United States
Environmental Protection
Agency
Office of
Radiation Programs
Washington DC 20460
ORP/CSD79-1
Radiation
&EPA
ALTERNATIVE
DISPOSAL CONCEPTS
FOR HIGH-LEVEL
AND TRANSURANIC
RADIOACTIVE
WASTE DISPOSAL
-------
This report was prepared as an account of work sponsored by the
Environmental Protection Agency of the United States government under
contract No. 68-01-3997. Neither the United States nor the United
States Environmental Protection Agency makes any warranty, express or
implied, or assumes any legal liability or responsibility for the
accuracy,, completeness or usefulness of any information, apparatus,
product or process disclosed, or represents that its use would not
infringe privately owned rights.
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Alternative Disposal Concepts for
High-Level and Transuranic Radioactive
Waste Disposal
Philip Altomare
Robert Bernard!
David Gabriel
Daniel Nainan
William Parker
Richard Pfundstein
May 1979
Contract Sponsor: EPA The MITRE Corporation
Metrek Division
Contract No.: 68-01-3997 1820 Dolley Madison Boulevard
Project No.: 15730 McLean. Virginia 22102
Oept.: W-53
MITRE Technical fleoort
MTR-7718
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FOREWORD
The Office of Radiation Programs carries out a national program
designed to evaluate the exposure of man to ionizing apd nonionizing
radiation, and to promote the development of controls necessary to
protect the public health and safety, and to assure environmental quality.
As part of this program, the office is developing standards for the
management and disposal of high-level radioactive wastes. A knowledge of
available technologies and their capabilities is necessary for the
development. This contract report examines a number of technologies
which have been proposed as alternatives to disposal of high-level wastes
in mined geological repositories.
Comments on this examination are welcomed; they may be sent to
the Director, Criteria and Standards Division (ANR-460),-Office of
Radiation Programs, U.S. Environmental Protection Agency, Washington,
D.C., 20460.
William A. Mills, Ph.D.
Director
Criteria & Standards Division
Office of Radiation Programs (ANR-460)
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ABSTRACT
Various alternatives have been proposed for the disposal of high-
level and transuranlc radioactive waste generated from the nuclear
electric power industry and the U.S. Defense program. The most
advanced disposal option, and the one under active development, is
the U.S. owned and operated deep-mined geologic repository. This
report reviews the primary alternative concepts to the geologic
repository, their present state-of-development and, to the extent
possible, their environmental Implications. The concepts included
are: transmutation, extraterrestrial disposal, seabed disposal,
ice sheet disposal, and other continental geologic disposal (matrix
of drilled holes, etc.). Projections of radioactive waste quantities
and the technologies for partitioning and fractionation of the waste
are also discussed.
This study reviewed information which was available through
approximately January of 1978.
iii
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TABLE OF CONTENTS
LIST OF ILLUSTRATIONS
LIST OF TABLES
1.0 INTRODUCTION 1-1
References 1-7
2.0 SUMMARY AND DISCUSSION 2-1
2.1 Disposal Options 2-2
2.1.1 Transmutation 2-2
2.1.1.1 Particle Accelerators 2-2
2.1.1.2 Nuclear Explosives 2-3
2.1.1.3 Fusion Reactors 2-3
2.1.1.4 Fission Reactors 2-3
2.1.2 Extraterrestrial Disposal 2-4
2.1.3 Seabed Disposal 2-7
2.1.4 Ice Sheet Disposal 2-9
2.1.5 Continental Geologic Disposal 2-11
2.2 Comparison of Disposal Concepts 2-13
2.3 Conclusions 2-19
3.0 QUANTITIES AND FORM OF HIGH-LEVEL AND TRANSURANIC 3-1
WASTE
3.1 Present and Projected Quantities of Waste 3-1
3.1.1 Existing Waste 3-1
3.1.2 Projected Quantities of Waste 3-7
3.2 Form of the Waste for Disposal 3-15
3.2.1 Spent Fuel 3-15
3.2.2 Reprocessed Waste 3-16
3.2.3 Partitioned and Fractionated Waste 3-19
References 3-21
4.0 PARTITIONING AND FRACTIONATION 4-1
4.1 Chemical Processes 4-2
4.1.1 Spent Fuel Reprocessing 4-2
4.1.2 Solvent Extraction 4-5
4.1.2.1 Actinides 4-5
4.1.2.2 Fission Products 4-9
4.1.3 Ion Exchange 4-10
4.1.3.1 Actinides 4-10
4.1.3.2 Fission Products 4-10
4.1.4 Precipitation Methods 4-11
4.1.5 Individual Nuclides 4-13
4.1.6 Other Methods of Partitioning 4-15
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TABLE OF CONTENTS (Continued)
4.2 Environmental and Health Considerations
4.3 Economic Impact
References
5.0 TRANSMUTATION
5.1 Transmutation Concepts
5.1.1 Particle Accelerators
5.1.1.1 Direct Bombardment by Charged
Particles
5.1.1.2 Coulomb Excitation
5.1.1.3 Photon Transmutation
5.1.1.4 Spallation Neutrons
5.1.2 Nuclear Explosives
5.1.3 Fusion Reactors
5.1.4 Fission Reactors
5.1.4.1 Lightwater Reactors
5.1.4.2 Fast Neutron Reactors
5.1.4.3 Thorium-Uranium Reactors
5.1.4.4 Actinide Cross-Sections
5.1.4.5 Fission Product Transmutation
5.2 Environmental and Health Considerations
5.3 Economic Impact
References
6.0 EXTRATERRESTRIAL DISPOSAL
6.1 Basis of Reference Studies
6.2 Space Disposal Concept
6.2.1 Waste Capsule and Reentry Shield
6.2.2 Launch Operations
6.2.3 Technical Feasibility
6.3 Environmental and Health Considerations
6.3.1 Normal Operations
6.3.1.1 Partitioning and Encapsulation
6.3.1.2 Terrestrial Transportation
6.3.1.3 Space Transportation
6.3.2 Abnormal Events
6.3.2.1 Launch Vehicle Accidents
6.3.2.2 Radioactive Waste Releases
6.3.3 Recovery and Contingency Planning
6.3.4 Shuttle, Waste Capsule Integration
6.3.5 Radiological Considerations
6.4 Economic Impacts
6.4.1 Partitioning
6.4.2 Encapsulation
vi
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TABLE OF CONTENTS (Continued)
6.4.3 Space Launch Costs 6-44
References 6-47
7.0 SEABED DISPOSAL 7-1
7.1 Ocean Characteristics 7-5
7.1.1 Continental Margin 7-7
7.1.2 Mid-Oceanic Ridge (MOR) 7-9
7.1.3 Ocean Basin Floor : 7-10
7.1.4 Criteria for Site Selection of Oce'anic 7-11
Provinces
7.2 Emplacement Techniques 7-14
7.2.1 Free Fall Penetration 7-14
7.2.2 Winch-controlled Emplacement 7-16
7.2.3 Drilled Holes 7-16
7.3 Environmental and Health Considerations 7-17
7.3.1 Engineering and Environmental Barriers 7'^
Against Waste Intrusion into the Biosphere
7.3.1.1 Waste Form 7-18
7.3.1.2 Canister 7-21
7.3.1.3 Sediment 7-25
7.3.1.4 Ocean 7-35
7.3.1.5 Summary - Barrier Effectiveness for 7-36
Waste Isolation
7.3.2 Research Needs 7-37
7.3.2.1 Ecological Implications of Thermal 7-41
Waste Heat
7.3.2.2 Hole Closure 7-41
7.3.2.3 Summary of Other Data Requirements 7-42
7.3.3 Radiological Impact Assessment 7-42
7.3.3.1 Source Term 7-44
7.3.3.2 Environmental Pathways to Man 7-47
7.3.3.3 Nuclides of Importance if Barriers 7-49
Maintain Expected Integrity
7.3.3.4 Dose Assessment 7-56
7.3.3.5 Operational and Transportation Risks 7-61
7.4 Economics 7-63
7.4.1 Cost Estimates 7-66
References 7-69
8.0 ICE SHEET DISPOSAL 8-1
8.1 Descriptions of Ice Sheet Disposal Concepts 8-1
8.1.1 Meltdown or Free Flow Concept 8-3
8.1.2 Anchored Emplacement Concept 3-5
8.1.3 Surface Storage Facility Concept 8-6
8.2 Status of Ice Sheet Technology Development 8-7
vii
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TABLE OF CONTENTS (Continued)
8.2.1 Emplacement
8.2.2 Transportation
8.3 Environmental Considerations
8.3.1 Availability of Ice Sheet Data and
Uncertainties
8.3.2 Long-Term Containment
8.3.2.1 Motions of Ice Sheets
8.3.2.2 Physical State and Rates of Ice Flow
8.3.2.3 Meltwater at Base of Ice Sheet
8.3.2.4 Long-Term Stability
8.3.3 Characteristics of Waste Forms
8.3.4 Site Requirements
8.3.5 Radiological Risks
8.3.6 Accidental Risks and Consequences
8.3.7 Additional Data Requirements
o • j • o s^irmnfi i*y
8.4 Capital and Operating Costs
8.5 Policy and Treaty Agreements
References
9.0 CONTINENTAL GEOLOGIC WASTE DISPOSAL
9.1 Concept Description
9.1.1 Solution-Mined Cavities
9.1.2 Waste Disposal in a Matrix of Drilled Holes
9.1.3 Waste Disposal in Superdeep Holes
9.1.4 Deep Well Injection
9.1.5 Hydrofracture
9.1.6 Rock Melting Concepts
9.1.6.1 Mined Cavity/Liquid Waste/Interim
Cooling
9.1.6.2 Mined Cavity/Solid Waste/Interim
Cooling
9.1.6.3 Deep Drilled Hole/Solid Waste/No
Interim Cooling
9.1.6.4 Solid Waste/Capsule/Deep Descent
9.2 Siting (Environmental) Considerations
9.2.1 Geologic, Hydrologic, Climatic and Other
Criteria Which May •Affect Long Term Confinement
9.2.1.1 Thermal Properties of the Host Rock
9.2.1.2 Engineering Properties of the Host
Rock '
9.2.1.3 Water Content of the Host Rock
9.2.1.4 Mineral Resource Potential
9.2.1.5 Geothermal Resource Potential
viii
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TABLE OF CONTENTS (Concluded)
9.2.1.6 Seismicity and Faulting
9.2.1.7 Depth of Disposal
9.2.1.8 Dimensions of Host Rock
9.2.1.9 Climate and Possible Change in
Climate
9.2.2 Pathways and Barriers of Migration of 9-23
Nuclides
9.2.2.1 The Waste Form 9-24
9.2.2.2 The Canister 9-26
9.2.2.3 Geologic System of the Host Rock 9-26
9.3 Technical Feasibility of Alternative Geological 9-28
Disposal Concepts
References 9-33
ix
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LIST OF ILLUSTRATIONS
Figure Number Page
2-1 .Alternative Waste Disposal Pathways 2-15
3-1 US Nuclear Power Growth Projection 3-8
3-2 Perspective on the Buildup of Spent 3-9
Fuel and Associated High Level Wastes
vs. Time (Nominal Growth Case,
Throwaway Cycle)
4-1 Conceptual Processing Sequence for 4-6
Actinide Partitioning
5-1 Enrichment Requirements for Actinides 5-9
Recycle
6-1 Extraterrestrial Disposal Process Steps 6-6
6-2 Transuranic Waste Capsule for Space 6-8
Disposal
6-3 Reentry Shield and Transuranic Disposal 6-8
Package for Solar Escape Destination
6-4 MHW Heat Source 6-14
6-5 Pad Configuration 6-16
6-6 Space Transportation Systems 6-17
6-7 Space Shuttle Launch-To-Landing Sequence 6-20
6-8 Number of Space Shuttle Launches Required 6-21
Per Year for Disposal of Only Actinides
Into High Earth Orbit or by Solar System
Escape. Prior 10-year Earth Storage
6-9 Radiological Recovery Sequence 6-35
6-10 Generalized Flow Diagram for Risk 6-39
Analyses
7-1 Engineering Concepts for Emplacement of 7-15
Radioactive Waste Canisters in the Seabed
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LIST OF ILLUSTRATIONS (Concluded)
Figure Number Page
7-2 Transport Processes of Radionuclides 7-19
from Seabed Disposal
7-3 The Proposed Standard Canister 7-23
8-1 Schematic of Operations in Ice Sheet 8-2
Disposal Systems for High-Level
Radioactive Wastes
8-2 Ice Sheet Disposal Concepts 8-4
8-3 Potential Cask-Canister Recovery 8-18
8-4 Overall Research and Development 8-20
Schedule Waste Disposal in Ice Sheet
9-1 Flow Diagram for Emplacement of 9-3
Solidified Waste
9-2 Flow Diagram for Emplacement of 9-4
Liquid Waste
9-3 Generalized Concept Solution Mining 9-7
Final Storage Facility
9-4 Solid Waste Emplacement in a Matrix 9-9
of Drilled Holes
xi
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LIST OF TABLES
Table Number Page
2-1 Summary of Disposal Concepts 2-16
3-1 Summary of Defense Waste Quantities 3-3
3-II Inventory of Major Fission Products 3-4
and Actinides in Hanford High-Level
Wastes Decayed to 1990
3-III Radionuclide Content - Savannah 3-5
River fligh-Level Wastes 1985
3-IV Typical Composition of Calcined Solids 3-6
Idaho Chemical Processing Plant
3-V Principal Radionuclides in Waste— 3-10
Throwaway Cycle
3-VI Estimated Range of Total Domestic High- 3-14
Level Waste Burden - (Circa 2010)
5-1 Comparison of Actinide Inventories For 5-11
Two Recycle Strategies
5-II Actinide Reaction Rates ,in Fast and 5-12
Thermal Reactors (Reactions/sec/Atom)
5-III Actinide Recycle From One 1200 MWe 5-13
LMFBR and Three 1200 MWe LWR's
5-IV Actinide Recycle Schemes 5-17
5-V Incremental Cost for Transmutation 5-25
of Actinides
6-1 Characteristics of Waste for Final Disposal 6-11
6-II Thermal Power and Radioactivity of Trans- 6-12
uranics in 10-Year-Old Waste
6-III Typical Launch Accidents 6-27
6-IV Mission Potential Fuel Release Events 6-32
6-V Mission Prompt Source Term Summary 6-33
xii
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LIST OF TABLES (Concluded)
Table Number Page
7-1 Characteristics of the Ocean Provinces 7-8
7-II Estimated Distribution Coefficients (Kj 7-30
and Retardation Factors (R.) In A
Typical Desert Soil
7-III Estimated Distribution Coefficients (Kd> 7-31
In Deep-Sea Sediments
7-IV Potential Barrier Effectiveness for Waste 7-38
Isolation
7-V Radionuclide Amounts in Initial Seabed 7-45
Repository
7-VI Concentration Factors 7-50
7-VII Pathways to Man and Modes of Exposures. 7-51
7-VIII Radionuclide Amounts After 106 Years 7-53'
of Decay
7-IX Levels of Natural and Fallout Radionuclides 7-60
in Sea-Water
7-X Estimated Dose and Dose Commitment From 7-64
Marine Food Chain For Loss of Plutonium
Package At Sea
7-XI Estimated Dose Commitment From Marine Food 7-65
Chain for Loss of A Spent Fuel Shipping
Cask Containing 3.1 MT of Uranium
7-XII Summary of Cost Data for Seabed Disposal 7-67
8-1 Capital and Operating Cost Items for Ice 8-24
Sheet Disposal
xiii
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1.0 INTRODUCTION
One of the major environmental, health, and safety concerns
related to nuclear power is the permanent disposal of radioactive
wastes. In particular, spent reactor fuel or reprocessed fuel waste
is characterized by high levels of radioactivity, with some fission
products and transuranlc radionuclei remaining as hazardous sub-
stances for more than a million years. Because of the hazard to
human health from radioactive wastes, these wastes must be placed in
disposal sites capable of containment for periods approaching geolo-
gic time scales.
The Office of Radiation Programs of the U.S. Environmental
Protection Agency (EPA) has a primary responsibility to establish
radiation protection standards. In carrying out this responsibility,
the EPA must assess the public health and the environmental impact
of radiation from all sources in the United States.
This study supports EPA's assessment of radioactive waste for
purposes of establishing environmental protection standards. It is
one of several concurrent studies sponsored by EPA in the evaluation
of high-level and transuranlc waste. These companion studies include
a MITRE study, Assessment of Waste Management of the Volatile Radlo-
nuclldes1 and the Arthur D. Little Inc. study, Technical Support
for Radiation Standards for High-Level Radioactive Waste Manage- •
meat.2
The Arthur D. Little (ADL) study provides a technical assessment
of the proposed U.S. disposal approach of placing high-level and
1-1
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transuranic radioactive waste in stable deep geologic formations.
For purposes of this report, this concept is referred to as deep-
mined geological repositories. The intent of the report is to exam-
ine alternative methods proposed for the disposal of high-level and
transuranic radioactive waste. These alternative concepts include
the following:
• Transmutation (Section 5) - nuclear conversion of radioiso-
topes to non-radioactive or short half-life isotopes
• Extraterrestrial Disposal (Section 6) - removal of waste from
the earth and disposal in space or on planetary bodies
• Seabed Disposal (Section 7) - placement of the waste in the
seabed thereby utilizing the ocean as an additional barrier
between the waste and man
• Other Continental Disposal - alternative methods for disposal
of waste on the earth land masses
For presentation purposes, the continental disposal is further
separated into an ice sheet disposal concept (Section 8) and conti-
nental geological disposal concepts (Section 9). Because several
disposal concepts require the separation of the waste into radionu-
clide groupings, Section 4 discusses the technology of partitioning
and fractionation of radioactive waste. Section 3 provides back-
ground on the quantities and forms of radioactive waste for disposal.
Section 2 provides a summary of the disposal concepts and compares
the merits of the alternative approaches.
At present there is no accepted method for the final disposal of
high-level or transuranic radioactive waste. The placement of these
1-2
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wastes in deep mines in geologically stable formations is the most
technically developed concept and therefore the one which offers the
most promise for early application. In this concept, a stable dry
geological formation is to be selected and the radioactive waste em-
placed in a mined area. These deep mined repositories can serve as
interim storage areas until the long term isolation capability of the
facility is confirmed or separate rooms may be backfilled as the
waste is emplaced with eventual sealing of all openings for final
disposal. The ADL study examines this concept in detail. The reader
is referred to reference 2 for a discussion of the geological reposi-
tory disposal concept.. A health risk assessment for geological
respository disposal is presently being prepared by EPA.
The difficulties in designating a final disposal method arise
primarily from the need to assure that these highly radibtoxic wastes
will be isolated from the biosphere for many thousands of years.
Predictions of geological and hydrological behavior over such time
periods are at best difficult and involve a large degree of uncer-
tainty. Research and development must, however, provide reasonable
assurance that the risk to present and future generations will be
acceptable. The definition "acceptable" is an issue unto itself
which has been and is being addressed by EPA.3>^>5
The form of the waste is significant in determining the method
of final disposal. As originally conceived, the spent fuel elements
are chemically processed to recover the usable uranium and plutonium
1-3
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as a part of the nuclear fuel cycle. A high-level radioactive resi-
due results from this reprocessing operation. , The residue, an aque-
ous raffinate, could be treated in several ways to form solids of
different degrees of leach resistance and further packaged for final
disposal. However, increasing concerns as to the potential for
diversion of nuclear materials to weapons production has resulted in
a moratorium on the reprocessing of commercial fuel in the U.S.
This decision produces substantial uncertainty as to the direction of
nuclear waste management programs. Spent fuel elements may or may
not be disposed of directly and reprocessed waste may or may not be
available for further treatment to meet the requirements of various
disposal options. Thus, the disposal of intact spent fuel must be
considered as a possible requirement.
The proliferation issue and potential for diversion of nuclear
*
materials are affecting the development and implementation of the
U.S. nuclear program. Different fuel cycles and reactor types are
presently under consideration. Development of the uranium-plutonium
fuel cycle may not occur in favor of the establishment of a throwaway
"cycle" (direct disposal of spent fuel elements), or the uranium-
thorium cycle. These latter fuel cycles have advantages in limiting
the accessibility to weapons grade material. Nuclear reactor types
may continue with the Light Water Reactors (LWR) to simply utilize
the uranium-235 resources, may shift to Heavy Water (D20) Reactors
to obtain greater utilization or burnup of the fissile U-235, or may
1-4
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shift to the Gas Cooled Reactor (GCR) or Light Water Breeder Reactor
(LWBR) but in centralized energy generating and fuel processing parks
where safeguards are more easily implemented. For the different
reactor types, different fuel handling and processing'facilities will
be required.
All of the above and more will affect the quantities, type, and
form of the nuclear waste. In proceeding with a discussion of alter-
native nuclear waste management concepts, it must be borne in mind
that there are many steps of research development and design. Final
selection, evaluation, and implementation is a complex process and is
influenced by economic, political, and technical factors.
Obviously, problems remain of both a technical and political na-
ture that must be resolved to determine the most appropriate disposal
method. For the present, therefore, it is prudent to continue to
consider each of the possible radioactive waste disposal methods and
to assume that processing of the spent fuel may be implemented either
for purposes of fuel recycle or for preparation of the wastes for
disposal.
Finally, although the studies presented herein are directed to-
ward radioactive waste, the treatment and disposal of other toxic
waste produced by man's activities are no less a concern. The con-
cepts, the problems encountered, and solutions derived for radioac-
tive waste will probably find application to treatment of waste from
other sources.
1-5
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REFERENCES
1. "Assessment of Waste Management of the Volatile Radionuclides,"
Draft, MTR-7719, MITRE, February 1978.
2. "Technical Support For the Radiation Standards For High-Level
Radioactive Waste Management," Tasks A to D, Draft, Arthur D.
Little Inc.
3. "U.S. EPA, Proceedings: A Workshop on Policy and Technical
Issues Pertinent to the Development of Environmental Protection
Criteria For Radioactive Wastes," Report: ORP/CSD-77-1, Reston,
Va. (1977).
4. "U.S. EPA, Proceedings: A Workshop on Policy and Technical
Issues Pertinent to the Development of Environmental Protection
Criteria for Radioactive Waste," Report: ORP/CSD-77-2,
Albuquerque, NM (1977).
