-------
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
-------
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
-------
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
-------
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
-------
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,
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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
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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
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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.
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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.
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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
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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
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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
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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
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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.
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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.
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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).
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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
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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
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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
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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
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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).
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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).
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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
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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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Ne56 Neel J. V. and Schull W. J., The Effect of Exposure to the
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Ne70 Nelson K. B. and Deutschberger J., Head Size at One Year as a
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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
-------
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
-------
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
-------
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)
-------
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
-------
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
-------
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
-------
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
-------
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
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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.
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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.
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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.
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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.
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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.
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APPENDIX B
GLOSSARY AMD ACRONYMS
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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.
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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.
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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.
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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.
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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.
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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.
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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.
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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
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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
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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
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