5. U.S. EPA, Background Report: "Considerations of Environmental
Protection Criteria for Radioactive Waste," February 1978.
6. Statement by President Carter on Nuclear Power Policy, The White
House, April 7, 1977.
1-6
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2.0 SUMMARY AND DISCUSSION
High-level and transuranic radioactive waste is created in the
commercial nuclear industry and the U.S. defense programs. High-
level waste in the context of this report is the highly radioactive
liquid, containing fission products and actinides, which is the
residue from the reprocessing to recover the uranium and plutonium
from the spent fuel. High-level waste may also refer to the unre-
processed spent fuel elements in the throwaway "cycle." The bulk of
this radioactive waste by the year 2000 will be from spent fuel dis-
charged from nuclear electric generating plants. This waste is char-
acterized by high specific radioactivity and is of particular concern
to human health and the ecosystem since some fission products and
produced radioactive actinide isotopes remain hazardous for hundreds
to millions of years.
The most developed concept so far for the disposal of high-level
and transuranic radioactive waste is the deep-mind geologic reposi-
tory. This method has reached the facility design and site selection
stage. Extensive technology research and development have been
undertaken and studies of geologic formations have been and are being
conducted to ensure the long-term isolation of waste from the bio-
sphere. Many alternative approaches to the final disposal of high-
level and transuranic radioactive waste have been proposed. While
none of these alternatives is as advanced as the deep-mined geologic
repository, they may supplement or replace this method at some future
2-1
-------
time if proven technically and economically practical and environ-
mentally acceptable.
There are several alternative disposal concepts that have been
considered in this report:
• Transmutation
• Extraterrestrial Disposal
• Seabed Disposal
• Ice Sheet Disposal
• Alternate Geologic Disposal Concepts
2.1 Disposal Options
2.1.1 Transmutation
Transmutation is the conversion of a radionuclide of undesirable
characteristics (long life or high toxicity) to a different nuclear
species by nuclear processes. The transmuted nuclide would have more
favorable characteristics for disposal by forming a stable or short--
lived isotope. Transuranic elements could be converted to a fissile
isotope which could be fissioned or recycled.
Several methods are considered for transmutation of radionu-
clides:
• Particle accelerators
• Thermonuclear or fission explosives
• Fusion reactors
• Fission reactors
2-2
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2.1.1.1 Particle Accelerators. Transmutation by particle ac-
celerators, while feasible, has not been determined to be practical.
The associated problems are high energy usage which can exceed the
energy generated in producing the waste, expected high cost, and
radioactive contamination.
2.1.1.2 Nuclear Explosives. It has been estimated that eleven
one-hundred kiloton nuclear detonations per year would be required
for transmutations of the long-lived fission products from each 1000
MWe reactor. It is not considered likely that this method of waste
disposal would be considered acceptable.
2.1.1.3 Fusion Reactors. Fusion reactors potentially have very
high neutron flux levels (1015 to 1016 neutrons/cm2 -sec). The
high energy neutrons produced in fusion reactors can be used directly
to cause neutron induced reactions, or thermalized for capture in
fission processes. Fluxes of this order of magnitude raise the pos-
sibility of transmuting not only actinides but also fission product
nuclides such as Kr-85, Zr-93, Tc-99, and 1-129, which are not consi-
dered practical for transmutation in fission reactors.
A sustained fusion, or thermonuclear, reaction has not yet been
achieved. A major breakthrough is required before this technology
can be realized.
2.1.1.4 Fission Reactors. Transmutation in fission reactors
entails removal of the selected radionuclides from the spent fuel,
2-3
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fabrication into new fuel elements or separate elements, and irradia-
tion to achieve transmutation by neutron capture.
Studies performed to date indicate that the transmutation of
important actinides (Np, Am, Cm, Pu, Bk and Cf) is feasible. Re-
search, development, and design studies are required, however, to
implement this technology. In particular, reaction cross-section
need to be measured, reprocessing and partitioning techniques have to
be developed or perfected, and reactor designs must be developed and
tested.
Transmutation in fission reactors of important long-lived fis-
sion products does not appear practical in that substantial removal
of the radionuclides in reasonable time periods is not achievable.
For example, reduction of the Tc-99 by a factor of 1000 could require
165 years, and to 10 percent 55 years. Specially designed reactors
that obtain high thermal: neutron fluxes from fast neutron reactors
could conceivably reduce the irradiation time. The practicality of
reactors of this type has not been evaluated.
2.1.2 Extraterrestrial Disposal
The concept of extraterrestrial disposal consists of placing a
capsule containing the waste in space where further contact with
earth is essentially eliminated. The space shuttle is being consi-
dered as the launch vehicle for extraterrestrial disposal because of
its lower cost, and the added safety and reliability of a manned
spacecraft.
2-4
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Several disposal destinations have been studied:
• High earth orbits
• Solar orbits
• Solar system escape
• Solar impact
• Lunar impact or landing
• Planetary impacts
High earth orbits are unattractive since there is a possibility
that the earth will recapture the waste. Solar, orbits are possible
for waste disposal, however, a portion of the solar system could
become contaminated following failure of the waste capsule. Solar
impact is not practical with currently available space vehicles.
Planetary impacts are ruled out at the present time by international
agreements. Solar system escape would provide complete disposal of
the waste from the earth and solar system.
Lunar landings offer some advantage in that the waste could be
stored with minimal risk and later recovered or launched to other
space destinations. International agreements would be required for
lunar disposal.
Extraterrestrial disposal of all high-level and transuranium
waste to be generated in nuclear power reactors is not currently
feasible. This would require an excessive number of launch opera-
tions, and both the cost and environmental effects would preclude
such an operation. Environmental effects include noise and sonic
2-5
-------
booms, acidic rain, reduction in upper atmosphere ion concentrations
and the local community interactions. The environmental effects are
expected to be of minimal significance for the normal anticipated
operations of some 50 to 100 flights per year by the year 2000 and
the additional launches that might be conducted for selected waste
disposal.
Extraterrestrial disposal of the actinides and separated long-
lived fission products is considered feasible, but not necessarily
practical. Studies presently being conducted by NASA are investiga-
ting space disposal concepts for the fission products and transuranic
waste with uranium reclaimed. Depending on the composition of the
waste, 100 to 250 space shuttle launches per year might be required
by the year 2000. The number of launches is affected by the degree
of separation of the fission products from the actinides, the age,
and the method of encapsulation and radiation shielding of the waste.
The encapsulation and reentry shield must be designed for maxi-
mum containment of the waste even in the event of a catastrophic
launch vehicle explosion and fire or reentry of the waste capsule
into the earth's atmosphere. The additional weight required for this
protection reduces the payload per launch resulting in increased cost
and increased number of launches for extraterrestrial disposal.
The major disadvantages of extraterrestrial disposal are the
potential for accidents and the cost. The waste form, the encap-
sulation method, the launch system, and the mission profile for
2-6
-------
extraterrestrial disposal have not been sufficiently defined for an
analysis of accidents and their consequences. Preliminary worst case
analyses do, however, indicate that they are potentially serious.
Accident risk can be substantially reduced by system design and
limitation on the quantities or types of waste for space disposal.
Whether space disposal of waste with the necessary safeguards is
economical, or whether the risk of accidents is acceptable, will
require extensive study. The risk of accidents with the potential
for releasing radioactive materials directly into the environmental
must be carefully evaluated.
2.1.3 Seabed Disposal
Seabed disposal is the controlled emplacement of radioactive
waste in deep sea sediments or rock formations under the ocean. The
ocean floors are divided into three principal physiographic provin-
ces: Continental Margin, Midoceanic Ridge, and Ocean Basin Floors.
Some of these areas may contain possible locations for controlled
emplacement of high-level radioactive waste. Potential sites for
high-level waste disposal will be selected on the basis of high geo-
logic stability; predictability for geologic time periods; limited
resource potential; biological nonproductivity and sediment charac-
teristics which are effective as barriers to radionuclide migration.
Based on sediment data from numerous drilling experiments, seismic
profiles, and bottom sediment photographs, the ocean areas in the
middle of the tectonic plates and the middle of the ocean gyres
2-7
-------
(mid-plate/mid-gyre) exhibit characteristics which are particular!;
attractive as seabed disposal sites. Sediment sampling experiments
are currently underway at two designated sites located in the middle
of the North Pacific mid-gyre region to establish the suitability of
these areas for high-level waste repositories.
An exact procedure for emplacement of radioactive waste canis-
ters will not be chosen until seabed disposal has been determined to
be feasible. However, several techniques are possible: free fall
penetration in sediments, winch controlled emplacement in clay sedi-
ment, and drilled holes into underlying rock formations. Free fall
penetration requires that the clay sediment have plastic properties
which will collapse to fill the resultant cavity in reasonable time.
Laboratory studies indicate that closure of the emplacement cavity
would occur. In winch controlled emplacement, laboratory studies
indicate that there may be some cavity closure problems and that a
sealant may be required. Deep sea drilling from a surface ship has
been demonstrated by several marine research centers. This emplace-
ment technique has the advantage that many canisters could be placed
in a single bedded area at depths of 100 to 500 meters. A hole seal-
ant would be required. To date, drilling techniques using sealants
for seabed disposal have not been demonstrated.
Much of the information needed to adequately assess the overall
feasibility of seabed disposal is not available. There are several
2-8
-------
areas which require further information, particularly the ability of
seabed disposal to act as a barrier to radionuclide migration:
• information on the characteristics of ocean provinces to
determine and establish their overall suitability as poten-
tial seabed disposal sites
• technological capabilities including transportation, ship-
ment, and placement of wastes
• leach rates for all radionuclides in proposed waste forms
V physical properties of deep sea sediments
• sorption and distribution coefficients of deep sea sediments
• retardation factors of sediments
• effects of thermal gradients on sediments (heat transfer pro-
perties)
• dynamic response of sediment to canister emplacement
• transport processes of radionuclides in deep sea'sediments
including structural and chemical properties and 'driving
forces
• transport processes in the water column, including diffusion
currents, advection, biological (feed web), and thermal
plume
Because of the uncertainties associated with seabed disposal, it
is not presently possible to conclude that this concept represents a
practical long-term solution to the waste disposal problem.
2.1.4 Ice Sheet Disposal
Disposal of high-level and transuranic waste in the Antarctic
and Greenland ice sheets has been proposed. The favorable features
of ice sheet disposal are geographic isolation, relative isolation of
2-9
-------
the waste from inhabited areas in the event of waste leakage, low
temperatures, and rapid heat dissipation.
Several waste emplacement concepts have been considered. A
meltdown or free flow concept emplacement would be accomplished by
predrilling-a shallow hole and allowing the thermal heat of the waste
canister to melt or free flow to the ice sheet basal. An anchored
emplacement concept would provide for 200- to 500-meter-long cables
anchored at the surface to hold the waste canister in place. A sur-
face storage facility has also been considered. The surface storage
facility would be mounted on jack-up pilings or piers resting on load
bearing plates. Cooling of the canisters in a surface facility would
be by natural air flow. Both the anchored emplacement and surface
storage would provide for retrievability.
At the present time, there is insufficient knowledge of the phy-
sics and history of ice sheets. International groups of glaciolo-
gists concluded that ice sheets could not seriously be considered for
radioactive waste disposal without further investigation in certain
areas of limited knowledge:
• the evolutionary processes in ice sheets
• the relationships of ice sheet behavior with climatic chan-
ges
• the nature of future climatic changes on the stability of ice
sheets
Ice sheets are not considered a feasible concept for the dispo-
sal of the long-lived radioactive waste at this time.
2-10
-------
2.1.5 Continental Geologic Disposal
The continental geologic disposal concept is to place radioac-
tive waste in stable geologic formations. The concept relies upon
• ^
the long-term stability and the nuclide retention capability of the
geology to isolate the waste for periods of millions of years. Since
/
water is a primary transport mechanism, the selected geologic forma-
tions must be essentially free of groundwater movement.
The deep-mined geologic repository is the most advanced concept
for the disposal of high-level and transuranic radioactive waste.
This concept has proceeded to the stage of facility design, and ef-
forts are underway to locate a politically and geologically accept-
able site. Deep salt deposits have received the most attention as a
suitable disposal media because of their demonstrated stability over
very long time periods, their homogeneity, and their capability of
plastic flow which would tend to seal cracks or fissures that may
develop from mining operations or as a result of temperature gradi-
ents. Crystalline rock formations such as granite or basalt, shales,
limestones, and certain clay beds are also being considered for dis-
posal sites.
The deep-mined geologic repository would consist of surface
facilities to receive and handle the waste, and mines 300 to 1500
meters deep in the selected rock formations'. Capability to repackage
the waste, if required, would be included at the surface facility.
2-11
-------
The waste would be emplaced in the floor of the mine shafts at spa-
cings limited by the heat production rate of the waste.
There are several alternative proposed geologic disposal con-
cepts :
• solution mined cavities in salt deposits
• matrix of drilled holes
• super-deep holes
• deep well injection
• hydro-fracture
• rock melting concepts
Solution-mined cavities, matrix of drilled holes, and super-deep
holes offer the possibility of deeper emplacement of waste than a
repository which is limited by mine opening constraints. The random
emplacement of waste packages in solution mining is such, however,
that only low-heat rate waste such as actinides can be considered.
The technology for super-deep boreholes has not been developed. A
matrix of drilled holes requires the development of a hole sealant
which will be an effective barrier to radionuclide transport.
Deep well injection and hydrofracture concepts may have applica-
tion to low-level and intermediate-level liquid waste, but the long-
term containment required for high level and transuranic waste has
yet to be proven.
In rock-melting concepts, the heat of the radioactive waste
melts the rock. The waste then descends to deeper depths or it mixes
2-12
-------
with the rock to form a waste-rock mix which eventually cools and
solidifies* The rock-melting concepts require research and develop-
ment to establish their practicality and to determine the physical
characteristics and behavior of rock-waste mixes.
The problem of assessing isolation capability for alternative
geologic disposal concepts is similar to those for the deep-mined
geologic repository. Studies underway to determine the environmental
acceptability of deep-mined geologic repository will therefore be of
interest for other geologic disposal concepts.
The major areas of uncertainty in the deep-mined geologic repo-
sitory are in the area of heat and mining effects on the host rock
formation and the assurance that the radionuclides will not escape
the repository as a result of natural events or accidental human in-
trusion over the long time periods required before they decay to
innocuous levels.
2.2 Comparison of Disposal Concepts
There are several concepts for the disposal of high-level and
transuranic radioactive waste which have the potential for eventual
implementation:
• geologic disposal (primarily the deep-mined repository)
• seabed disposal
• extraterrestrial disposal for certain separated waste
• fission transmutation for actinides
• fusion transmutation for actinides and long-lived fission
products
2-13
-------
SPENT FUEL-
Uranium,
Uranlun-plutonlua,.
Uranlum-Thorlua,
Recycle
i— »-No B
1
1
1
1
1
1
I— «.&ep
m«. ^
ua.
eprocesslng^
1
roc
Disposal Of CHCAT-ULAl
•Spent Fuel DICAToULAl
__ Storage of
* Spent Fuel '
No Partitioning Preparation
or Fractionation Disposal
Short Half-
Fission Pro
^ Fractionation
"^(Fission Products)
Long Half
* Fission Pro<
t Partitioning
fActlnldes)
1 GEOLOGIC
l_ EXTRATERRESTRIAL
Llf Preparation. 1 GEOLOGIC
Encapsulation ' EXTRATERRESTRIAL
-Life Preparation. i EXTRATERRESTRIAL
iucts "Fabrication"' "'"l— FUSION TRANSMUTATION
Preparation. I EXTRATERRESTRIAL
Fabrication . n«.Tnu i-i>»ieu»».«.u
Legend
-^-^— Pathway
^— — Questionable pathway
O Option point
Volatile
Radlonuclldes:
* C-l«. 1-129 '
Kr85. H-2
Half-Life
C-14, 1-129
.Short Half-Life
Kr-85. H-3
Preparation
Encapsulat ion
Preparation
Encapsulation
CEOLOCIC
SEABED
EXTRATERRESTRIAL
•
I FUSION TRANSMUTATION
CEOLOCIC
SEABED
ENGINEERED STORAGE
—— FUSION TRANSMUTATION
FIGURE 2-1
ALTERNATIVE WASTE DISPOSAL PATHWAYS
-------
The various disposal pathways for commercial high-level and trans-
uranic radioactive waste management are shown in Figure 2-1.
Present U.S. policy has deferred the reprocessing of commercial
spent fuels. The spent fuel must, therefore, either be disposed of
directly or committed to retrievable storage. Disposal options for
spent fuel would be limited to geologic or seabed disposal. However,
if the spent fuel is stored, it could eventually be returned for
reprocessing and the alternative disposal options as indicated in
Figure 2-1 would then be possible. The technical development and
environmental studies of the alternative concepts have not advanced
to a stage where quantitative comparisons can be made. In particu-
lar, the environmental, health, and safety aspects, as well as the
probability for accidental release and the consequences of such
releases, must be assessed for each step of the waste management pro-
cess before a meaningful comparison can be made. The relative merits
of the alternative disposal schemes are presented. No attempt is
made herein to rank the desirability of the disposal options nor
should any preference be implied.
The state of development, the major problems, and the advantages
of the alternative disposal options are listed in Table 2-1.
The deep-mined geologic repository is the most advanced disposal
concept and offers the earliest possibility for implementation. The
major problem facing the acceptance of the deep-mined geologic reposi-
tory is the reasonable demonstration that isolation can be maintained
2-15
-------
TABLE 2-1
SUMMARY OF DISPOSAL CONCEPTS
to
1
Disposal Concept
Deep Geologic Disposal
Alternate Geologic
Disposal
Seabed Disposal
Ex I: rater res trial
Disposal
Fission Reactor
Transmutation
!Fusion Reactor
Transmutation
State-of-Developmen
Advanced state of
development
Early stage of
development
Early stage of
development
Early stage of
development
Early development
for actinides,
questionable appli-
cation to fission
products
Dependent on fusion
reactor development
Major Problems
Advantages
Assurance of long term
isolation is required
Proof of isolation,
technical development
is needed
Data is required for
proof of concept and
long term isolation
High potential of
accidents and accident
consequences unknown
Research, development
and design needs
Major breakthrough in
fusion development
needed
In an advanced stage
of development
Possible economics,
deep disposal possible
Added barrier to human
environment and ocean
dilution
Elimination of long
term uncertainty
Elimination of long-
lived actinides
Elimination of long-
lived actinides and
fission products
-------
for thousands to millions of years. It must be determined that
mining and the effects of waste heat will not result in pathways to
the biosphere. Groundwater or radionuclide migration must be absent
or of sufficiently slow rate that with the sorption capability of the
host rock or other geologic media, radioactive materials are not
transported to the environment in biologically significant quanti-
ties. Natural events, earthquakes, vulcanismsr, meteorite impact, or
accidental intrusions by man which would result in the release of the
waste must be of negligibly low probability so as to be acceptable to
society. Numerous studies are being conducted by the Environmental
Protection Agency, the Department of Energy, the Nuclear Regulatory
Commission, the Geological Survey and others to determine the accep-
tability of deep-mined geologic repositories for radioactive waste
disposal. It is not intended nor is it within the scope of this
study to evaluate the acceptability of deep-mined geological reposi-
tories.
Alternate continental geologic disposal concepts may have some
advantages over repositories economically and perhaps in deeper em-
placements of the waste. These options require technical develop-
ment.
Seabed disposal is an attractive alternative in that an addi-
tional barrier exists between the waste and the human environment and
few direct exposure pathways exist. For example, the oceans are not
used for drinking water or for irrigation. The only direct exposure
2-17
-------
pathways are the ingest ion of marine animals for food and some limit-
ed ingestion of marine plants. Ocean dilution of radionuclides which
may escape the repository would also reduce the biological hazard.
Data concerning the containment capabilities of seabed disposal is
not presently adequate to implement this disposal method.
Extraterrestrial disposal removes the waste from the earth and
with proper selection of space destinations essentially eliminates
the uncertainty of future terrestrial contamination. There is, how-
ever, a potential for accidents in which the waste capsules may con-
taminate the earth. The probabilities (risk) and impacts (consequen-
ces) of accidents have not been analyzed. The risk and consequences
can be minimized by design approaches although economics might be
affected substantially, i.e., small amounts of actinides or long-
lived fission products per launch. Further analysis is required
before extraterrestrial disposal becomes an acceptable alternative.
Transmutation of the actinides in fission reactors has been con-
sidered an attractive disposal concept by researchers. The long-
lived actinide radionuclides could essentially be eliminated by this
approach. In the event that fusion reactors become practical, both
the actinides and the long-lived fission products could conceivably
be eliminated by transmutation utilizing the high neutron flux of
these reactors. However, a major technical breakthrough is needed
s
before fusion reactors can be considered practical and fission trans-
mutation requires research, development, and design. Neutron
2-18
-------
absorption and reaction rates (cross-sections) with the long-lived
actinides must be determined. Nuclear reactors must be developed,
designed, and tested for the transmutation process. Further, parti-
tioning, fractionatlon, and fabrication methods must be developed for
the long-lived actinides and fission products.
In the extraterrestrial and transmutation process, It is likely
that there will be short half-life waste that will require either
geologic or seabed disposal. The time period for isolation will,
however, be substantially reduced; from millions of years to perhaps
a thousand years. The uncertainty of future events which might re-
lease the waste to the environment would correspondingly be reduced.
While extraterrestrial disposal and transmutation appear as favorable
concepts, it should be borne in mind that chemical separation and
other processing facilities are not perfect in their operation. Some
fraction of the long-lived radionuclides will remain with the shorter
half-life material for terrestrial disposal. The added operating
facilities will have some radioactive material releases and will
increase the risk for accidents. These factors must also be con-
sidered in radioactive waste management and in the evaluation of
disposal concepts.
2.3 Conclusions
Deep-mined geologic repositories offer the greatest potential as
a near-term approach to final disposal of high-level and transuranic
radioactive waste. In the event that repositories are deemed
2-19
-------
unacceptable for the final disposal of radioactive waste, or if
disposal is deferred for other reasons, then it will be necessary to
place the waste in long-term storage until alternative methods of
disposal are developed and accepted. Storage of spent fuel elements
is probably most desirable since it provides for the greater options
of final disposal with the least economic penalty.
At present, none of the alternatives to geologic repositories
have reached a stage of development to be considered acceptable
methods of final disposal. They do, however, have potential for
development to practical approaches and several would reduce the
uncertainty of long-term containment.
Seabed disposal offers an additional barrier to transport of
radioactive material to biologically active regions and provides
dilution to reduce the biological hazard.
Extraterrestrial disposal and transmutation have the potential
to remove the long-lived radionuclides from the earth and thus reduce
the long-term uncertainty of waste disposal.
The possibility also exists of employing a multiple approach to
radioactive waste disposal; a combiantion of fission transmuation of
actinides, extraterrestrial disposal of selected long half-life fis-
sion products, and geologic or seabed disposal of short half-life
radioactive waste is one example. Whether such an approach is eco-
nomically, technically, or environmentally acceptable remains to be
determined.
2-20
-------
The discussion of disposal alternatives has been primarily
directed to spent fuel from commercially operated reactors. Existing
defense wastes are a special problem in that they exist in forms
which are not readily adapted to further treatment. Extraterrestrial
disposal and transmutation are therefore not likely to be attempted
for these wastes. Accordingly, either geologic or seabed disposal
would be anticipated for the Defense waste final disposal.
It has not been possible in this report to assess the radiologi-
cal health risk of the alternative disposal concepts. Studies are
presently being conducted by EPA, DOE, and others to determine the
long-term health risk of geologic repositories. Similar studies are
required for comparative evaluation of alternative disposal concepts.
2-21
-------
3.0 QUANTITIES AND FORM OF HIGH-LEVEL AND TRANSURANIC WASTE
Radioactive waste may originate from a variety of sources:
certain mineral processing activities; medical, industrial, and sci-
entific radioisotope applications; nuclear power reactors; and U.S.
Defense waste programs. This report deals with wastes from the spent
/
fuel of nuclear power reactors and certain wastes from the U.S. De-
fense program. These wastes pose the greatest hazard to the environ-
/
ment and the long-term welfare of society. They are characterized by
high specific radioactivity and contain elements of atomic number
greater than 92 (transuranium elements). The transuranics are char-
acterized by long half-life and high radiotoxicity and are therefore
of particular concern.
At the present time, the Defense waste represents the greater
bulk of waste for disposal. By the year 2000, the commercial waste
will, however, far exceed the defense waste in total radioactivity
for treatment and disposal, even if only the low projections of
installed nuclear power are realized.
3.1 Present and Projected Quantitites of Waste
3.1.1 Existing Waste
The estimated inventory (in 1977) of spent fuel from operating
commercial nuclear reactors is about 2,500 metric tons (MT). This
spent fuel is primarily stored at the reactor sites. In addition,
there exist approximately 77 million gallons of Defense program
high-level waste stored at government facilities at the Hanford
3-1
-------
Reservation, Savannah River Reservation, and Idaho National Engineer-
ing Laboratory.^ A backlog of 1800 MT of Defense program-related
spent fuel has been accumulated from the Hanford N Reactor for pro-
cessing and an additional amount of 400 to 900 MT/year is expec-
f\
ted. A small amount of high-level radioactive liquid waste is
stored at the now shutdown Nuclear Fuel Services Plant at West
Valley, New York.
The Defense waste, which represents the bulk of the present
waste, exists in several different forms: solidified calcine powder
(Idaho); salt cake; sludge; residual liquor (Hanford and Savannah
River); and capsules of strontium and cesium (Hanford).
The quantities of fission products, actinides,* and contained
/
sodium for each of the U.S. Government high-level waste storage sites
are shown in Table 3-1. ^ The sodium is non-radioactive but is used
in Defense waste programs in the form of NaOH to neutralize the ni-
tric acid used in the treatment of irradiated fuel at Hanford and
Savannah River. This permits'the use of less expensive carbon steel
tanks. The sodium is important, however, in that it complicates
further processing for waste disposal as noted below.
The radionuclide content of the Defense waste is not well known
(the program dates back to the 1940"s) but representative composi-
tions are given in Tables 3-1I, 3-III, and 3-IV.
Actinides are elements of atomic number 89 or higher. They include
the radioactive decay daughter products of the transuranium
elements. Some of these isotopes and their daughter products are
hazardous alpha radiation emitters.
3-2
-------
TABLE 3-1
SUMMARY OF DEFENSE WASTE QUANTITIES
Site
Hanford
Savannah River
Idaho
TOTALS
Millions
of
Gallons
51
22
3
76
Radioactivity, Ci
Sr-90
Plus
Cs-137
2.4 x 108
8
2.6 x 10
7
4.4 x 10
8
5.4 x 10
Total
FP1
2.5 x
3.2 x
8.0 x
6.5 x
s
io8
8
10
7
10
8
10
Uranium
7.1 x
•v 8 x
^ 1 x
7.1 x
IO2
o
10"
o
10U
2
10
TRU
1.4 x
7.4 x
1.0 x
8.8 x
IO5
5
103
3
10J
5
103
Total
FP's
60
57
9.2
130
Wt. (MT)
Uranium
900
•v50
**• 2
952
TRU
.52
.44
.02
.98 .
Na
Content
66,000
30,000
30
96,030
Source: Arthur D. Little, Inc., estimates, Reference 3
-------
TABLE 3_n
INVENTORY OF MAJOR FISSION PRODUCTS AKP ACTISIDCS
IN HANFORO HIGH-LEVEL WASTES DECAYED TO
1990
Radioactivity (Cl)
Udlonuclida
nitlon Produces:
H-3
C-14
Sr-90
Zr-93
Tc-99
C4-113a
Sb-12S
Sb-126
1-129
Cs-137
Ca-144
Pm-147
Sa-151
Eu-152
Eu-154
Eu-155
Actinides:
U-233
U-235
U-236
Np-237
Pu-238
Pu-239
Pu-240
Pu-241
Am-241
Sale
Cake Sludge
* *
2.0 x 106 4.5 x 107
* 6.9 x 103
* • *
* 5.0 x 103
* 2.0 x 104
* 9.6 x 106
* *
5.0 x 106 5.0 x 10S
• 9.9 x 106
* 1.0 x 106
• 1.4 x 106
* 1.5 x 103
* 7.3 x 104
* 7.4 x 104
* 4.0 x 102
* 1.3 x 101
* 3.0 x 102
* 1.0 x 102
4.0 x 102
2.1 x 104
5.2 x 103
6.0 x 104
5.0 x 104
Residual
Liquor
1.1 x 104
6.0 x 105
*
3.1 x 104
*
*
*
4.7 x 101
1.8 x 107
*
*
*
*
*
•
*
*
*
*
*
*
*
*
*
Capsules Total
fc
1.1 x 104
<1.6 x 104
5.8 x 107 1.06 x 108
6.9 x 103
3.1 x 104
5.0 x 103
- 2.0 x 104
9.6 x 10°
4.7 x 10l
1.0 x 108 1.3 x 108
9.9 x 106
1.0 x 106
1.4 x 106
1.5 x 103
7.3 x 104
7.4 x 104
4.0 x 102
1.3 x 10l
3.0 x 102
1.0 x 102
4.0 x 102
2.1 x 104
5.2 x 103
6.0 x 104
5.0 x 104
*Contain» tract quancitias of thasa Isotopa*.
fota; Daughtar nuclidas aoc lilted; curia value* are for parent nuclide -only.
Source; ERDA-76-43. UC-70, "Alternatives for Managing Wastes froa Reactors and
Post-Fission Operations in the LWR Fuel Cycle," Volume 2, U.S. Energy
Research and Development Administration, May 1976.
3-4
-------
TABLE 3-III
RADIONUCLIDE CONTENT*
SAVANNAH RIVER HIGH-LEVEL WASTES (1985)
Radionuclide*
Fission Produces: Total Activity (Ci)
Sr-90 1.3 x 108
Ru-106 1.8 x 106
Cs-137 1.3 x 108
Ce-144 1.1 x 107
Pm-147 4.6 x 107
Sm-151 4.2 x 106
Accinides:
Pu-238 6.0 x 105
Pu-239 2.4 x 104
Am-241 6.0 x.104
Cm-244 6.0 x 104
*L)aughter nuclides in decay chains are noc listed. Curie values are of
important nuclides only.
Source: Alternatives for Long-Term Management of Defense High-Lev*!
Radioactive Waste—Savannah River Plant. ERDA 77-42/1, U.S.
Energy Research and Development Administration, May 1977'.
3-5
-------
TABLE 3-IV.
TYPICAL COMPOSITION OF CALCINED SOLIDS
IDAHO CHEMICAL PROCESSING PLANT
Composition. We.
ZrO
HgO
Ca as CaF«
Fission product and other
oxides, fluorides
Nitrogen as NJD.
Aluminum ~
(Non- fluoride)
Waste
85
0
1
0.3
2.4
0
4.8
Zirconium
(Fluoride)
Waste
8
34
0
0.9
0.1
54
0.5
Bulk Density
1,100 kg/m3 1,600 kg/ta3
Based on: Alternatives for Long-Tera Management of High-Level Defense
Waste—Idaho Chemical Processing Plant. Preliminary Draft,
May 1977. Kearney, M.S., & Walton, £.D., Long Tera Management
of AEC/I3DA Generated High-Level Radioactive Waste, AlChZ
Symposium Series 154: 45-51, IS76. Reference 3.
3-6
-------
3.1.2 Projected Quantities of Waste
The quantities of-future waste are primarily dependent upon the
growth of the commercial nuclear power industry. Estimates of
installed nuclear power range from less than 400 to 1000 GWe in the
year 2000. An estimate by S.M. Stoller Corporation is shown in
Figure 3.1.3 in 1975, the Energy Research and Development Agency
(ERDA), now incorporated into the Department of Energy (DOE), pro-
jected nuclear generating capacity on a low growth scenario to 380
GWe in the year 2000.
Actual quantities of nuclear waste will be dependent upon a
number of factors as previously noted; however, projections for a
nominal case of 700 GWe to about the year 2010 are given in Figure
3.2. For reference purposes, light-water reactors typically dis-
charge 25.5 MTH/GWe-year.* Included with this discharge is approx-
imately 0.9 MT of fission products and 0.26 MT as transuranics (TRU).
For the total lifetime (30 years) of 700 GWe added capacity,
5.36x1O6 MTHM of spent fuel would be discharged.3
The estimated range of total U.S. high-level waste to the year
2010 is shown in Table 3-V.3
The significant radionuclide composition of the commerical waste
will vary with age. Table 3-VI presents the significant radionu-
clides (greater than 1 percent of total activity) for time periods up
*1130 MWe PWR, 30-year lifetime, 70 percent capacity, 33 percent
thermal efficiency. MTHM means metric tons (1000 kg) of Heavy
Metal.
3-7
-------
500
o>
ra
>
Q.
ra
O
TD
_0>
1
300
200
100
50
0
B.
SMSC Assessment of the
True Economic Potential
of Nuclear Power
D. Nominal Projection
E. Schlesinger's
Year 2000
Estimate
A.
Current Utility
Timetable for
Existing Projects
C. ERDA Low-Growth
Scenario (9/76)
I I
I I
J L
77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 00
Year
Source: The S. M. Stoller Corporation. 6/17/77.
Reference 3 FIGURE 3-1
U.S. NUCLEAR POWER GROWTH PROJECTION
-------
I
10'
Product*
X
X
X
TRU
Socnt
1980
198S
1990
1995
3000
10*
a
I
o
to4
103
Nott: W«it» Conum B««d on Ttn- Y wr 0*CJy Tim*.
FIGURE 3-2
PtBSPtCTtVg ON TX€ BUILDUP OF SPSNT FUEL AND ASSOCIATED
HIGH LEVEL WASTES VS. TIME (NOMINAL GROWTH CASE. TWRCWAWAY CYCLE)
Source: Arthur D. Little Inc., Reference 3
3-9
-------
TABLE 3-V
PRINCIPAL RADIONUCLIDIIS IN WASTE—
THROWAWAY CYCLE
NUCLIDE
H-3
C-14
Fe-55
Co-60
Nl-59
Ni-63
X.r-93
Nb-93m
All Others
TOTAL
H-3
Kr-85
Sr-90
Y-90
Zr-93
Nb-93m
Tc-99
HALF-LIFE
© •
12.26y
1 5730y
2.60y
5.26y
8 JL IflAy
92y
1.5 x 10 y
13. 6y
I2.26y
10.76y
27. 7y
64.01)
1.5 x 10 y
13. 6y
2.12 x 10 y
RADIOACTIVITY AT VARIOUS DECAY TIMES. Ci/MTHM (7
10(1) YEARS
10(2) YEARS
10(3) YEARS
10(4) YEARS
10(5) YEARS
)©
10(6) YEARS
HULLS
* 1.07(-1)
1.52(-2)
1.69(2)
2.52(3)
2.36(2)
—
2.93(3)
4.16(2)
5.98(3)
6.00(4)
6.00(4)
6. 69 (-4)
1.50(-2)
1.66(0)
1.20(0)
—
1.22(2)
2.61(0)
6.52(3)
6.52(3)
1.35(-2)
1.36(-1)
5.52(-2)
6.09(-2)
0.03(0)
1.92(0)
FISSION
—
1.86(0)
1.36(0)
.1.43(1)
4. 54 (-3)
1.52(0)
5.50(-2)
5.79(-2)
0.02(0)
1.65(0)
PRODUCTS
—
1.86(0)
1.86(0)
1.38(1)
—
6.97(-l)
5.28(-2)
5.28(-2)
0.03(-1)
8.06(-1)
—
1.78(0)
1.78(0)
1.03(1)
—
3.48(-2)
3.48(-2)
0.07(-2)
7.03(-2)
—
1.18(0)
1.18(0)
5.44(-l)
UJ
-------
TABLE 3-V (Cont.)
PRINCIPAL RADIONUCLIDES IN WASTE-
TIIROWAWAY CYCLE
Ul !/*• Y f\I7
NULL IDE
Tc-99
Pd-107
Sn-126
Sh-126
Sb-126m
1-129
Cs-134
CH-i:i5
Cs-137
Ba-137m
Pm-147
Sm-151
Eu-154
All Others
TOTAL
Fb-209
Pb-210
Pb-214
Bi-210
Bi-213
HALF-LIFE
©
2.12 x 105y
7 x 106y
50m
7
1.7 x 10 y
2.046y
3.0 x 106y
30. Oy
2.554m
2.62y
87y
16y
3.30li
20.46
26 . 8m
5.013d
A 7m
RADIOACTIVITY AT VARIOUS DECAY TIMES, Ci/KTIIM(T)
10(1) YEARS
10(2) YEARS
10(3) YEARS
10(4) YEARS
10(5) YEARS
©
10(6) YEARS
FISSION PRODUCTS (concluded)
__
—
—
—
—
—
9.18(3)
—
8.64(4)
8.08(4)
7.87(3)
—
— —
0.10(5)
3.20(5)
__
—
—
—
—
—
—
—
1.08(4)
1.01(4)
—
5.68(2)
—
0.01(4)
3.46(4)
1.43(1)
—
5.60(-1)
5.60(-1)
5.55(-l)
—
—
2.23(-l)
—
—
—
4.37(-l)
— —
0.06(1)
2.69(1)
1.38(1)
—
5.26(-l)
5.26(-l)
5.2K-1)
—
—
2.23(-l)
—
—
—
—
— —
0.07(1)
1.99(1)
1.03(1)
—
2.82(-l)
2.82(-l)
2.79(-l)
—
—
2.18(-1)
—
—
—
--
— —
0.02(1)
1.52(1)
5.44(-l)
1.05(-1)
—
—
—
3.62(-2)
—
1.77(-1)
—
—
—
—
— —
—
3.21(0)
ACTINIDES AND DAUGHTERS
—
—
—
—
_^
—
—
___
—
—
_—
—
—
—
—
—
4.19(-1)
9.80(-1)
9.80(-1)
9.80(-1)
4.19(-1)
9.40(-1)
4.70(-1)
4.70(-1)
4.70(-1)
9.40(-1)
to
I
-------
TABLE 3-V (Cont.)
PRINCIPAL RADIONUCL1DF.S IN WASTE—
TIIKOWAWAY CYCLE
NUCL1DE
111-21 4
Po-210
Po-213
Po-214
Po-218
At-217
Kn-222
Fr-221
Ru-225
Ra-226
Ac-225
Th-229
Th-230
Th- 2 34
Pa- 2 33
Pu-234m
U-233
U-2J4
U-236
U-238
N|>-237
Np-239
Pu-238
Pu- 2 39
Pu-240
HALF- LIFE
©
19.7u.
138. 40d
4.2 x 10-6s
1.64 x l(Hs
3 . 05m
3.23 x 10-2a
3.8229d
4.8m
14.8d
1602y
10. Od
7340y
8.0 x I0''y
24.10d
27. Od
1.0175m
1.62 x 105y
2.47 x 105y
2.39 x 107y
4.51 x l()9y
2.14 x 10&y
2.346cl
86. 4y
24,390y
6580y
RADIOACTIVITY AT VARIOUS DECAY TIMES, Ci/MTIIM G.
^+^1
10(1) YEARS 10(2) YEARS 10(3) YEARS 10(4) YEARS
10(5) YEARS
)©
10(6) YEARS
ACTINIDES AND DAUGHTERS (continued)
— _ _ '—— _ _
— — — —
__ __ — __
_;_ — — —
— — — —
— — — —
— — — —
— — _:_ —
—
—
— — — —
— — — —
— — — —
— — — —
— — — —
— — — —
^
— — — —
— — — —
— — — —
— — — —
6.94(0)
2.19(3) 1.69(3) — —
3.30(2) 3.22(2) 2.52(2)
4.87(2) 4.44(2) 1.77(2)
9.80(-1)
9.80(-1)
4.10(-1)
9.80(-1)
9.80(-1)
4.19(-1)
9.80(-1)
4.19(-l)
4.19(-1)
9.80(-1)
4 . 19(— 1)
4 . 19 (—1)
9.78(-l)
—
1.19(0)
— •
4.18(-1)
1.53(0)
—
—
1.19(0)
—
—
1.98(1) '
—
4.70(-1)
4.70(-1)
9.20(-1)
4.70(-l)
4.70(-1)
9.40(-1)
4.70(-1)
9.40(-1)
9.40(-1)
4.70(-1)
9 . 40 (—1) ••
- 9 .40(— 1)
4.70(-1)
3. 14(-1)
8.86(-l)
3. 14(-1)
9.40(-1)
4.1K-1)
3.88(-l)
3.14(-1)
8.86(-l)
—
—
—
—
-------
TABLE 3-V (Concluded)
PRINCIPAL RAD10NUCLIDKS IN WASTE—
THROAWAY CYCLE
NUCLIDE
Pu-241
Pu-242
Am-241
Am-243
Cm-244
All Others
TOTAL
HALF-LIFE
©
13.26y
3.79 x 105y
458y
7.95 x 103y
17. 6y
RADIOACTIVITY AT VARIOUS DECAY TIMES,
10(1) YEARS
10(2) YEARS
10(3) YEARS 10(4) YEARS
Ci/MTHM (T)(3
^^/ > — •
10(5) YEARS
)
10(6) YEARS
ACTINIDES AND DAUGHTERS (concluded)
7.95(4)
1.73(3)
1.35(3)
0.08(4)
8.56(4)
1.11(3)
3.86(3)
0.08(3)
6.96(3)
9.24(2) 6.94(0)
0.05(3) 0.08(2r
1.73(3) 4.51(2)
1.45(0)
0.15(1)
4.03(1)
2.80(-1)
0.04(1)
1.73(1)
I
H-«
to
1. Half-lives are reported in seconds (s), minutes (m), hours (h), days (d), and years (y).
2. Numbers in parentheses represent powers of ten.
3. Dashes indicate a value less than one percent of the total in a given column. Tritium and carbon-
14 values are exceptions.
-------
TABLE 3-VI
ESTIMATED RAMCE OF TOTAL DOMESTIC HIGH-LEVEL WASTE BUKPEH*
(CIRCA 2010)
lo
I
Category of Waste*
1*. Commercial Waste*
(throwaway fuel cycle)
Ib. Commercial Waste*
(mixed oxide recycle) ,
2. Waste from Defense
Program* •
Spent Fuel
(MTIIM)
High Level Wastes
Other Associated Waste*
Total
Radioactivity
(Cl)
.11
Fission
Product*
(HT)
TUI
(MT)
lodlne-129
(Cl)
Carbon-14
(Cl)
Miscellaneous
(CD
3.1-7.7 x 10* 1.3-3.2 x 10" 11.000-27.000 3.100-7.700
(Contained In speat fuel)
1Q11
700-1.600 1.3-3.2 x 10* 1.4-l.i x 10* 0.9-2.3 x 10*
6.5 » 10
1)0
1.2
n.a.
•Quantities of commercial wastes based on llfetlne production for range of gross nuclear capacity
additions (400-1.000 CW) keyed to LWR generation. Data are /or 10-year-old wastes. Quantities
and characteristics of non-cosaerclsl waste* keyed to existing Inventory.
"Miscellaneous" consist* of: Cladding hulls, fuel assembly structure, entrapped TRU. and entrapped
fission product*.
Source: Arthur D. Little, Inc., Reference 3
-------
to one million years for a throwaway fuel cycle. The fission prod-
ucts are the major source of radioactivity up to 100-200 years.
Beyond 1000 years, among the fission products, only Zr-93, Tc-99,
•> f
Pd-107, 1-129, and Cs-135 are significant. The neutron activation
radionuclides C-14 and Ni-59 also remain significant after 1000
years' decay. Short half-life daughter products from the radioactive
decay of the long half-life processes are also significant contribu-
tors to the radioactivity beyond 1000 years.
Plutonium and americium are the primary radioactivity sources
from about 200 to beyond 10,000 years. Past 100,000 years, the
actinides daughter products become significant contributors to the
radioactivity source. The alpha-emitting actinides are, of course, a
potential major health hazard throughout their lifetimes.
3.2 Form of the Waste for Disposal
The form of the waste for disposal is dependent upon the policy
decision regarding reprocessing and the disposal option ultimately
selected. The waste form can, however, be generically considered to
be one of three types: 1) spent fuel; 2) solidified and packaged
residue from reprocessing; 3) solidified and packaged-partitioned and
fractionated waste.
\
3.2.1 Spent Fuel
Spent fuel may be treated in several ways in preparation for
disposal. The fuel elements, following a period of aging to facili-
tate handling and to reduce the radioactivity and heat generation,
3-15
-------
would be encapsulated. It is also probable that portions of the fuel
assemblies, i.e., nozzles, end boxes, etc., would be separated from
the remaining hardware to reduce the total mass and volume. The fuel
assembly hardware is initially contaminated with fisqion products
and transuranium elements. A proposed standard requiring materials
contaminated with greater than lOnCi/gm to be disposed of in a
Federal repository may result in this material requiring the same
disposal as solidified high-level waste, unless advanced methods of
decontamination and transuranic element removal are developed.
The fuel elements could be melted and recast in a form which
facilitates handling and disposal. This later option could be par-
ticularly important for disposal options such as seabed disposal
where a specially formed waste capsule may b°. required. In the case
of melting the fuel elements, consideration must also be given to the
collection and disposal of volatile compounds that will be released.
Relatively long-lived volatile radionuclides such as 1-129, C-14,
Kr-85, and tritium could be released.
3.2.2 Reprocessed Waste
Where reprocessing is performed to recover uranium, uranium and
plutonium, or uranium-plutonium-thorium, the aqueous raffinate will
be further treated to form a solid waste. Advanced forms of solidi-
fied waste are granularized calcine and glass.
The calcined product is of approximately the same volume as the
liquid waste. The vitrification of waste requires the addition of
3-16
-------
borosilicate or phosphate glass. The waste glass form produced from
high-level liquid waste is from 60-80 liters/MTU. The cladding and
hulls of the fuel elements are treated and disposed of separately. If
assumed to be compacted to 70 percent of the theoretical density,
about 60 liters/MTU would be formed.^
The form of the Defense waste for final disposal has not as yet
been specified. The INEL waste is at present a solid calcine and
could readily be converted to the higher leach resistant glass form.
The Hanford and Savannah River waste, however, has a high sodium
content which makes the conversion to glass more difficult. In order
to keep the Na content of the glass below 10 percent to facilitate
conversion to glass, 10*> metric tons of glass waste would be
produced. If the Na is removed, the limiting factor is the uranium.
At 40 percent, uranium plus fission product content, 2.8 x 10-*
metric tons would be produced. If the Na and uranium are removed,
the waste for disposal would be only 300 metric tons of glass.•*
The calcine requires packaging to contain the loose granules,
however, all waste forms require containment to protect against ex-
posure of workers and to provide radiation shielding during handling
and shipping. The containment is also necessary to avoid leakage or
contamination in the event of accidents and to provide resistance
against corrosion and leaching of the waste in the disposal environ-
ment. Carbon steel, stainless steel, and titanium have been sug-
gested as waste form encapsulation materials. Titanium has been
3-1-7
-------
projected to have the longest containment lifetime — up to 1,000
years. 3 The type of encapsulation material will be dependent upon
the length of time that- container integrity is determined to be
required. Containment for several hundreds to 1000 years is adequate
for isolation required for the shorter half-life fission products.
However, the encapsulation material cannot assure containment for the
long-lived fission products and transuranium elements. The packaging
and encapsulation material is important in assuring containment for
the period of time during which the waste may have to be retrieved.
In the reprocessing of spent fuel, certain volatile radionu-
clides will be released for which collection and immobilization tech-
nologies are under development. The volatile radionuclides of con-
cern are: Kr-85, C-14, 1-129, and tritium. The Kr-85 and tritium
have relatively short half-lives and therefore require isolation from
the environment for shorter periods of time — on the order of a hun-
x
dred years. Possible forms for disposal of these radionuclides are
listed below:
Radionuclide Half-Life Possible Disposal Form
Tritium (H-3) 12.26 y Polymer impregnated concrete
or Polyethylene organic
compounds
Carbon-14 5730 y CaC03 in concrete
Krypton-85 10.76 y Carbon steel pressure vessels
or Zeolite crystal lattice
Iodine-129 1.7x10^ y Barium lodate incorporated in
concrete
3-18
-------
The disposal of these volatile radionuclides is discussed in re-
ference 4.
3.2.3 Partitioned and Fractionated Waste
There are advantages to partitioning and fractionating the waste
to separate the long half-life from the short half-life radionu-
clides. These separate fractions could possibly be disposed of by
more economical methods. Partitioning and fractionation are required
for the transmutation and extraterrestrial disposal methods.
For the transmutation disposal method, long-lived elements would
be fabricated into targets for particle acceleration and fusion re-
actors or into fuel elements for exposure in fission reactors. Par-
titioned waste for extraterrestrial disposal would be encapsulated in
special containers acceptable for space disposal (see Section 6). It
is assumed that the residual material would be solidified in the
*
calcine or glass form as noted above for reprocessed waste.
3-19
-------
REFERENCES
1. Dr. T. English et. al., "An Analysis of the Technical Status of
High Level Radioactive Waste and Spent Fuel Management Systems,"
JPL 77-69, Jet Propulsion Laboratory, Pasadena, Cal., December
1977.
2. "Alternatives for Managing Wastes from Reactors and Post-Fission
Operations in the LWR Fuel Cycle," ERDA-76-43 Battelle, Pacific
Northwest Laboratories, May 1976.
3. "Technical Support for the Radiation Standards for High-Level
Radioactive Waste Management," Task A and B, Draft, Arthur D.
Little, Inc.
4. P.M. Altomare et al., "Assessment of Waste Management of Volatile
Radionuclides," MTR-7718, MITRE Corporation, McLean, Va., May
1979.
3-20
-------
4.0 PARTITIONING AND FRACTIONATION
Radionuclides produced in nuclear power reactors include acti-
nides (caused by neutron capture in the fertile materials) and fis-
sion and activation products. Their half-lives vary over a wide
range—from minutes to millions of years. Current plans are to
treat high-level waste (spent fuel elements or solidified repro-
cessing waste) as a single entity in storage, solidification, and
disposal (temporary or permanent). This procedure may be adequate
for disposing high-level wastes, but othar disposal alternatives
exist which require the waste to be separated into its components—
actinides, fission products, and volatiles. The optimum waste sys-
tem management could consist of several of the disposal alternatives
discussed in this report. If the waste could be separated into frac-
tions which have comparable half-lives, short-lived fractions might
then be placed in deep-mined geological repositories where they would
decay to innocuous levels in times during which isolation could more
reasonably be assured, i.e., thousands of years. Long-lived frac-
tions could be considered for other treatment: transmutation to
short-lived, nonradioactive nuclides or fissile species; extrater-
restrial; or other types of disposal. Initial considerations were
based on the concept of minimizing the long-lived impurity content
of short-lived fractions so that after a period of about a thousand
years, the short-lived fraction would represent no significant
radiological toxicity. The actual percentage of long-lived nuclides
4-1
-------
allowable would be determined by technical limitations and the par-
ticular nuclide, since not all are equally hazardous.
In separating the long-lived nuclides from the short-lived ones,
emphasis has primarily been placed on separating the actlnide ele-
ments from the fission products since these elements not only have
long life-times but are also highly radiotoxic. The chemical separa-
tion of actinides from fission products is generally referred to as
partitioning. In some cases it may be necessary to separate each
individual type of nuclide both chemically and isotopically. For
example, it is often desirable to separate the element curium from
other actinides because of its intense radioactivity. The separation
of individual elements from mixtures is referred to as fractionation.
Partitioning and fractionation are appropriate only for reprocessed
waste. Isotopic separation may be necessary in situations where the
transmutation of a stable or relatively harmless isotope of a given
chemical element tends to augment rather than reduce the radiological
hazard. However, it must be noted that isotopic separation is an ex-
tremely expensive process by currently available techniques.
4.1 Chemical Processes
4.1.1 Spent Fuel Reprocessing
The radiological and chemical releases in partitioning are
related to the chemical processes that are involved. In most cases
the irradiated fuel is first dissolved in HNC>3 and the solution is
fed to a solvent extraction stage where the Pu and U are separated
4-2 •
-------
from the other constituents and subsequently recovered. In most
reprocessing plants, the primary extraction is done by the Purex
process using tributyl phosphate (TBP) as the solvent. The residual
waste solution from solvent extraction contains about 99.9 percent
of the nonvolatile fission products and almost the whole original
actinide content except U, Pu and some Np. The fraction of U and Pu
reaching the waste stream depends on the efficiency of the separation
process. A value between 0.1 and 0.5 percent is considered as a
design objective by present methods, although present recoveries may
be less. In addition to these, there are other chemical impurities
such as organic solvents, nitric acid, and corrosion products from
plant vessels.
The waste is treated to remove organic solvents and then concen-
trated by evaporation, because there is strong economic incentive to
reduce the volume. The segregation of highly active wastes from low-
level wastes and the minimizing of salts in the waste stream are of
particular importance in volume reduction. Highly irradiated fuel
from LWRs produces several hundred litres of waste/MT of fuel pro-
cessed.
The nature of the hazard from the fission product differs from
that due to actinide components because the actinides are, in gen-
eral, alpha emitters, and are a primary health hazard only if in-
gested into the body. The fission products present both internal
and external hazards. The alpha activity is initially dominated by
4-3
-------
curium isotopes, after several years decay, americium, and after
several thousands of years, plutonium become the controlling actinide
in terms of the number of curies. In the very long term (millions of
years), Np-237 and U-238 have the greatest dose impacts.
In actinide partitioning, the main problem is the removal of Pu,
Am, and Cm. It is necessary to maintain plutonium in an extractable
form at very low concentrations because of its very high radio and
t
chemical toxicity and to avoid the possibility of a criticality ac-
cident. With Am-Cm processing, the major difficulty is separating
these elements without generating large amounts of radioactive chemi-
cal wastes. Experience in regard to the operation of radio-chemical
plants which utilize extensive recycle of the waste streams is lim-
ited. Although the optimum process for each actinide has not yet
been definitively established, removal of Cm, Np, and most of the
plutonium by adding an extra extraction cycle to the Purex process
is considered a strong possibility.
A multifaceted waste management scheme would require separa-
tion or partitioning of the high-level waste into its principle
components—actinides and fission products. For a successful util-
ization of the disposal of radioactive wastes by the transmutation
technique, such separation is an absolute necessity. The reason for
this is the different neutronic behavior of actinides and fission
products.
About 99.5 percent of the uranium and plutonium in the spent
fuel of light-water reactors is recovered by present reprocessing
4-4
-------
techniques. The other 0.5 percent is lost to the high-level wastes.
Studies on the feasibility of partitioning actinides from high-level
wastes have been carried out at Battelle Northwest Laboratories, Oak
Ridge National Laboratory, and EURATOM, in Ispra, Italy. Some
specialized techniques are being developed at other laboratories, but
the process developments have not progressed to the stage where it is
possible to determine cost-benefit tradeoffs. The separation pro-
cesses with the greatest potential are solvent extraction, ion
exchange, and precipitation (or some combination of these methods).
4.1.2 Solvent Extraction
Solvent extraction is the most widely used technique because of
a high degree of selectivity and purity of solvents available, the
possibility of continuous operation, and the availability of.a wide
variety of suitable industrial scale extraction equipment with possi-
bility of automation, remote control, high level of productivity, use
of a wide range of concentrations, etc. There are, however, dis-
advantages such as the inflammability and toxicity of extraction
liquids and the possibility of radiation damage to them, thus
reducing their effectiveness.
4.1.2.1 Actinides
Figure 4-1 from a paper by Bond and Leuze*, shows a conceptual
processing sequence for actinide partitioning based on a combination
of modified Purex processing and secondary processing of the
high-level waste.
4-5
-------
DISSOLVER SOLUTION
I
ON
U, Np, Pu RECYCLE
Pu RECYCLE
„ J
t *
'
L PUREX
PLANT
MM II
H U U
C/J JM
Np
r ^
Pu
r
1 PRODUCT
PURIFICATION
ill
U Np Pu
1
f
PLUTONIUM
REMOVAL
\
Am-Cm + R.
REMOVAL
E.
^ SUPPLEMENT
^ EXTRACT I
INTERIM
0-5 yr
WASTE
^ RECYCLE
Am-Cm ^ Am-Cm
+ R.E. ™r», « „ Am-Cra
FROM R.E.
|
^r ^
HIGH LEVEL
WASTE
MANAGEMENT
ARY
ON
AGE
1
I
Am-Cm
*R.E. - Rare Earths
Source: Bond and I.euze, Reference I.
FIGURE 4-1
CONCEPTUAL PROCESSING SEQUENCE FOR ACTINIDE PARTITIONING
-------
There are major radiological and chemical problems yet to be
resolved:
• Recycle of low- and intermediate-level wastes in Purex
• Adequate U and Np recovery by Purex
• Recovery of actinides sorbed on solids and of "inextractable
Pu"
• Adequate Am-Cm removal from waste without greatly increasing
the waste volume
• Actinide recovery from miscellaneous wastes, burnable waste,
cladding hulls, spent ion-exchange resin, HEPA filters, etc.
Careful process control will be necessary to ensure that
actinides are not released along with fission products. For example,
when Purex feed is stored at high temperature, zirconium and molyb-
denum salt crystals are formed which contain up to two percent
plutonium. Also, zirconium hydrolyzes at high temperatures to form
colloids that carry plutonium.
As yet, no simple solvent extraction method has been developed
for partitioning all of the actinides. A multi-step solvent extrac-
tion process based on more than one solvent has the greatest possible
chance of success. The processes for the extraction of U, Np, and Pu
are different from that for Am and Cm.
The Purex process using tributyl phosphate (TBP) has been demon-
strated on a plant scale for the separation of the U, Np, and Pu with
recoveries of up to 99.9 percent, 90-95 percent, and 99.8 percent,
respectively.2
4-7
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Modifying the Purex process for complete separation of Am-Cm
and higher transuranium elements from all fission products does not
appear feasible, but separation above 95 percent of the trivalent
(Am-Cm) actinides and lanthanides by TBP extraction from solutions
heavily salted with metal nitrates has been achieved.3
Solvents with the greatest potential for the partitioning of Am
and Cm are di-ethyl hexyl phosphoric acid, bidentate organophosphorus
compounds, and dibutyl phosphonate. Solvent extraction using biden-
tate organophosphorous reagents for the removal of trivalent acti-
nides and lanthanides from high-level purex waste is being experi-
mented on at the Idaho Nuclear Engineering Laboratory.^
In solvent extraction, solvent additives to improve the degree
of separation will often give rise to excessive amounts of inert
materials harmful to waste processing or disposal. Scientists at
Battelle Pacific Northwest Laboratory are investigating a process
for the separation of Am and Cm from the bulk of fission products
(especially lanthanides) by solvent extraction that does not involve
additives other than HNC^.5
Since the extent to which various pathways to man's environment
reduces the risk due to long-lived nuclides is not completely estab-
lished, permissible concentrations of long-lived isotopes in the
short-lived fraction cannot be defined. Concentrations varying from
one nanocurie to ten microcuries per gram have been studied. At
1CP nCi/g, only americium will be of concern, and the separation
4-8
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factor required is only 6. At 1 nCi/gm, which represents the same
risk as the naturally occurring radioactivity in man's surroundings,
the plutonium separation requirement is about 99.9 percent. Such
removal factors are greater than those attainable.
4.1.2.2 Fission Products
The extractants generally in use for the separation of fission
products fall into groups of organic phosphorous^compounds, amines,
substituted phenols, ketones, etc. The best known extraction process
is the use of di-2 ethylhexylphosphoric acid (HDCHP) and tributyl-
phosphate for the extraction on Sr and the rare earths at ORNL.
The amine group includes primary, secondary, and tertiary amines
and quaternary ammonium salts. The only fission products extractable
by primary amines are Ru, Zr, Tc, and the rare earths. Tertiary
amines used for the isolation of Ru include trialkylamines with chain
lengths of six to nine carbon atoms. Dipicrylamine is used for the
separation of cesium.
The use of ketones has been sporadic, such as the use of a mix-
ture of thenoyltrifluoracetone (TTA) and tributylphosphate in CCl^
for the extraction of Sr. Other extractants such as carboxylic acids
are also in use. For example, naphthenic acid (which is ten times
cheaper than HDCHP) is used in the Soviet Union in connection with
the isolation of Sr and Y from neutral or alkaline solutions and
extraction of Zr, Mb, Ru, Cs, and Pm.
4-9
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4.1.3 Ion Exchange
The ion exchange method has several advantages such as simpli-
city of operation and equipment, and the possibility of using multi-
stage arrangements. There are also drawbacks such as the slowness of
'the process, large volume used for elution, and unsuitability for use
with "uncharged substances or with colloids [e.g., polyantimonic acids
[H3Sb305(OH)8]3 or (H5Sb506(OH)18)].
4.1.3.1 Actinides
Ion exchange methods for actinide separation are still at the
laboratory stage. It has been shown that Am and Cm can be parti-
tioned by the use of two ion exchange steps^, with recovery capa-
bility 99.9 percent or greater. First the lanthanidesi actinides,
and some of the other fission products are sorbed on a cation
exchange resin column and selectively eluted with HN03. The acti-
nides and lanthanides are then separated by cation exchange chro-
matography on a second column. There are some problems yet to be
resolved such as the conversion of actinide-bearing ion exchange
resins to forms suitable for waste disposal, and the treatment of
the waste streams generated in the chromatographic separation.
4.1.3.2 Fission Products
There are many hundreds of cation and anion exchangers being
produced with different selectivities for particular ions. Some ion
s
exchange resins of the organic synthetic type include hydroxyiso-
butyric acid, lactic acid, ethylene diamine tetracetic acid (EDTA),
4-10
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hydrazinediacetic acid (HDA), and hydroxyethyl ethylene diamine
triacetic acid (HEDTA).
Synthetic resins undergo radiation damage accompanied by gradual
reduction in capacity. Inorganic substances (such as hydroxides,
salts of acids with multivalent metals, insoluble ferrocyamids, alu-
mino silicates, etc.) do not have this drawback. Following are a few
well known processes. MnC>2 has been used for the purification of
Pm. Polyantimonic acid ([H3Sb305(OHg]3 has been used as a
selective sorbent for Sr. Zirconium phosphate, which is a well
studied product, is used for the sorption of Cs. Salts of hetero-
polyacids, such as (Mfy^HPMoj^C^O* used in packed columns
easily take up heavy alkali elements. Alumlno silicates, which can
be divided into clays and zeolites, are highly resistant to radiation
damage. Clays, which are cheap and abundant, are used mainly in
v
connection with the treatment of low and medium activity wastes. A
large number of zeolites have been used for Isolating Cs, Sr, Y, Ce,
Ru, and other medium A elements.
4.1.4 Precipitation Methods
Precipitation methods make use of the low solubility of certain
compounds. They date back to the days of Mme. Curie and Hahn and
Meitner and are therefore well established. However, when applied
to high level waste they entail the problem of remote handling of
solids. They may best be used in conjunction with solvent extraction
and ion exchange.
4-11
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Methods for obtaining crude concentrations of Pu, Am, and Cm by
oxalate precipitation have been developed at EURATOM in Ispra, Italy.
Multiple stages of this type of precipitation have resulted in essen-
tially complete removal of the Am-Cm mixture.
The separation of actinides from high-level waste solutions as
hydrous oxides or associated hydroxyphosphates through the hydrolysis
of urea or hexamethylenetetramine is being attempted in several labs
in the U. S. and Germany.** This method, known as homogeneous pre-
cipitation, has the advantage that the reagents would not contribute
to the volume of the high-level wastes. It also avoids the effects
of introduction of a variety of other chemical substances.
The insolubility of sulphates of alkaline earths, oxalates of
rare earths, and of double salts (such as alums) of alkali metals
makes precipitation a very useful procedure for such fission pro-
ducts. For example, the best known method for isolation of Cs is
the precipitation of CsAlCSC^^. 121^0 (cesium aluminum
sulphate). The heteropoly acids with heavy alkali elements form
slightly soluble salts, e.g., phosphotungstic acid ^PW^C^g or
phosphomolybdic acid ^PMo^O^O* Ferrocyanides are another
type of material used to take up alkali metals, especially Cs.
Coprecipitation is used in the isolation of Sr and rare earths
(e.g., Ce, Pm). The former is precipitated with PbS04. The rare
earths are precipitated as a double sulfate with sodium.
4-12
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4.1.5 Individual Nuclides
The separation of 85Kr, 90Sr, 93Zr, 99Tc, 129I, and
s are of particular interest due to their health effects, long
half-lives, and/or difficulties of containment for long periods.
Krypton-85. A spent fuel reprocessing plant with a daily capa-
city of five tons of fuel produces 35,000 curies of °*Kr per day.
It would be desirable to keep dilution of the gas to a mimimum,
therefore, the free space for cutting and dissolving of fuel parts is
kept very small. Because °%r is a noble gas, there is no neces-
sity for chemical separation; physical separation methods include
adsorption on solid materials and in liquids, low temperature distil-
lation and diffusion. . A more simple approach would be adsorption on
activated charcoal or molecular sieves at laboratory temperature.
Strontium-90. The main emphasis for separating and refining Sr
has been on precipitation or coprecipitation methods using a car-
bonate, an oxalate or lead sulphate; ion exchange methods based on
the use of organic resins and inorganic synthetic materials; and
extraction methods involving the use of the di-2 ethylhexyl phos-
phoric acid (HDEHP).
Zirconium-93. The oldest method of separating zirconium is
based on sorption with silica gel and elution with oxalic acid, which
forms a soluble complex with zirconium. Extraction with HDEHP or TBP
is another possibility. Ion exchange on resins with complex func-
tional groups has also been found feasible.
4-13
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Zirconium, along with niobium, is obtained as a precipitate in
alkaline wastes from the Purex process. In acid wastes these
elements are partly found in the solution and partly adsorbed to
solid siliceous deposits from which considerable amounts can be
extracted by leaching.
Technetium-99. The principal starting material for obtaining
technetium is alkaline Purex wastes from which Cs has been isolated.
t
In acidic and alkaline solutions, especially in the presence of oxi-
dizing agents, Tc is isolated as a pertechnate TcO^.. Technetium
can also be a by-product in the preparation of UF5 from reprocessed
uranium. It is separated from UF5 by adsorption in MgF2, and is
then refined by anion exchange or solvent extraction by a tertiary
amine.
Iodine-129. Iodine-129 is a volatile radionuclide released
during spent fuel reprocessing. The use of silver- and lead-
exchanged zeolites for recovery from the reprocessing off-gases and
storage of 1-129 is now being studied at Idaho National Engineering
Laboratory. Both collection and fixation of iodine are accomplished
in the same process. About 1.5 cubic meters of lead-exchanged
zeolite will be required annually to collect the iodine generated by
a plant which reprocesses five tons of fuel per day. Immobilization
of iodine in cement and glass is also being attempted.
Cesium-137. As a well-known gamma- and beta-ray energy
standard, the separation and purification of ^'Cs has been done
4-14 •
-------
for a very long time. Co-crystallation of Cs with
was developed at Oak Ridge in the 1940s. The solution containing Cs
is saturated at 80°C. with ammonium alum and cooled to 15°C. Crys-
tals of this material are separated. At Hanford, Cs is obtained from
alkaline wastes which are passed through a bed filled with alumino-
silicate. Maximum selectivity for the uptake of Cs from a solution
containing NaNO-j or NaNC^ is achieved at low temperatures.
The use of heteropoly acids is well suited for obtaining Cs
from highly acidic waste solutions. The process has been tested in
the U.S., U. K., and France. Cs is selectively absorbed by salts of
multibasic acids of readily hydrolysable elements, including Zr, Tc,
Sr, U, Th, and Ce salts of phosphoric, molybdic tungstic, antimonic
and arsenic acids. Hexafluorophosphate, tet^afluorophosphate, and
hexafluoroarsenate are among the extracts of Cs which have been
tested.
4.1.6 Other Methods of Partitioning
A technique that is being pursued with some success at the
Lawrence Livermore Laboratory is the chemical separation of transplu-
tonium elements from the chemically analogous lanthanides.^ This
technique uses the formation of stronger complexes by virtue of the
farther spatial extent of the 5f electron orbits of actinides in com-
parison to the 4f electron orbits of the lanthanides.
Partitioning of actinide elements from high-level wastes using
laser photochemical separation is being evaluated at the Brookhaven
4-15
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National Laboratory.^ This process involves reactions that a
molecule undergoes subsequent to electronic excitation by a light
quantum. A general survey of the photochemical spectral region is
required to determine the feasibility of introducing light into the
complex process mixtures and to determine whether there are appro-
priate numbers of wavelengths to carry out selective photochemical
reactions. If successful, this technique can be used for the par-
titioning as well as fractionation of the individual actinides and
fission products.
4.2 Environmental and Health Considerations
The full range of the environmental impacts of applying parti-
tioning and fractionation techniques to radioactive waste is not easy
to assess because the techniques are not yet well established. It is
expected that the design and construction of nuclear fuel cycle fa-
cilities using partitioning and fractionation of waste would, at the
earliest, be in the 1990s. The time of implementation is dependent
upon several things: a decision to proceed with spent fuel repro-
cessing, without which partitioning and fractionation cannot occur;
the establishment of a need, for example, the commitment to a dis-
posal concept requiring this partitioning of waste; and the rate to
which research and development is funded.
The implementation of a partitioning and fractionation technol-
ogy will be dependent upon the balance of the positive and negative
impact on the environment and the health effects. Advancement of
4-16
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this technology is necessary for certain alternative radioactive
waste disposal methods, in particular transmutation and extrater-
restrial disposal. These alternative disposal methods are positive
contributions to the extent that, singly or in combination, they
reduce the risk to the environment and society both for the present
and future generations.
Partitioning and fractionation will increase the steps in the
handling of radioactive waste and thus will increase the radiological
risk. As an adjunct to partitioning end fractionation of waste:
• The total volume of waste to be handled increases due to
the chemical process involved;
• The quantities of low-level radioactive waste and contami-
nated facilities and equipment to be treated and disposed
of increase;
• There are usually some small releases of radioactive
materials and pollutants to the environment;
• There is an increased risk in occupational exposure of
workers;
• Additional transportation with associated risks may be
required;
• The potential for accidents will be increased.
Quantification of the potential environmental and health ef-
fects is not possible with the information available and estimation
of these effects is not within the scope of this study. However,
it is reasonable to assume that the impacts would be less than those
from spent fuel reprocessing.8,9,10 jn tjje context that reproces-
sing will be acceptable after consideration of environmental, health,
4-17
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and political factors, it can be anticipated that partitioning and
fractionation will also be acceptable. In the final assessments of
the alternative disposal methods, those methods requiring partition-
ing and fractionation must include the associated impacts in the
benefit and effects evaluation.
4.3 Economic Impact
The cost of partitioning high-level waste into a long-lived and
a short-lived fraction will certainly increase the cost of nuclear
fuel processing. The estimate made so far has been preliminary
because many of the techniques are still at the laboratory level.
The cost depends on the degree of separation desired and the number
of elements which must be separated from the short-lived fraction.
The most conservative estimate is that in BNWL-1907,^ where
the cost of separation to an actinide concentration level of 1000
nCi/gm is set at $4/ton of uranium. The corresponding figures for
100, 10, and 1 nCi/g are $1,400, $3,900 and $4,200, respectively.
These figures "probably are significantly low" according to the
authors.
Another cost estimate was made on separating 99 percent of the
actinide elements only from high-level waste. The process was devel-
oped by Koch, et al, in Germany. In this process, the volume of con-
centrated high-level waste per unit mass of irradiated fuel is about
seven times less than that of the feed to the reprocessing plant.
Accordingly, the basic reprocessing cost ($35,000/ton) was reduced
4-18
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by a factor (x)' because of the reduction in plant size. The re-
sulting cost was further modified by comparing the number of process
cycles for partitioning to the number required for reprocessing. The
total cost estimates are as follows:^
Cost/tons
Actinides plus 1% of fission products $ 10,000
Actinides less U + 1% of fission products $ 15,000
Actinides less U + 0.1% of fission products $ 20,000
A still higher estimate for actinide partitioning has been
quoted by Brown and Goldstein.5 it is claimed that actinide par-
titioning cost will be comparable to nuclear fuel reprocessing costs
which will be $324/kg U. Evidently, there is a wide discrepancy
among the three estimates of at least two orders of magnitude. The
conclusion is that at the present time, accurate predictions of the
cost of partitioning are not possible.
What can definitely be said is that actinide partitioning will
increase the cost of electric power and the cost of waste management
research. Additionally, there are other comparable long-term hazards
such as the low-level solid wastes generated at the fuel fabrication
facilities, where 0.5 percent of the processed Pu and U are lost. It
is indeed hard to compute the economic aspects of such problems.
4-19
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REFERENCES
1. W. D. Bond and R. E. Leuze, "Feasibility Studies of the
Partitioning of Commercial High-Level Wastes Generated in Spent
Nuclear Fuel Processing," Annual Progress Report for FY-1974,
ORNL-5012, January 1975.
2. R. E. Burns, et al, "Technical and Economic Feasibility of
Partitioning Hanford Purex Acid Waste," BNWL-1907, Battelle
Pacific Northwest Laboratories, Richland, WA, May 1975.
3. J. M. McKihben, et al, "Partitioning of Light Lanthanides from
Actinides by Solvent Extraction with TBP," DP-1361, E. I. duPont
Nemours and Co., Aiken, SC, August 1974.
4. L. D. MeIsaac, J. D. Baker and J. W. Tkachyk, "Actinide Removal
from ICCP Wastes," ICP-1080, Allied Chemical Corporation, Idaho
'Falls, ID, August 1975.
5. E. J. Wheelwright, et al, "Partitioning of Long-lived Nuclides
from Radioactive Waste—FY 1975 Annual Report,"'• Management of
Radioactive Waste: Waste Partitioning as an Alternative,
Proceedings of NRC Workshop, Seattle, WA, June 1976.
6. R. Forthmann. and G. Blass, "Fabrication of Uranium-Plutonium
Oxide Microspheres by the Hydrolysis Process," Journal of Nuc-
lear Materials 64, p. 275 (1977).
7. V. Kowrim and 0. Vojtech, "Methods of Fission Product Separation
from Liquid Radio-active Wastes," At. Energy Rev., Vol 12(2), p.
215, June 1974.
8. H. C. Burkholder, M.O. Cloninger, D. A. Baker, and G. Jansen,
"Incentives for Partitioning High-level Waste," USAEC Report,
BNWL-1927, Battelle Pacific Northwest Laboratories, Richland,
WA, November 1975.
9. B. Verkerk, "Actinide Partitioning: Arguments Against," LAEA-SM
207/41, International Symposium on the Management of Radioactive
Wastes from the Nuclear Fuel Cycle, Vienna, March 22-26, 1976.
10. Y. Sousselier, J. Pradel, and 0. Cousin, "Le Stockage a tres
long terme des produits de fission," IAEA-SM-207/28.
11. H. G. Koch, et al, "Recovery of Transplutonium Elements from
Fuel Reprocessing High-level Waste Solutions," Report.No.
KFK-1651, Karlsruhe, Germany, November 1972.
4-20
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REFERENCES (Concluded)
12. S. L. Beaman and E. A. Altken; "Feasibility Studies of Actinide
Recycling in LMFBR as a Waste Management Alternative" American
Nuclear Society Annual Meeting, Toronto, Canada. June 1976.
13. S. Raman. C. W. Nestor, and J. W. T. Dabbs; "A Study of the
233y _ 232^h Reactor as a Burner for Actinide Wastes."
Conference on Nuclear Cross-sections and Technology, Washington,
D.C. March 1975.
14. J. W. T. Dabbs; "The Nuclear Fuel Cycle and Wastes: Cross-
Section Needs and Recent Measurements," ORNL/TM-5530, Oak Ridge
National Laboratory, Oak Ridge, TN, August 1976.
15. "High-Level Radioactive Waste Management Alternatives,"
BNWL-1900, Battelle Northwest, Richland, WA, Volume 1, May
1974.
4-21
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5.0 TRANSMUTATION
One of the alternatives being considered for the management
of long-lived radioactive wastes is to transmute them into stable or
short-lived radioactive or fissionable isotopes. If this is feasible,
the quantity of waste containing long-lived radionuclides could be
reduced significantly, and the time required for isolation of the
waste shortened.
5.1 Transmutation Concepts
The process of transmutation is accomplished by any of the
following devices:
• Particle accelerators;
• Thermonuclear or fission explosives;
• Fusion reactors;
• Fission reactors.
Each type of device has to be judged on the basis of certain criteria
including overall energy and waste balance and the rate of transmuta-
tion. A favorable overall energy balance means that the energy
required to dispose of the waste should be less than the energy fur-
nished by the nuclear reactor which produced the waste, preferably by
an order of magnitude or better. A conceivable exception would be
when the era of nuclear fission power comes to an end and there are
other plentiful energy sources available which can be economically
used for the disposal of the fission power wastes. The criterion of
overall waste balance is self evident: the waste disposal program
5-1
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should not create more hazardous waste than it removes. This is not
as trivial as it first appears. TV process of transmutation in some
cases is similar to the original process which created the waste. A
successful transmutation rate would be greater than the natural decay
rate of the nuclide. More precisely, the product of the particle flux
(0) which induces the transmutation and the cross-section (<7) for the
transmutation process should be much greater than the natural decay
constant of the nuclide (A), i.e.,0<7»A.
5.1.1 Particle Accelerators
At least four accelerator transmutation methods are conceivable:
(1) direct bombardment by charged particles of several hundred MeV
energy; (2) coulomb excitation in order to augment the y9-de cay rates;
(3) photon transmutation using electron bremsstrahlung; and (4) use of
neutrons released as a result of spaHation by high energy particles.
5.1.1.1 Direct Bombardment by Charged Particles. The direct
nuclear reaction of charged particles from accelerators is not parti-
cularly attractive for radioactive waste transmutation. Most of the
long-lived fission products are intermediate or high atomic number
nuclei. Proton penetration for such nuclei requires energy of sev-
eral tens or hundreds of MeV. It has been estimated that nuclear
reaction with direct bombardment by charged particles expends at least
five times the energy in transmuting the waste than was acquired in
creating it.^
5-2
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5.1.1.? Coulomb Excitation. Beta-decay from certain metastable
nuclear excited states proceeds more rapidly than that from ground
states. This situation applies only in certain exceptional cases.
For example, the 10.8 year Kr-85 has a metastable state at 310 keV
which decays with a half-life of 4.4 hours. Unfortunately the cross-
section for Coulomb'excitation is so small that the energy requirement
is higher by three orders of magnitude than the nuclear fission energy
which produced the waste.^
5.1.1.3 Photon Transmutation. Electrons accelerated to several
tens of MeV produce a shower of photons from bremsstrahlung, but the
yield of photons is too small and the energy required is found to be
at least two orders of magnitude greater for the actual transmutation
of waste nuclei than the energy produced during the creation of the
waste.1
5.1.1.4 Spallation Neutrons. High energy acceleration with
proton energy greater than 1000 MeV could provide a continuous source
of neutrons by spallation in suitable targets (e.g., Pb-Bi). After
moderation in a suitable medium, thermal neutron fluxes up to 10^-°
n/cm2 sec can be expected and can be used for transmutation. The
energy required to transmute one fission product nucleus such as
Cs-137, Tc-99, or Sr-90 was estimated to be between 23 and 110 MeV.2
Thus this method would at best be marginal in satisfying the energy
balance criterion. With a proton beam power of 65 MW, it is esti-
mated that two spallation accelerators are needed to handle the
5-3
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inventory of the above mentioned isotopes. However, at a flux of
neutrons/cm2 sec, it takes 14 years to eliminate 99 percent of
and 80 years for 137£s<3 j^e radioactive contamination caused
by proton interaction with structural materials, the lead target,
etc., may create more wastes than it can transmute, but these are
expected to be short-lived.
Another possibility with high energy accelerators is to use the
radioactive fission product as the target for the protons. A study
team of the Japanese Industrial Forum has speculated that 85 l^Cs
nuclei could be transmuted per incident proton.1
5.1.2 Nuclear Explosives
Transmutation using fission and thermonuclear explosive devices
has been evaluated as technically feasible.^ The procedure is to
partition the actinides and to lower them into a drilled hole along
with the explosive device, seal the hole, and set off the device. The
neutrons produced in the explosion transmute the waste. It is esti-
mated that an average of 3.5 one-hundred kiloton thermonuclear deto-
nations would be required annually to transmute the Np, Am, and Cm
produced every year in a 1000 MWe light-water reactor. It should be
borne in mind that the fission products resulting from the actinide
transmutation and the unconsumed fissile material of the device will
remain in place along with those resulting from the nuclear explosion.
Transmutation of long-lived fission products from each 1000 MWe LWR
requires more than 11 one-hundred kiloton detonations. The concept of
5-4
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transmutation by means of nuclear explosive devices is not considered
practical because of the inordinate number of explosions needed to
cover the nuclear power capacity of the world.
5.1.3 Fusion Reactors
Fusion reactors potentially have very high neutron flux levels
(10^ - 10*6 neutrons/cm^ sec). The high energy neutrons produced
in fusion reactors can be used directly to cause neutron induced
reactions or thermalized for capture in fission processes.
A study of actinide transmutation in the blanket of a conceptual
thermonuclear fusion reactor has been made by Wolkenhauer, Leonard and
Gore for both deuterium-deuterium (D-D) and Deuterium-Tritium (D-T)
reactions. The flux of the neutrons from the plasma reactions could
be augmented by a factor of 2.5 by having beryllium in the blanket
(using the reacton 9Be + n -»2 4He + 2n), thus, fluxes up to 3xl016
neutrons cm~2 sec~l could be realized, which is about 1000 times
greater than in an LWR. Fluxes of this order of magnitude raise the
possibility of transmuting not only actinides but also fission product
nuclides such as Kr-85, Zr-93, Tc-99, and 1-129 for which there is no
practical way of transmutation using fission reactors (see below).
In addition to capture and fission processes, there are other possi-
bilities such as (n, 2n), (n, 3n) and (n, charged particle) reactions
for nuclides such as Np-237, Pu-237, and Am-234. Since a sustained
controlled thermonuclear reaction has not yet been achieved, use of
5-5
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this technique for waste management has to await a breakthrough in
controlled thermonuclear reactor technology.
If, and when, fusion reactors become commercial, it is very
likely that fission reactors will no longer be built, thus the fusion
reactors will only have to transmute whatever inventory of fission
products and actinides are left. In the long run, transmutation by
fusion reactors may become unnecessary.
5.1.4 Fission Reactors
The suggestion to use neutrons from fission reactor neutrons to
to transmute radioactive waste was made as early as 1964.6 4 study
by Claiborne made at the Oak Ridge National Laboratory is perhaps
the most extensive study of the subject to date. It is the general
consensus of this and later studies that trans-nutation of actinides
in fission reactors is technically feasible. Kubo,° and Kubo and
Rose,^ extended Claiborne's work and have shown that actinide recy-
cling in thermal reactors is not only technically feasible, but is
an attractive waste management concept. In the scheme visualized
by Claiborne, the chemical processing of the irradiated fuel rods is
separated into three parts: (1) 99.5 to 99.9 percent of uranium and
plutonium stored or recycled because of their fuel value; (2) fission
products and approximately 0.1 to 0.5 percent heavy elements; and (3)
99.5 to 99.9 percent actinides other than U and Pu. Uranium and Pu
are then recycled into the fresh fuel by adding uniformly to every
rod of a 3.3 percent enriched UC>2 fuel for a PWR.
5-6
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There are a number of parameters which have a bearing on the
feasibility and effects of actinide transmutation in fission reac-
tors :
• mass and composition of actinides being recycled;
• the rate at which the recycled actinides are fissioned
in the various types of fission reactors;
• the effect of the recycled actinides on fission reactor
criticality and reactivity;
• the effect of the recycled actinides on fuel fabrication,
shipping reprocessing, etc.
It has been estimated that a pressurized water reactor producing
1000 MWyr(e) electric produces about 22 kg of actinide waste after
recovery of Pu and U, of which Np, Am and Cm constitute 70, 23 and 6
weight percent, respectively. These are typical values and the exact
composition depends on the reactor characteristics and the recovery
techniques.10
5.1.4.1 Light-Water Reactors. A pressurized water reactor
using U-235 and U-238 fuel with 3.3 percent U-235 enrichment has
been considered as a typical transmuting reactor in the study by
Claiborne.'
Sustained recycle of the actinides in a pressurized water reac-
tor results in an equilibrium mass of about twice that produced per
year without recycling.^ It has been estimated that a typical ac-
tinide transmutation rate is about 6 percent for each year that the
actinides are in the transmutation reactor.
5-7
-------
The introduction of actinides in the fuel rods affects the neu- '
tronic behavior of the reactor. The infinite multiplication factor
kco, which is the ratio of the neutron production rate to the neutron
destruction rate assuming no leakage (reactor of infinite size), is
one criterion for neutronic behavior of reactors. The effect on kco
is a function of the time spent by the actindde in the reactor, but on
the average the effect is to decrease kco by 1 percent.9 This, of
course, seriously affects the reaction design, the economics and the
uranium resource utilization.
The incorporation of the actinides into the fuel elements can
be done uniformly, or in certain selected fuel rods. The mixture
of the actinide isotopes into the fuel elements will present some fab-
rication problems due to decay heat, gamma-ray dose rate, and neutron
emission rate, thus causing additional fabrication costs.
The increase in uranium enrichment required to achieve the same
energy output as for 3.3 wt percent U02 fuel without actinides is
shown in Figure 5-1 for two strategies. In the first strategy, the
actinides are distributed uniformly in all the fuel rods, and in the
second they are concentrated into every tenth fuel rod. Assuming
one-third of the core is discharged every year (corresponding to an
average burnup rate of 22,000 MWd/ton), the enrichment of the fuel
must be increased from 3.3 to about 3.43 percent for the first strat-
egy. Recycling actinides in ten percent of the rods increases the
demand on uranium enrichment to 3.47 percent on the average. The
reason for the higher enrichment requirement is that higher
5-8
-------
H
Z
W
o
CJ
ro
CM
1
Ed
Cd
§
3.45
3.40
3.35
3.43
3.30
ACTINIDES IN
EVERY TENTH
U02 ROD'
ACTINIDES IN
EVERY U02 ROD
10
NO. OF RECYCLES
Source: Kubo and Rose, Reference 9.
FIGURE 5-1
ENRICHMENT REQUIREMENTS FOR ACTINIDES RECYCLE
5-9
-------
concentration causes additional self-shielding of the resonances,
requiring a larger inventory of each isotope before the burnup rate
equals the production rate. Table 5-1 compares the actinide
inventories for the two recycle strategies. It shows that actinide
inventory is not reduced as much by the recycle of actlnides in a few
rods as in the recycle in every fuel rod.
5.1.4.2 Fast Neutron Reactors. It was recognized early in the
recycling studies that fast neutron reactors would cause a faster
burnup of the actinides than thermal reactors. There are two very
obvious reasons for this: 1) the fission-to-capture ratio* is gen-
erally higher for fast reactor neutrons; and 2) the flux of the fast
reactors is typically 5x10^ n/cm^ sec as against 3x10^ for
LWRs. The combination of flux and cross-section rates results in
higher fission rates of the actinides compared to thermal reaction
•..
(see Table 5-II).
Beaman and Aitken^ tried to determine the equilibrium cycle
condition for a recycle scheme involving one 1200 MWe Liquid Metal
Fast Breeder Reactor (LMFBR) and three LWRs of comparable power,
using only the LMFBR as the transmuting reactor. In their calcula-
tions, they assumed a two-year period for reprocessing and fabrication
between the time of discharge, from the reactor and time of loading in
the LMFBR. The batch stays in the LMFBR for 402 days. Table 5-III
*Fission to capture ratio is the ratio of the number of neutrons
resulting in actinide fission to the number of neutrons absorbed by
the actinide nucleus resulting in isotopic transmutation to another
isotope of the actinide.
5-10
-------
TABLE 5-1
COMPARISON OF ACTINIDE INVENTORIES FOR TWO RECYCLE STRATEGIES
Part A - Actinidea Recycled in All Roda
Recycle No.
0
1
2
3
4
5
6
7
8
9
Part B - Actinides
Recycle No.
0
1
2
3
4
5
6
7
8
9
Np237
521.28
703.92
921.49
993.84
1031.99
1052.15
1062.81
1068.45
1071.44
1073.02
Recycled in One
Np237
521.28
811.78
994.46
1118.27
1192.93
1253.06
1299.96
1334.71
1360.02
1379.70
Am241
64.61
76.41
77.51
77.81
77.95
78.03
78.06
78.00
78.10
78.10
Rod In Ten
Am241
64.61
73.37
74.98
75.85
76.31
76.68
76.96
77.15
77.27
77.35
Actlnide
Am242
.58
.77
.79
.80
.80
.80
.80
.80
.80
.80
Actinlde
Am242
.58
.71
.75
.78
.79
.80
.81
.81
.82
.82
Inventory (Cms/KT
An>243
78.63
100.41
104.67
.105.12
104.94
104.72
104.58
104.49
104.44
104.42
Inventory (Gma/MT
An>243
78.63
103.43
111.98
115.43
116.81
117.66
118.01
110.16
118.22
118.26
of Heavy
Cm242
7.89
9.95
10.15
10.15
10.15
10.15
10.14
10.14
10.14
10.14
of Heavy
Cm242
7.89
9.68
9.90
9.98
10.01
10.04
10.06
10.07
10.08
10.09
Metal)
Cm243
.12
.28
.33
.34
.34
.34
.34
.34
.34
.34
Metal)
Ctn243
.12
.27
.32
.34
.35
.35
.36
.36
.36
.36
Cn>244
22.90
76.66
116.20
139.81
152.96
160.08
163.88
165.90
166.97
167.53
Cm244
22.90
76.55
120.63
152.73
175.55
193.78
203.55
208.71
211.28
212.44
Cm245
1.07
5.75
9.63
12.02
13.37
14.10
14.49
14.70
14.81
14.87
Cm245
1.07
5.96
10.44
13.75
16.05
17.88
18.95
19.57
19.91
20.10
Source: Kuba and Rose, Reference 9
-------
TABLE 5-II
ACTINIDE REACTION RATES IN FAST AND THERMAL REACTORS
(Reactions/sec/Atom)
Oi
M
KJ
Fast Spectrum
Half-Life.
Isotope Years
Np237 2.14 x 106
Am21*1 433
Ara2l*2ra 152
Am21*3 7370
Cm2"1* 17.9
Thermal Spectrum
Fission Capture Fission
Reaction Rate Reaction Rate Reaction Rate
2.2 x ID"9
2.7 x 10-9
4.7 x 10'°
1.39 x 10" *
3.47 x 10"9
1.03 x 10"a
2.35 x 10~B
9.69 x 10 9
4.5 x 10"9
2.77 x 10-9
6.
6.
1.
1.
4.
18 x
18 x
49 x
55 x
02 x
10-
10"
10-
10-
10-
12
10
7
11
10
Capture
Reaction Rate
4.9
1.38
1.33
3.18
5.9
x
x
X
X
X
10
10
10
10
10
-8
-V
-7
-8
-9
'Average Total Flux = 6.93 x 1015 in Core Zone 1
**Averap,e Total Flux = 3.09 x 10II§
Source: Hainan, Reference 11
-------
TABLE 5-III >
ACTINIDE RECYCLE FROM ONE 1200 MWe LMFBR
AND THREE 1200 MWe LWR's -
Cycle No.
2
4
6
8
10
12
14
16
18
20
22
24
26
28
30
Total
Actinidesa
1.47 + 2
2.28 + 2
2.75 + 2
3.02 + 2
3.19 + 2
3.30 + 2
3.37 -1- 2
3.41 + 2
3.44 + 2
3.47 + 2
3.48 + 2
3.49 + 2
3.50 + 2
3.50 + 2
3.51 + 2
Total
Pub
3.05 + 1
4.42 + 1
5.05 + 1
5.36 + 1
5.51 + 1
5.58 + 1
5.62 + 1
5.63 + 1
5.64 + 1
5.66 + 1
5.66 + 1
5.66 + 1
5.66+1
5.66 + 1
5.66 + 1
Total
Actlnides + Pu
1.77 + 2
2.72 + 2
3.25 + 2
3.56 +:i2;
3.74 + 2
3.86 + 2
3.93 + 2
3.92 + 2
4.01 + 2
4.03 + 2
4.04+2
4.05 + 2
4.06 + 2
4.06 + 2
4.08 + 2
Total Actinides
if not recycled
2.26 + 2
4.52 + 2
6.78 + 2
9.04 + 2
1.13 + 3
1.36 + 3
1.58 + 3
1.81 + 3
2.03 + 3
2.26 + 3
2.49 + 3
2.71 + 3
2.94 + 3
3.17 + 3
3.39 + 3
a) Actlnides include: Np, Am, Cm, Bk,Cf
b) Pu results from Np neutron capture or decay of higher atomic
number isotopes
Source: Seaman and Aitken, Reference 12.
5-13
-------
lists the total weight of actinides remaining after a specified number
of cycles, and the total weight which would be accumulated if the ac-
tinides are not recycled. The number of cycles required for equilib-
rium of a particular isotope increases with increasing atomic weight
because of the production of higher atomic number isotopes by neutron
capture in the lower atomic number isotopes.
During their lifetime (assumed to 40 years), the four reactors
would have produced about 3620 kg of actinides; with recycling this
would be reduced to 690 kg, thus reducing the actinide quantities by
a factor of 5.2 over the life-time of the reactors. If the reactors
are replaced by another generation of comparably powered reactors
and the recycling is continued, the equilibrium concentrations will
remain the same and a reduction factor of more than 10 is achieved
over a period of 80 years.
Actinide recycle might affect the transmuting reactor in several
ways. These could include the increase in fissile inventory, reactiv-
ity of the core, and breeding ratio. These effects were explored by
Beaman and Aitken^ by a comparison between the "reactivity worths"
of standard fuel assemblies and target recycle assemblies defined as:
n
' N,
Where N. is the atom density of the it1 nuclide, V. , the average
number of neutrons it emits per fission, ff.ft a. are the one-group
1L 1 cL
microscopic fission and absorption cross-sections in the number of
5-14
-------
reactive materials. The actinides, because of their larger absorp-
tion cross-section in comparison with U-238 (which forms the bulk of
the fuel assembly), have a negative worth, but vo, is relatively large
.*
for the actinides and this almost compensates for the absorption. The
decrease in reactivity worth is only slight for a 50-50 U-238 actinide
mix replacing an equal number of standard fuel assemblies. The worth
of the core can be restored in the "worst" case situation by addition
of plutonium, amounting to about 3.4 percent. Such an increase in
the plutonium and the decrease in U-238 which has been replaced by
actinides causes a decrease in the breeding ratio of the reactor. It
has been estimated that an equilibrium cycle load'of actinides will
decrease the breeding ratio by a rather modest amount of 1 percent.
Actinide recycling can also cause power peaking problems in a
fast reactor. A fuel assembly completely loaded with recycle acti-*
nides can produce about twice as much power as a standard fuel as-
sembly and could cause severe heat transfer and reactivity problems
in the reactor. One way to avoid this problem is to mix the acti-
nide with a dilutent. The most obvious dilutent is U-238, not only
because it is plentiful but because it contributes to the breeding
and minimizes the heat transfer effects. A logical choice is an
assembly of 50 percent U-238 and 50 percent actinides. It has been
estimated that the power output of such a fuel assembly.after the
attainment of actinide equilibrium varies from 8.2 MWth to 9.5 MWth,
whereas the standard fuel assembly produces between 8.2 and 8.5 MWth,
and this is judged to be a reasonable match.
5-15
-------
Table 5-IV, by Beaman and Aitken,12 lists five possible acti-
nide recycle schemes. Each subsequent scheme has a greater safety
margin and involves higher costs than the previous one. The only
exception is Scheme 5, which relaxes the requirement on lanthanlda
fission product separation.
5.1.4.3 Thorium-Uranium Reactors
Light-water reactors using a mixture of U-235 and U-238 are the
major types of thermal reactors that are in commercial use in the
f
United States. Neutron capture by U-238 results in production of Np,
Pu, and higher elements which contribute to the bulk of the actinlde
problem. A possible alternative would be reactors which use Th-232
as the fertile material and U-233 as the nuclear fuel. In such a
reactor, the production of nuclides with mass numbers above 237 is
negligible because of the large number of neutron caputures necessary
to produce them. The recycling of actinides in such a reactor has
been the basis of a study by Raman, Nestor, and Dabbs. The Np, Am,
Cm, and higher isotopes, together with 0.5 percent of the U and Pu
isotopes from a U-235 and U-238 reactor, were considered as wastes
to be recycled in a 1000 MWe pressurized water reactor which uses
the U-233 and Th-232 cycle.
In 60 years, which corresponded to ten recycling periods, nega-
tive buildup gradients were established for all isotopes except the
5550 year Cm-246 and the 2.55 year Cf-252. Both of these are sponta-
neously fissionable materials and therefore require additional care
in transportation and fuel processing.
5-16
-------
ACTINIDE RECYCLE SCHEMES
In
I
Initial
Reprocessing
Remove U, Pu, Np, Am, Cm,
Bk, and Cf from spent fuel
Reprocess as in 1 above
and further remove curium
for storage
Reprocess spent fuel such
that the U and Pu are
separate from the Np, Am,
Cm, 3k, and Cf
Reprocessing spent fuel
such that U, Pu, and Np
are separate from the Am,
Cm, Bk, and Cf
Reprocess as in 3 or 4
above, carrying some of the
lanthanide fission products
with the Am, Cm, Bk, and Cf
Fabrication
Actinide
Irradiation
Fabricate pins ccontaining In all fuel pins of
U, Pu, Np, Am, Cm, Bk and Cf a LMFBR
Fabricate as in 1 above
without curim
Fabricate fuel pins
containing U, and Pu;
fabricate target pins
containing Np, Am, Cm,
Bk, and Cf, and a possible
diluent
Fabricate fuel pins
containing U, Pu, and Np;
fabricate target pins
containing Am, Cm, Bk, Cf,
and a possible diluent
Fabricate as in 3 or 4
Curium allowed to
decay; irradiate
after radiation
levels have fallen
In target pins
initially containing
only Np, Am, Cm, Bk,
Cf, and a possible
diluent
Np irradiated in
fuel- pins; Am; Cm,
Bk, and Cf irradi-
ated in target pins
Irradiate as in 3 or
4
Reprocessing of Pins
Containing Recycled
Actinides
Similar to initial
reprocessing
Similar to initial
reprocessing
i) Reprocess target pins
separately from fuel
pins
ii) Mix material from
target pins with
material from spent
fuel; reprocess in a
manner similar to
initial reprocessing
i) Reprocess target pins
separately from fuel
pins
ii) Mix material from
target pins with
material from spent
fuel; reprocess in a
manner similar to
initial reprocessing'
Reprocess recycle
pins as in 3 or 4
-------
5.1.4.4 Actinide Cross-Sections
The quantitative prediction of various nuclei produced, trans-
muted, and fissioned in reactors is necessary for systematic manage-
ment of actinide wastes. Such predictions are made with the aid of
special computer programs which use as input the relevant cross-
sections for capture, fission, or other processes that are caused by
the neutrons. Lacking detailed experimental values at the present
time, most calculations utilize "effective values" in the thermal,
resonance fast neutron regions.
Several laboratories in the United States have cross-section
measurement programs for various actinide nuclei. The Oak Ridge
National Laboratory High Flux Isotope Reactor has been used to ob-
tain the cross-sections for the heavier actinides in the thermal and
resonance regions. The Idaho Experimental Breeder Reactor (EBR II)
is being used to provide integral cross-section data by the irradia-
tion of purified samples of the isotope for several years and sub-
sequent mass spectrometric and radiometric analysis of the sample
after a certain cooling-off period. The Los Alamos Radiochemistry
Group has also made integral cross-section measurements in critical
assemblies using activation and fission chamber techniques.
Cross-section measurements can also be made with the aid of
accelerators. The electron linear accelerators provide a versatile
pulsed source of neutrons whose energy can be measured to a fair de-
gree of accuracy by time-of-flight techniques. The Lawrence Liver-
more Laboratory Linear Accelerator is being used for cross-section
5-18
-------
measurements in the neutron energy range 0.1 to 30 MeV on several
isotopes of uranium, plutonium and curium.
A detailed program for the measurement of actinide cross-sections
has been formulated at the Oak Ridge Linear Accelerator. Some of the
proposed and current measurements have been discussed by Dabbs.^
One of the chief difficulties with cross-section measurements is the
difficulty in producing isotopically pure samples.
5.1.4.5 Fission Product Transmutation
The significant fission products that have half-lives greater
than 10 years and therefore need storage for more than 100 years in
order to reduce their activity by a factor of 1000 are H-3 (12.33
yrs), Kr-85 (10.73 yr), Sr-90 (29.0 yr), Zr-93 (9.5 x 10 yrs), Tc-99
(2.13 x 1015 yrs), 1-129 (1.7 x 107 yrs), and Cs-137 (30.1 yrs).
Carbon-14 (5730 years) is also present from activation of impurities
of nitrogen in the fuel elements. Of these, tritium and C-14 can be
ruled out as candidates for transmutation because their capture cross-
sections for both thermal and fast neutrons are very small, of the
order of microbarns. Even at a flux of 10*7 neutrons/cm sec, the
transmutation constant is only about 10"*-* sec~^ compared to the
—12 —1
natural decay constant of 10 sec for the relatively long-lived
C-14.
Tc-99 and 1-129 have the highest thermal neutron cross-sections
of the remaining radionuclides of 44.5 and 34.5 barns and effective
fast neutron cross-sections of 0.2 and 0.24 barns, respectively. At
5-19
-------
13 2
a flux of 3 x 10 neutrons/cm sec, reduction of the technetium ac-
tivity by a factor of 1000 would require 165 years, and to 10 percent
would require 55 years, corresponding to an annual reduction of 4.3
percent. Even though fast .reactor fluxes are much higher, the much
lower cross-section makes these time periods even longer. Thus trans-
mutation of long-lived fission products is considered impracticable,
except perhaps with high energy (>GeV) accelerators.
5.2 Environmental and Health Considerations
The topics considered so far concern the burn-out efficiency
for the actinides in fission reactors, but there are other considera-
tions. One of the main results of recycling actinides would be the
augmentation of spontaneous fission activity associated with the fuel.
This, along with the intense activity, is a factor in the handling of
actinides for chemical separation and other processes. The neutron
source strength in irradiated fuel is also important in the design
of shielding and it affects the reactivity status of reactors that
have been shut down (i.e., its closeness to criticality).
Recycling in thermal reactors results in the production of the
spontaneously fissile nuclide Cf-252. Recycling in fast reactors
produces, in addition, the fissile nuclides Cm-244 and Cf-250. The
V
short half-life of these isotopes, namely 18 years for Cm-244 and
13 years for Cf-250, make for high specific activity.
In recycling schemes under consideration, it is often necessary
to fabricate the actinide holding fuel rods without those elements
5-20
-------
whose isotopes have high neutron activity. For example, if curium is
removed from the recycle scheme, neutron sources for (a, n)jreactions
with the fuel are reduced by a factor of 18.5 and spontaneous fission
neutron sources are reduced by a factor of 3100.
Further, there is the problem of highly intense gamma ray emis-
sion from such isotopes as Am-243 and Np-239. Such large dose rates
may necessitate recycling in a special small throughput remotely
maintained facility.
Another potential problem regarding actinide recycling is the
buildup of plutonium isotopes such as Pu-238 which is reprocessed
along with plutonium fuel in discharged fuel assemblies. The high
alpha activity of this nuclide may dictate the maintenance of the
f
fabrication facility for target assemblies separate from-other fuel
assemblies.
Actinide transmutation necessarily requires partitioning the
actinide elements from the fission products, and in many instances
fractionation of individual actinide elements (or at least groups
of them) from other actinides. The techniques for partitioning and
fractionation were discussed previously. One significant feature of
partitioning and fractionation is that they would require additional
radiological protection in the fuel reprocessing plants. As the ac-
tinide elements are recycled in fission reactors, actinides of higher
atomic numbers and masses are produced by the successive capture of
neutrons. These higher elements decay by o, |3, and Y radiation and
5-21
-------
some undergo spontaneous fission, thus increasing the radiological
risks.
The need for nucleav data on the actinide elements includes
those for the measurement of body burden and for the estimation of
internal dose. The details should include P-decay energies, Auger
electron yields, fluorescent yields (X-ray), etc. for each element
produced and the daughter nuclides. All of these depend on the
details of the decay scheme of each nuclide produced, which need to
be well established.
The phenomenon of spontaneous fission has greater radiological
consequences than the other types of decay. If spontaneous fission
occurs 1 percent of the time compared to the other modes of decay,
the resulting dose will be comparable to that from other modes. Over
80 percent of the dose from spontaneous fission will be imparted
to the organ in which the radionuclide is deposited. In the gastro-
intestinal tract, however, the fission fragments do not penetrate the
mucosa overlying the radio-sensitive cells, so in this part of the
human body a significant portion of the dose is imparted by neutrons,
(3-particles and V-rays rather than the_ fission fragments.
From the radiological point of view, short-lived isotopes which
cause the greatest concern are the following:
5-22
-------
Decay mode(s)
241Pu
243
Pu
P.v
242^
244^
244Cm
250,
cf
15 yrs
4.98 hrs
152 yrs
10 hrs
18.1 yrs
3.2 hrs
13.0 yrs
2.65 yrs
Spontaneous fission
cross-section (barns)
if any
1,110
3,000
2,300
3,000
3,750
Implementation of a technology for a transmutation of radio-
active waste will have similar environmental and health impacts as
those wastes for partitioning and fractionation of waste. In ad-
dition, irradiation targets or elements will have to be fabricated,
handled, and transported. Additional facilities and waste management
process steps can be expected to have some effluent releases to the
environment, to increase the occupation exposure of workers, and to
increase the risk of accidents. The transportation of materials from
chemical separation facilities to preparation and fabrication plants
and to and from irradiation facilities will require special consider-
ation to minimize the risk to the general population.
5.3 Economic Impact
Transmutation of actinides, even though technically feasible,
involves economic penalties. There are several reasons for this:
5-23
-------
• With actinide recycling, all uranium oxide fabrication will
have to be remotely handled; cost increases up to five times
have been estimated for remote handling. Such cost increases
can be minimized by recycling in only a small fraction of the
fuel rods, say 10 percent, thus fabricating the other 90 per-
cent without a cost penalty.
• Neutron dose rates of up to 10^ neutrons/sec per ton of
fuel material are realized after a few recycles, primarily due
to Cf-252. The transportation of these materials from the
reprocessing plant to the actinide target facility involves
the cost of heavy neutron shielding. This could be minimized
by having the target manufacturing facility as an integral
part of the reprocessing plant.
• Thesneutronic penalty incurred in the recycling of actinides
has already been discussed. As seen before, the enrichment
in the fuel rods must be raised from 3.3 wt percent to 3.47
percent for the case of recycling in 10 percent of the fuel
rods, which is assumed as the reference for estimating costs.
The estimated annual incremental costs for the transmutation of
actinides are listed in Table 5-V1. The figure of $45 million (1973
dollars) can propagate to an increase in the cost of electricity. One
thousand tons of fuel corresponds to the reprocessing requirement of
33, 1000 MW PWRs per year, which at 70 percent capacity fuel will pro-
duce about 2 x 10^ GWh electricity per year.
As shown by Beaman and Aitken,^ the reduction in the breeding
ratio is very small (1 percent) and the economic penalty is negligi-
ble.
5-24
-------
TABLE 5-V
INCREMENTAL COST FOR TRANSMUTATION OF ACTINIDES
Component Annual Cost/1000 Tons of Fuel*
($ x 106)
Partitioning 10
Fabrication 21
Enrichment j^
Total 45
*Cost in 1973 dollars.
Source: Battelle, Pacific Northwest Laboratories, Reference 1.
5-25
-------
REFERENCES
1. K. J. Schneider and A. M. Platt, High-Level Radioactive Waste*
Management Alternatives, V. 4, BNWL-1900, Battelle Pacific
Northwest Laboratories, Richland, WA, 1974.
2. G. A. Bartholomew, "Spallation Type Thermal Neutron Sources,"
Seminar on Intense Neutron Sources, CONF-660925, September
19-23, 1966, Proceedings TID-4500, p. 637.
3. ERDA-76-43, "Alternatives for Managing Wastes from Reactors
and Post-Fission Operations in the LWR Fuel Cycle," Report
Coordinated by the Battelle Pacific Northwest Laboratories,
V. 4, May 1976.
4. M. Goldstein and E. Nolting, Proposal No. IBR-72 2706, Inter-
national Business and Research, Inc. Proposal to USAEC
January 24, 1972.
5. W. C. Wolkenhauer, B. R. Leonard, and B. F. Gore; "'Transmutation
of High-Level Radioactive Waste with a Controlled Thermonuclear
Reactor" BNWL-1772. Battelle Pacific Northwest Laboratories,
Richland, WA., Sept. 1973.
6. M. Steinberg, G. Wotzak, and B. Manowitz; "Neutron Burning of
Long-Lived Fission Products for Waste Disposal" BNL-8558,
Brookhaven National Laboratory, Upton, N.Y., Sept. 1964.
7. H. C. Claiborne "Neutron-Induced Transmutation of High-Level
Radioactive Waste" ORNL-TM-3964, Oak Ridge National Laboratory,
Oak Ridge TN, Dec. 1972.
8. A. S. Kubo; "Technology Assessment of High-Level Waste Manage-
ment" Sc. D. Thesis Massachusetts Institute of Technology,
April 1973.
9. A. S. Kubo and D. J. Rose; "Disposal of Nuclear Wastes" Science
183 (4118) pp 1205-1211. Dec. 21, 1975.
10. A. G. Croff; "Parametric Studies Concerning Actinide Transmuta-
tion in Power Reactors", Trans. Am. Nucl. Soc. 22 pp. 346-347.
November 1975.
11. S. Raman, "Some Activities in the United States Concerning
Physics Aspects of Actinide Waste Recycling" The Advisory Group
Meeting on Transactinium Isotope Nuclear Data, Karlsruhe
W. Germany Nov 3-7, 1975.
5-26
-------
REFERENCES (Concluded)
12. S. L. Seaman and E. A. Altken; "Feasibility Studies of Actinide
Recycling in LMFBR as a Waste Management Alternative" American
Nuclear Society Annual Meeting, Toronto, Canada. June 1976.
13. S. Raman. C. W. Nestor, and J. W. I. Oabbs; "A Study of the
233y - 232-fh Reactor as a Burner for Actinide Wastes."
Conference on Nuclear Cross-sections and Technology, Washington,
D.C. March 1975.
14. J. W. T. Dabbs; "The Nuclear Fuel Cycle and Wastes: Cross-
Section Needs and Recent Measurements", ORNL/TM-5530, Oak Ridge
National Laboratory, Oak Ridge TN, August 1976.
5-27
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6.0 EXTRATERRESTIAL DISPOSAL
Radioactive nuclear waste launched deep into space without any
possibility of return to earth is permanently removed from our en-
vironment. The long-lived wastes, with half-lives of thousands to
millions of years, may thus be disposed of without concern for the
long-term integrity of their containers. This attractive possibil-
ity has created and sustained the interest in extraterrestrial waste
disposal for the last ten years or more.
The most extensive studies were conducted by the National Aero-
nautics and Space Administration (NASA) with contributions from ERDA
(now DOE) and were published in 1973-74.1»2 These two studies were
performed concurrently and some authors are common to both. Their
work established the technical feasibility of space disposal of
transuranium wastes, estimated the costs, and assessed the safety
implications. The scope of the work was based on utilization of
existing technologies in order to avoid any implication of unreality
or a desire to promote any particular idea. This paper summarizes
the results, updates the cost estimates, and further assesses risks
and benefits.
The results of a more recent study of extraterrestrial disposal
of radioactive waste conducted by Battelle Columbus for NASA are in-
cluded to some extent in this report.-* This latter study is only a
part of several concurrent studies sponsored by NASA. When complete,
this study will provide an updated assessment of the feasibility and
6-1
-------
risk of extraterrestial disposal. Since the assumptions and techni-
cal approach will be more advanced than those of references 1 and 2,
they should be consulted as available.
6.1 Basis of Reference Studies
The studies referred to were based on an assumed nuclear capacity
of 1000 GWe.l»2 This is consistent with the presently projected
upper limit of installed nuclear power in this time period. The
estimated weight of waste accumulated after removal of uranium and
plutonium by the year 2000 was estimated at 9000 metric tons (MT) of
fission products and 1200 MT of actinides. The 1200 MT of actinides
reduces to 300 metric tons if the separation of uranium is complete.
These results compare on the low side to those of reference 4, which
estimates 9,000 to 22,000 MT of fission products and 700 to 1,600 MT
of transuranium products for 400 to 1,000 GWe gross installed capacity
for a mixed oxide recycle (see Section 3.0). The estimates of refer-
ence 4, however, are based on the total waste produced over the
30-year plant life. All of these wastes would not be available for
disposal in the year 2000. Several options for the space disposal of
reactor wastes were considered in the studies:
A. Launching all wastes;
B. Removing fission products, uranium and thorium, and
launching only the transuranimum elements with 1.0
percent, 0.5 percent or 0.1 percent of the fission
products remaining;
C. Same as B, but with 99 percent of the curium removed.
6-2
-------
It was evident in the early studies performed and the more recent
study of reference 3 that an impractically large number of launches—
thousands of flights per year by the turn of the century—would be
required following option A. /Similarly, by the year 2000, approxi-
mately 15,000 metric' tons of spent fuel were estimated to be generated
per year. Launching of this large mass is also considered impracti-
cal. Accordingly, extraterrestrial disposal of spent fuel from the
"throwaway" cycle is also impractical.
Option C was considered because curiura-244 with a half-life of 18
years is responsible, after removal of uraniuim and plutonium, for all
but 15 to 20 percent of the actinide radioactivity and about 10 per-
cent of the heat in 10-year old light-water reactor wastes. Removal
of the curium substantially reduces the heat removal and shielding
requirements thereby allowing an increase in -launch payload and a
corresponding decrease in cost and number of launches. It was assumed
that the spent fuel would be held for at least 10 years, and possibly
much longer, prior to processing. A longer period of terrestrial stor-
age of the waste would, of course, allow the curium to decay and thus
reduce the heating and shielding problems. Realizing that an optium
hold time would actually be used based on costs of holding, encapsula-
tion, and transportation (launch), the studies of curium-244 removal
were not pursued in detail. The limited results of the study indicate
an approximate 50 percent reduction in extraterrestrial disposal costs
if the curium-244 is removed from 10-year old reactor wastes.
6-3
-------
It should be borne in mind, therefore, that the costs presented for
the cases without curium removal may be conservative.
Primary attention in the studies of extraterrestrial disposal
has been given to option B and the percentages of fission products
were treated parametrically in some instances. It was judged that
separation technology would more nearly satisfy the 1.0 percent fis-
sion product content, so primary emphasis was given to this case
although a ten-fold reduction in fission product content (0.1 percent)
could provide a reduction of up to 50 percent in program costs.
Very long-lived fission products such as Zr-93, Tc-99, and the
volatile radionuclides 1-129 and C-14 were not considered for sepa-
ration from the fission products and extraterrestrial disposal along
with the actinides. The iodine fraction of the total waste is approx-
imately 0.1 weight percent and the technetium fraction is even less.
The actinide fraction consisting primarily of neptunium, plutonium,
americium, and curium is approximately two percent. The chemical
form and packaging that would be chosen for iodine and technetium
and other long half-life fission products have not been determined.
As will be evident from cost breakdowns subsequently presented, the
major cost is in the transport of waste to the space destination.
Extraterrestrial disposal of the long-lived fission product wastes
will be costly and will have to be weighed against the advantages of
reduced potential health effects to future generations.
6-4
-------
The waste fraction for space disposal primarily discussed in the
balance of this paper is the separated actinides with one percent of
all fission products remaining, and uranium removed and aged ten years
from reactor withdrawal.
6.2 Space Disposal Concept
The required steps in space disposal are shown in the simplified
diagram of Figure 6-1. Spent fuel is withdrawn from storage, repro-
cessed, and partitioned into fission products and actinides with the
uranium removed. Volatile radionuclides are released during repro-
cessing and the long half-life radionuclides 1-129 and C-14 could be
collected and prepared for space disposal. Uranium may or may not be
separately extracted for re-use. Fission products are assumed to be
prepared and disposed of -by different methods. The transuranium pro-
ducts are processed and encapsulated, then shipped to the launching
site where they are launched for space disposal.
6.2.1 Waste Capsule and Reentry Shield
The capsule and shield must provide the following:
• Integrity for the time of use up to final space disposal
• Safety in ground handling
• Shielding
• Integrity in case of accidents
• Cooling and heat transfer
• Handling and attachments
• Subcriticality
6-5
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SPENT FUEL
STORAGE
VOLATILE •
RADIONUCLIDE£ L
1-129, C-14 ;
_ _ _ -t
I IMMOBILIZATION I
J IN SOLID l—1
I MATRIX !
FUEL
REPROCESSING
WASTE
PARTITION
OX
TRANSURANICS
PREPARATION
ENCAPSULATION
h
RECOVERED
URANIUM &
PLUTONIUM
•
CURIUM
i
i
i
•
!
- •
FISSION
PRODUCTS
1
PREPARATION
AND
TERRESTRIAL
STORAGE
T LONG LIVED I
J FISSION
, PRODUCTS
I PREPARATION J
•
TRANSPORT
TO LAUNCH
SITE
LAUNCH
SPACE
DISPOSAL
FIGURE 6-1
EXTRATERRESTRIAL DISPOSAL PROCESS STEPS
-------
The reference design chosen in the NASA study for the waste
payload is shown in Figures 6-2 and 6-3. The waste is compacted and
enclosed in coated tungsten spheres 3.3mm in diameter. These tiny
spheres are mixed into a matrix of lithium hydride, copper, or alu-
minum for shielding and thermal conductivity. This large matrix is
then compacted and enclosed in successive layers of coated tungsten,
lithium hydride, and stainless steel. Design criteria have included
the following:
• Radiation level of 1 Rem/hr or less at 1 meter
• Low temperatures throughout to avoid material degradation
• Ability to withstand launch fires, explosions, impacts
• Ability to withstand reentry temperatures, and pressures
• Ability to withstand surface impacts and burial
The design which has evolved for a payload to a solar system escape
mission (Figures 6-2 and 6-3) has the following parameters:
• Outside diameter, 1.5 meters
• Outside diameter impact shell, 0.98 meters
• Total weight, 3,270 kg*
• Weight of transuranics, 113 kg
• Weight of fission products, 40 kg
• Weight of reentry shield, 4.5 kg
• Dose, 1 Rad per hour at 1 meter
• Thermal power, 9.2 KW
*Weight capacity of the selected launch vehicle,
6-7
-------
TUNGSTEN CAPSULE FOR
HIGH TEMP STABILITY
VOID
TRANSURANICS AND LiH 50 VOL% Al VOLUME FOR
HELIUM BUILDUP
TUNGSTEN / ACTINIDE
SHIELDING v / OXID£
PARTICLES
STAINLESS
STEEL
ALUMINUM OXIDE COATING
FOR OXIDATION RESISTANCE
LITHIUM HYDRIDE-
ALUMINUM MATRIX
BORON PARTICLES
3.3 Him
FIGURE 6-2
TRANSURANIC WASTE CAPSULE FOR SPACE DISPOSAL
SS IMPACT SHELL
LiH NEUTRON SHIELD
TUNGSTEN GAMMA SHIELD
REENTRY SHIELD
TOTAL PACKAGE WEIGHT: 3270 GK
TRANSURANIC WASTE
IN'MATRIX
Source: NASA TMX-2911, Lewis Research Center, Reference 1
FIGURE 6-3
REENTRY SHIELD AND TRANSURANIC DISPOSAL
PACKAGE FOR SOLAR ESCAPE DESTINATION
o-c
-------
As will be evident in the subsequent discussion on safety, these
precautionary measures provide a safety factor in the event of a
launch accident. In addition, as shown in Figure 6-3, a reentry
shield is required to protect the waste capsule in the event of ac-
cidental high Velocity reentry into the atmosphere from space. The
reentry shield provides for intact reentry of the waste package thus
i
providing for enhanced potential for recovery. The rather sophisti-
3
cated packaging of the waste minimizes the possibility of release on
reentry impact and provides long-term containment in the event the
waste package is not recovered.
The combined weights of the shielding, impact shells, and tung-
sten shields are nearly 2200 kg and the weight of the reentry shield
is about 400 kg. Even modest reductions in shield weight would sub-
stantially improve the waste payload although obviously not on a one-
for-one basis. The cost of encapsulation is of the order of a few
percent of the cost of extraterrestrial waste disposal. It is clear
therefore that cost is no barrier to efficient capsule design.
The features of the design which have been developed analyti-
cally or experimentally are as follows:
o Not breached by pressures of 2400 atmosphere
o Not penetrated by aluminum fragments with speeds up to
500 feet/second
o Not damaged by short term fireballs
o Inner shell contains waste in five minute solid propel-
lant fires (shield is lost)
6-9
-------
• Survives vertical ballastic reentry at 11 km/sec
• May be breached by impact on hard granite but may not
release waste (contained in tungsten protective .lay-
ers)
• Outer shell will rupture if deeply buried in earth but
waste will be contained by inner shell
• In the various accidents to be considered there may be
deformations or loss of shielding which could increase
radiation
Comparison of these features to the accident environments pre-
viously discussed shows the design to be qualitatively favorable.
The encapsulation processes are quite complex. The state-of-
the-art for these processes is in an advanced stage-.of development
from many years of experience in encapsulation of radioisotope heat
sources. Further research and development would be required for the
waste but no fundamental problems are anticipated. Plant facility
designs for encapsulation would be similar to existing facilities
for these operations.
The waste fraction presumed to be launched has the approximate
composition given in Table 6-1. Other long-lived fission products
may be included as previously mentioned. The thermal power and
radioactivity of the actinides from different reactor types are given
in Table 6-II.
There are many other options for.composition of the partitioned
encapsulated fraction that can be considered. The fraction finally
selected will optimize the benefits, risks, and costs. The extensive
6-10
-------
TABLE 6-1
CHARACTERISTICS OF WASTE FOR FINAL DISPOSAL
Material
Li-6
Li-7
Cu
0
Al
H
Np-237
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Am-241
Am-243
Cm-244
Atoms/cc x
11.2
13.8
18.9
4.06
13.5
25.0
1.57
0.0124
0.0552
0.0409
0.00600
0.00391
0.112
0.185
0.0440
1,
0.
8/cc
0.1120
0.1610
,9900
,1080
0.6050
0.0420
'0.6180
0.0049
0.0219
0.0163
0.0024
0.0016
0.0448
0.0746
0.0178
Total g in
Single Sphere
•
6,325
9,092
112,376
6,099
34,164
2,360
34,898
276
1,237
920
135
88
2,530
4,213
1,005
Note: Sphere volume » 56.6 liters
source reference 2
Source: Battelle Pacific Northwest Laboratories, Reference 2.
6-11
-------
TABLE 6-1I
THEKMAL POWER AND RADIOACTIVITY OF TRANSURANICS IN 10-YEAR-OLD WASTE
LWR-U LWR-Pu HTGR LMFBR-AI LMFBR-GE
cr>
I
Thermal Radio- Thermal Radio- Thermal Radio-
Power^a) activity^) Power^3) activity^1*) Power^3^ activity^)
Total Act in ides
Less U 69.9 2,350 1,230 36,900 617 25,000
Curium 60.4 1,727 1,144 32,617 36.7 1,051
Percent of .Total
in Curium 85 73 93 89 6 4
Thermal Radio Thermal Radio
Power'3' activity(k) Power^3' activity'*1)
169 7,140 141 5,530
41.3 959.1 57.5 1,633
24 13 41 30
'a'Thermal power is in watts/MT of U + Th
(^Radioactivity is in curies/KT of U + Th
Source: Battelle Pacific Northwest Laboratories,
High-Level Radioactive Waste Management Alternatives,
Section 8. May 1974
-------
analysis required to determine the most favorable mix has not yet been
performed.
The actual payload, 113 kg of transuranics, requires over 3000 kg
of protective encapsulation, or about 27-1/2 times the payload weight.
Such a small payload margin, if decreased by further design and de-
velopment refinement, could significantly increase the program costs.
However, the conservation assumptions of the studies performed and
the many potential options in design or choice of protective devices,
shielding, and launch operations make it more likely that higher pay-
loads could ultimately be achieved.
Alternative Waste Capsule Designs
There are a variety of waste capsule design approaches. One
such approach is illustrated in Figure 6-4 which shows the reentry
protection and encapsulation for a modern radioisotopic electric
generator heat source. The outer graphite cylinder provides the re-
entry protection and the inner graphite and metallic spheres provide
the radioactive material containment. Extensive testing and analysis
have shown this design to be safe for experimental flights on current
launch vehicles.
238
Each heat source contains about 6 kg of PuO. and the assembly
weighs approximately 20 kg. As many as four of these devices with 24
kg of fuel and 288,000 curies have been launched in a single flight.
Research and development for enhanced safety, reduced weight,
and lower cost heat sources is continuing. One such concept is to
6-13
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SPACER
END CAP LOCK RING
SPHERE LOCK
GRAPHITE AEROSHELL
POST-IMPACT SHELL
FUEL
IMPACT SHELL
RETAINING TRAY -^ W
END CAP
TIE BOLT
LOCK RING
LAMINATED END CRUSH-UP
SPHERE SEAT PLATE
FUEL SPHERE ASSEMBLY
r- ABLATION SLEEVE
COMPLIANCE PAD
LAMINATED END CRUSH-UP
Source: General Electric, Doc. No. 775054206, Reference 5,
FIGURE 6-4
MHW HEAT SOURCE
6-14
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separate the radioactive material into a large number of small con-
tainers called "PADS."6 This concept is illustrated in Figure 6-5.
Some of the potential of total risk may be substantially reduced by
more advanced capsule designs.
6.2.2 Launch Operations
The space shuttle with its upper stages was selected in the NASA
study because it performed the missions of interest at lowest cost.
The costs for all other existing launch vehicles were appreciably
higher. The launch vehicles considered are shown in Figure 6-6. The
following list summarizes the cost of launch vehicles for high earth
orbit or solar orbit destinations.*
Launch Costs, $/kg of Payload
Titan III E/Centaur 4920
Saturn V 4590
Saturn V/Centaur 4390
Space Shuttle/Tugs 2940
y.1 launched vehicles except the shuttle are scheduled to be phased
>ut by the 1980s.
The destination studies in the NASA study were as follows:
• High earth orbits
• Solar orbits
• Solar system escape
• Lunar impact or landing
• Planetary impacts
6-15
-------
30-Mnd 3 Graphite
Reentry Shield and
Impact Shell
Molybdenum
Strength
Member
Weld Lid
in Place
PPO Fuel
encapsulation -»
Container
Mall - Indium or Pt/lr
30-Mod 3 Graphite
Reentry Shield
Source: BNWL-975, Battele Pacific Northwest Laboratories, Reference 6.
FIGURE 6-5
PAD CONFIGURATION
6-16
-------
100 i—
80
60
40
20
A
A
A
Siturn V
Titan IIIE/Centaur
£
£1
LWc
Space Shuttle
a
Space tuglnutte pacUgt
Source: NASA TMX-2911, Lewis Research Center, Reference 1.
FIGURE 6-6
SPACE TRANSPORTATION SYSTEMS
6-17
-------
Solar impacts are not possible without planetary swingbys because
the velocity requirements are so high that present launch vehicles
cannot provide the necessary boost. Planetary swingbys pose the pos-
sibility of contaminating their surfaces in violation of present in-
ternational agreements.^ For the same reason, planetary impacts are
ruled out at the present time. High earth orbits and solar orbits are
less attractive because there is no guarantee that the earth will not
at some time recapture the waste, or portions of it, in the event that
its encapsulation fails. High earth orbits and solar orbits are, how-
ever, more attractive than solar system escape on a cost basis in that
space transportation cost could be reduced by a factor of four or
five.
Lunar impacts or landings offer some potential advantages. Waste
deposited on the moon could ultimately be recovered if that were to
become desirable. The cost of lunar missions could also be attrac-
tive. The moon could be a useful staging point for finally launching
the waste into deep space. However, current international agreements
eliminate the lunar destination. The solar system escape destination
is one which can be considered to permanently dispose of the waste for
thousands to millions of years required and is the mission considered
herein even though it is the most costly. It should be realized that
for the very long time span that is contemplated for waste disposal,
many advancements in launch vehicles, encapsulation, and other tech-
nologies may greatly reduce cost. Studies of extraterrestrial dis-
posal are continuing and specific costs for today's technologies
6-18
-------
should not by themselves be the basis for permanently discarding the
space option.
A typical space shuttle launch sequence is shown in Figure 6-7.
Two such launches would be required for a solar system escape mis-
sion. In one launch, the payload would be an expendable tug upper
stage with the waste capsule, and in the other, a reusable tug. The
two tugs would rendezvous in high earth orbit and fire successively,
accelerating the payload to escape velocity. Such missions are ex-
pected to be routine by the 1980s.
There is at least a daily launch opportunity for the solar escape
mission. It may be targeted to miss planets without difficulty. The
waste will escape the solar system in about twenty years. The number
of shuttle launches per year required to dispose of the ten-year-old
waste is shown in Figure 6-8. Depending on the composition of the.
waste, 100 to 250 launches per year would be required by the year
2000. If the launches are made from the existing launch facility
(Kennedy Center) together with the normal anticipated space program,
a modest expansion of launch'facilities would be required. If other
launch sites are to be considered, substantial expense would be in-
volved in creating a new facility.
6.2.3 Technical Feasibility
The entire technology of extraterrestrial waste disposal is in
the conceptual stage with the exception of the space shuttle, which
is in its development phase. However, the processes of partitioning
6-19
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Solid-fueled rxket-molor
ISRM) burnout and jettison
Launch
Landing
Source: NASA XMX-2911, Lewis Research Center, Reference 1.
FIGURE 6-7
SPACE SHUTTLE LAUNCH-TO-LANDING SEQUENCE-
6-20
-------
CsJ
as
u
ft.
X
u
M
r-
U
su
CO
350,—
300
250
200
150
100
50
FISSION PRODUCTS IN
ACTINIDE WASTE, PERCENT
1.0 \
.1
SOLAR
SYSTEM
-ESCAPE
.OOl/
CURIUM REMOVED
i.o\-
l HIGH
1 DEARTH
\ ORBIT
.ooy
1980
1990
2000
2010
Source: NASATMX-2911, Lewis Research Center, Reference 1.
FIGURE 6-6
NUMBER OF SPACE SHUTTLE LAUNCHES REQUIRED
PER YEAR FOR DISPOSAL OF ONLY ACTINIDES
INTO HIGH EARTH ORBIT OR BY SOLAR SYSTEM
ESCAPE. PRIOR 10-YEAR EARTH STORAGE.
6-21
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and encapsulation are similar to those for fuel reprocessing and heat
source encapsulation and therefore would benefit from a considerable
experience background. The space launch operations are being devel-
oped as a part of the nation's current space program. The launching
f
of radioactive materials has been commonplace in the past decade. As
will be discussed subsequently, the optimization of launch operations
and vehicle and capsule designs can substantially enhance safety.
It is evident that a research and development program of substan-
tial scope and cost for adoption of the space program to radioactive
waste disposal would be required to establish the final practicality
even though it can be considered to be technologically feasible.
Consideration of the magnitude of effort required to complete such a
program would indicate a likely time span of around twenty years to
maturity, although this could probably be shortened if a crash program
were to be undertaken.
6.3 Environmental and Health Considerations
The environmental issues which concern extraterrestrial disposal
of radioactive waste can be divided into two parts: those due to
normal operations, and those due to abnormal events such as accidents
or unplanned events.
6.3.1 Normal Operations
Normal waste extraterretrial operations include partitioning of
waste and encapsulation, terrestrial transport, and space transport.
6-22
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6.3.1.1 Partitioning and Encapsulation. Partitioning requires
the construction and operation of plants similar in size and nature
to the presently constructed fuel reprocessing plant, but differing
in the detailed processes that will be used. These processes will
depend on the ultimate choice of the fractions to be separated, the
chemical composition chosen for the products, and the types of waste
to be processed. Similarity to reprocessing plants leads to the con-
clusion that some chemical and radioactive material releases would be
expected. Some thermal pollution of local water sources is likely and
plant construction and land use will intrude on the local environment.
These factors have been considered for a typical reprocessing plant
in a Draft Environmental Statement and found to have minimal adverse
effects on local environments or populations.?
Encapsulation of the waste involves an operation on a consider-
ably smaller scale than fuel reprocessing or a waste partitioning
operation. Accordingly, during normal operations, no significant
releases would be anticipated and no significant environmental
intrusions would be expected.
6.3.1.2 Terrestrial Transportation. Presumably, terrestrial
transportation would conform to the Federal Code Part I 10CFR7, or
its successor, which prescribes the normal and accident provisions
for protection and transportation of radioactive materials.
Occupational and general public radiation exposure could poten-
tially be higher than that for other disposal concepts in that both
6-23
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partitioned waste for extraterrestrial disposal and residual waste
(fission products) for an alternative disposal require transport
to the disposal or launch site. In the event that a remote island
launch site were to be constructed, sea transport would also be in-
volved (see Section 7.0, Seabed Disposal). No significant effect
on the environment would be expected.
6.3.1.3 Space Transportation. Space shuttle launches at the
Kennedy launch site are expected to approach 50 to 100 a year soon
after the year 2000 and impose some safety hazard even if no radio-
activity is released. Nuclear waste disposal missions could increase
the frequency of launch by factors of 2 to 4 in the next few decades.
Environmental studies by'NASA° have identified several potential
effects which, based on the current traffic models, are thought to
be of minimal significance. These will, however, be more significant
if the number of flights is substantially increased for radioactive
waste disposal. These effects include noise and sonic boom, acidic
rain, slight reduction in upper atmosphere ion concentrations, and
the common local community interactions.
The environmental effects on the upper reaches of the atmosphere
depend on the type of vehicle employed. The chemical effluents can
cause reduction in the local ion concentration in the ionosphere,
thus affecting radiowave propagation. A special type of acidic rain
can occur from the propellant emission of hydrogen chlorine. The
6-24
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ozone depletion contribution for 100 launches is around 0.33 percent
per year.^
The possibility of acidic rain, toxic emissions, launch noise,
and sonic booms would make a remote island site more acceptable than
an established launch, area such as the Kennedy Space Center. The
requirement for a new site will depend upon the number of flights as
affected by the nuclear waste mix, the form chosen for disposal, and
the results of future impact assessments.
The annual energy requirements for materials and propellants for
one hundred space shuttle flights per year have been estimated to
require 4 x 10*3 kilojoules. This represents about 2.8 percent of
the electric power to be generated in the year 2000, assuming an in-
stalled capacity of 638 GWe at an availability factor of 70 percent.
Increasing the launch role will increase the magnitude of these
effects and also introduce the need for additional site development
with some modest construction impact. Additional study would there-
fore be required to assess these factors. No radioactive releases
would occur under normal launch operations.
6.3.2 Abnormal Events
Abnormal events, or accidents, have some potential of occurring
at each stage of the waste handling process; partitioning and frac-
tionation of waste, encapsulation of waste, transportation of waste
and launch, and space transport of waste. A detailed assessment of
the risk and consequences of extraterrestrial disposal has not been
6-25
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performed. Partitioning, fracdonation, encapsulation, and transport
of waste are not expected to be greatly different from operations
that are currently performed in the nuclear industry and U.S. space
programs. To the extent that these current operations are presently
acceptable or will be acceptable for other disposal alternatives, it
can'be assumed that similar operations for extraterrestrial disposal
can be made equally acceptable. The major difference between extra-
terrestrial disposal and other alternative concepts is the launch and
space transport stages. This phase of the extraterrestrial disposal
is of particular concern because of the potential consequences of
failure. The balance is the rather complete removal of the waste
and the corresponding' elimination of risk to future generations.
6.3.2.1 Launch Vehicle Accidents. The typical launch accidents
are summarized.in Table 6-III together with a listing of the resultant
events and requirements.
Accident evaluation is commonly divided into four phases:^
Phase 0 - Prelaunch
Phase 1 - Ignition until the impact point clears the launch
area
Phase 2 - Ascent to parking orbit
Phase 3 - Parking orbit to escape
Prelaunch and launch area (Phase 0 and 1) could involve the fol-
lowing:
• Catastrophic explosion and fire
• System failure while the vehicle is near the launch pad
6-26
-------
TABLE 6-1II
TYPICAL LAUNCH ACCIDENTS
Launch Phase Accident Environment Requirements Comments
Phase 0
Phase 1
Phase 2
Phase 3
Mishandling
of payload
or propel Ian t
Explosion and
fire, vehicle
tumbling and
Impact
Failure to
reach earth
orbit
Failure to
reach escape
velocity
Explosion, propel Ian t fireball,
Impact on pad, liquid residual
fire, solid propellant fire,
Impact of fragments
Similar to Phase 0
Ballistic reentry, land or
water Impact
Extended solar or elliptical
earth orbit, possible reentry
Comprehensive, rigid
ground procedures with
checks and controls
Capsule must not be
breached by overpressure,
light temperature, solid
propellant fire, shrapnel
or debris and burial
Capsule must withstand
reentry temperature and
pressure loada, Impact
loads, submergence, burial
Capsule Inert In space
environment, reentry similar
to Phase 2
Very low probability
Source can be limited by
Integrated design of containment
and launch vehicle and by
operating procedures
Low probability with
shuttle reentry capability
Possible recovery by tug
In some modes
o-
I
N>
-•J
-------
• Guidance and control system failure causing the vehicle
to strike the ground resulting in explosion and fire
Ascent and parking orbit to escape accidents (Phases 2 and 3)
could involve the following:
• Explosions and fire at high altitudes
• System failure resulting in short-lived orbit or
powered reentry
• Maneuver or docking accidents with reentry
Present day and past launch vehicles were usually designed for
unmanned operation and incorporated little or no redundancy. The
Saturn V used for the Apollo missions and some versions of the Atlas
and Titan vehicles used in the earlier manned missions employed lim-
ited redundance. In general, the liquid propellant vehicles have
experienced a mission success reliability in the range of 88 to 100
percent with a median of 94 percent. Thus, six missions out of 100
failed to achieve the objective, but not all failures would result in
loss of the waste. A recent Titan Centaur launch was calculated to
have the following vehicle loss probabilities:5
Phase 0 - 0.0043
Phase 1 - 0.000712
Phase 2 - 0.0355 (with 0.025 land impacts)
Phase 3 - 0.01925
Total - 0.059762
Thus, six out of 100 such launches would be expected to experience
vehicle loss whereas the overall vehicle mission reliability is esti-
mated to be less than 90' percent. A vehicle loss is not normally
6-28
-------
expected to release radioactive material since the waste form and
container can be designed to withstand extreme environmental condi-
tions. Also, an intact waste container may be recovered.
6.3.2.2 Radioactive Waste Releases. Radioactive material could
be released from accidents during every phase of the launch. Assum-
ing that the future containment technology and chemical forms of the
c
waste are at least equal to, and possibly superior to, the technology
238
used for PuCL radioisotope heat sources utilized in present
space programs,prelaunch accidents, i.e., prior to placement on a fuel
launch vehicle, are unlikely to result in radioactive material
releases. The most likely release areas would be the launch area
during Phase 1 accidents and land or sea impact on a worldwide basis
during Phase 2 and 3 accidents. The potenti?! impact areas would be
more closely defined in. the event of Phase 2 accidents through flight
path selection. The rather sophisticated designs for radioisotope
heat sources are unlikely to burn up on reentry and result in atmos-
pheric releases. Less sophisticated, though more economical designs
may be more subject to release of radioactive waste in the atmosphere
in the event of an accident.
As noted previously, containment capsules can be designed to
withstand hostile environments. Therefore, any accident of sufficent
severity to breach a modern container during Phase 1 accidents will
necessarily involve explosion and fire. In many cases, and particu-
larly in the case of the shuttle with its large inventory of hydrogen
and oxygen and solid propellants, the fire will be of such intensity
6-29
-------
that a portion of the waste released will be vaporized. The radiolog-
ical release during this phase will, in all probability, be vapor and
larger particles near the launch site.
Phase 2 and 3 accidents may produce impacts on land or water or,
though less likely, upper atmosphere releases. Modern containers will
not be breached by most impacts. However, some probability exists
that impact on hard rock may cause a potential release. The release
t
in the event of hard rock impacts will consist of a respirable frac-
tion and larger particles. The respirable fraction will become
airborne and will then settle out in accordance with dispersion
mechanisms. If disturbed by natural or artificial events, the parti-
cles may become partially resuspended and be redeposited.
In the event of a reentry of the waste capsule and water impact,
the time and rate of release will be dependent upon the damage sus-
tained by the capsule (fire and/or explosion of vehicle, reentry
he'ating, impact damage, hydrostatic pressure). If the waste capsule
is buried in unconsolidated sediments of low thermal conductivity,
the container may fail as a result of high temperatures.
The assessment of the risk associated with the space launching
of radioactive materials is a complex and difficult task. Each poten-
tial failure mode of the mission must be examined and a probability
determined. Each instigating failure will in turn have branching
probabilities of events that may lead to a resulting accident of suf-
ficient magnitude to lead to the release of radioactive materials.
6-30
-------
A detailed analysis has not been conducted to predict the acci-
dent probabilities and the associated consequences for extraterres-
trial waste disposal.
Although a detailed risk assessment has not been conducted, the
analysis performed for the radioisotope thermoelectric generator de-
signed for space applications provides insight to the risk and quan-
tities of released material that might be expected. For the type of
isotopic heat source shown in Figure 6-4, an analysis was conducted
for a final safety analysis report which presented the probabilities
238 <\
for release of PuO« from affected fuel sphere assemblies (FSA).->
In this analysis, the isotopic heat source contained twenty-four Fuel
Sphere, Assemblies with a total of 7 x 10^ curies of Pu-238. The
potential mission accidental release events and prompt fuel release
probabilities are presented in Table 6-IV.5 The predicted quantities
of release and corresponding probabilities are given in Table 6-V.
When an explosion and fire exist, the vaporized material will be
lifted by the fireball and will have an effective height of release
(Heff) above the launch area as shown in Table 6-V. In the analysis
for this radioisotope thermoelectric generator, the predicted proba-
bilities of prompt release were small (on the order of 10~5 to 10"®)
and the predicted quantities of release were also a small fraction
(10~^) of the total inventory.
The analysis conducted for the RTG is not directly applicable to
radioactive waste disposal. In particular, the ratio of total weight
6-31
-------
TABLE-6-IV
MISSION POTENTIAL FUEL RELEASE EVENTS
Mission
Phase
0 - Prelaunch
1 - Launch
Area.
2 - Ascent
3 -Orbit
Initial Accident
none
A. Explosion and fire
in Centaur
B. Tumbling vehicle -
guidance/control
malfunction
Spacecraft ballistic
re-entry due to launch
vehicle malfunction
Launch vehicle mal-
function resulting in
prompt re-entry or
orbit decay
1. Multiple skip re-
entry
2. All other re-
entries
Mechanism Causing
Fuel Release .
none
1. Spacecraft impacts on concrete
launch pad side-on with RTCPs
hitting first
a. No contact with burning
UTP-3001
b. Contact with burning UTP-
3001
2. Spacecraft impacts on sand near
launch pad side-on with RTG's
hitting first
1. Centaur/SC impacts on concrete
launch pad nose first
a. No contact with burning
UTP-3001
b. Contact with burning UTP-
3001
2. Centaur/SC impacts on concrete
launch pad sldc-on with RTG's
hitting first
a. No contact with burning
UTP-3001
b. Contact with burning UTP-
3001
3. Centaur/SC impacts on sand near
launch pad side-on with RTG's
hitting first
HSA impacts on rock following re-
entry
a. High velocity Impact
b. Low velocity impact
HSA Impacts on rock following re-
entry
a. High velocity impact
b. Low velocity impact
HSA impacts on rock following re-
entry
a. High velocity Impact
b. Low velocity impact
Location
Affected
none
Launch
Pad
Launch
Complex
Launch
Pad
Launch
Pad
Launch
Complex
Ground
Track
28* N to
28* S
28* Nto
28* S
Mission
Probability ol
Fuel Release
— * -
none
•B
5. 0 x 10 *
.7
5.0x10
-8
.8.3x10
-ft
9.5x10°
1.8 xlO"7
-8
1.3x10
— T
2.1x10
2.2xlO*7
_7
2.2x10 \
5.1x10
-10
8.3x10 i°
1.5x10
-S
1. S x 10 I
3.5x10
Source: General Electric, Doc. No. 77SOS4206, -Reference 5
6-32
-------
TABLE 6-V
MISSION PROMPT SOURCE TERM SUMMARY
Mission Phase
Launch Area (1)
Ascent (2)
Orbit (3)
Region
Affected
Launch Pad
Launch
Complex
Ground Track
28e N-28' S
No. of
FSAs
1
1
1
1
1
1
1
1
3
3
11
11
6
2
5
2
3
2
6
2
Probability
1.6 (-6)
4. 6 (-7)
3.1 (-6)
8. 9 (-7)
1.3 (-7)
1.6 (-8)
4.0 (-7)
4. 6 (-8)
1.6 (-7)
4.5 (-7)
2.0 (-6)
4.0 (-6)
8.3 (-5)
3.3 (-7)
3.3 (-7)
1.5 (-6)
2.2 (-5)
9.3 (-10)
9.3 (-10)
4.4 (-9)
Location
air
air
air
Air
air
air
air
air
surface
surface
surface
surface
buried
surface
surface
surface
surface
surface
surface
surface
»eff
Meters
260-3970
105-1695
260-3970
105-1695
260-3970
105-1595
260-3970
105-1595
1
1
1
1
-
1
1
1
1
1
1
1
Amount Released
mCl
<4«r
-
- '
-
-
-
-
-
-
3.2
4.0
14
15
6.6
2.7
6.8
2.2
3.2
2.7
6.8
2.2
Vapor
387
387
183
183
387
387
183
183
_
_
—
-
-
-
-
-
-
-
Total
387^
387 1
183 |
18:J
387"^
387 1
183J
issj
549"^
753J
2620^
2830J
1098
366
915
366
649
366
915
366
Max. No.
FSAs
f12
1 12
v
r 4
/ 4
V.
r
f
-
6
15
6
9
6
15
6
Probability
6.7 (-26)
9. 0 (-21)
1.0 (-12)
4.1 (-12)
9.1 (-6)
6.6 (-5)
—
7.8 (-14)
7. 8 (-14)
3.6 (-13)
7.3 (-11)
3.0 (-16)
3.0 (-15)
1.4 (-14)
Amount Released
mCl
<4i«
-
-
-
6.1
16
—
8.1
20.3
6.6
9.7
8.1
20.3
6.5
Vapor
3012
3012
936
936
—
~
-
• -
-
-
-
-
-
-
Total
3012
3012
936
936
"••
.*;
936
3012
^
,, "^
1098
2745
1098
1647
1098
2745
1098
LO
Source: General Electric, Doc. No. 77SOS4206, Reference 5
-------
to weight of waste may be significant if large quantities must be
transported. The analysis performed for the RTG does indicate, how-
ever, that the risk of prompt release of waste in extraterrestrial
disposal can be reduced to low values. To determine whether this will
lead to an excessive number of required flights, whether the risk is
acceptable, or whether the costs are unreasonable will require further
system design and evaluation.
Regardless of whether the waste capsule survives the accidental
event, it must be assumed to fail prior to the radioactive decay
elimination of the long-lived radioisotopes. Recovery of accidentally
released waste capsules is therefore an important aspect of extrater-
restrial disposal.
6.3.3 Recovery and Contingency Planning
Safety during space launches has been enhanced by the use of
operational procedures that have been developed to counter accidents
that may occur and procedures developed to isolate and recover radio-
active material. Figure 6-9 partially illustrates the sequences of
actions that would be undertaken in the event of a Phase 1 accident.
Should an accident occur, immediate measures, as indicated, are under-
taken to ensure the safety of the launch personnel and the public, the
protection of the environment, and to expeditiously recover the radio-
active material. These practices have been refined and improved and
are presently operational.^
6-34
-------
ON/OFF SITE RAD
ASSESSMENT
RADIOLOGICAL
ASSESSMENT AIRCRAFT
EXPLOSIVES ORD
ASSESSMENT
CAPSULE LOCATED
INTACT
PAD OR LIFTOFF
ABORT
LAND IMPACT
IMPACT CONVOY TO
' SCENE
CONTROL CENTER
INITIATE FIRE
CONTROL
RENDER SAFE
INITIATE CAPSULE
SEARCH
(MOB ILE MONIT. TEAMS }
REMOVE CAPSULE
TO STORAGE
REQUEST ASSISTANCE
FROM SUPPORT
AGENCIES
OCESN IMPACT
OCEAN
RECOVERY
PLAN
FIRE ASSESSMENT
MEDICAL ASSESSMENT
FLIGHT HARDWARE
ASSESSMENT
CAPSULE LOCATED
RUPTURED
SECURE AREA
INITIATE CONTAM.
CONTROLS
EVALUATE
CONTAMINATION
LIMITS
REMOVE DEBRIS TO
STORAGE
Source: Manned Space Craft Center, Houston,
Reference 10.
FIGURE 6-9
FINAL
DECONTAMINATION
RADIOLOGICAL RECOVERY SEQUENCE
6-35
-------
If an accident should occur during Phase 2 or 3, return to the
earth would be on land or water remote from the launch site. In
either case, there is a high probability that the capsule or capsules
can be located through a combination of worldwide tracking during the
descent, signal devices in the payload, and aircraft search and detec-
tion devices. Such aircraft are already available and have been used
in several accident situations. Water recovery is possible and has
been accomplished to a substantial depth, but not at all depths. The
capsule design must therefore prevent the catastrophic release of
radioactive material in the deep ocean.
Current detection and recovery capabilities are scaled to very
rare occurrences of emergencies. It is likely that -substantial en-
hancement of these operations would be necessary in the .event that
space waste disposal is used.
6.3.4 Shuttle, Waste Capsule Integration
There are severaL significant implications to be drawn from the
information available. Safety is very strongly determined by inte-
grated vehicle characteristics, encapsulation techniques, and opera-
tional activities. Experience has shown that the proper combination
of these can substantially reduce the risk of space flight operations.
Research and development are continuing and further enhancement of
safety can be expected in future missions.
The space shuttle and tug combinations represent the most ad-
vanced state-of-the-art of launch vehicle design presently known.
6-36
-------
It is anticipated that catastrophic failures would be substantially
less probable than for former launch vehicles. The combination of the
shuttle's ability to recover safely from many previous accident situ-
ations, its redundant systems, and the presence of a pilot with
capability to take remedial actions will contribute to a reduced prob-
ability of accidental releases. The ability of the tugs to rendezvous
and recover waste from aborted missions will also contribute to the
safety of space operations.
A wide variety of encapsulation and system approaches is possi-
ble. For example, it is possible to consider such options as launch-
ing only small amounts of waste at one time as "piggy back" payloads
for shuttles that are not fully loaded and collecting them at a space
depot. While the practically of such concepts remains to be deter-
mined, there are many options to optimize safety and minimize risk.
6.3.5 Radiological Considerations
Disposal in space of a fraction of the nuclear waste may affect
the ecosystem during normal operations and, in the event of accidents,
may result in the release of radioactive materials to the environment.
The primary concern in extraterrestrial waste disposal is accidents
during the launch phases.
The steps to compute radioactive waste release consequences are
as follows:
1. Determine the probability of the accident
2. Determine the source terms and their probability
6-37
-------
3. Predict the movement or dispersion of released material
through the environment by atmospheric or aquatic disper-
sion processes
4. Determine the probable number of people exposed
and the probable doses received
A generalized diagram for risk analysis in a space vehicle launch
sequence is shown in Figure 6-10. Such an analysis has not been con-
ducted for the space disposal of radioactive waste. However, recent
studies have been conducted by Battelle Columbus for the National Aero-
nautics and Space Administration (NASA) based on worst case analy-
sis.3
The Battelle study assumed a 5500 kg nuclear waste payload in a
calcine powder waste form. The study considered five options of waste
mix and five types of abnormal events or accidents.
The waste mixes considered were as follows:
• The fuel is leached from its clad and the entire dis-
solved solution is solidified and shipped into space;
• 99.5 percent of the uranium is recovered from the dis-
solver solution. The remaining dissolver solution is
solidified and sent into space;
• 99.5 percent of the uranium and plutonium is recovered
from the dissolver solution. The remaining dissolver
solution is solidified and sent into space;
• 0.1 percent of the uranium and plutonium, the balance
of the actinides and all the rare earths except cerium
are solidified and sent into space;
• 99.5 percent of the uranium and plutonium and a minimum
of 94 percent of the technetium are recovered from the
dissolver solution. Only the technetium is sent into
space.
The abnormal events considered were as follows:
6-38
-------
SOURCE: NUS Corporation, Reference 9.
FIGURE 6-10
GENERALIZED FLOW DIAGRAM FOR RISK ANALYSES
-------
• On- or near-pad catastrophic launch vehicle explosion and fire
(major impact to lower atmosphere);
• Launch vehicle explosion and fire at high altitude or reentry
and burnup of the nuclear waste payload from orbit (major im-
pact to upper atmosphere—followed by chronic impact to lower
atmosphere);
• Water impact resulting from launch vehicle failure or intact
reentry of the nuclear waste payload from orbit (major impact
to ocean or fresh water);
• Land impact resulting from launch vehicle failure or intact
reentry of the nuclear waste payload (major impact to land).
The entire inventory of 5500 kg was assumed to be released in the
worst case analysis.
The conclusion of the Battelle study is as summarized below:
"For the five types of events considered here, a catastrophic
on- or near-pad launch vehicle" failure at KSC, resulting in the
rupture and release of the radioactive waste payload, is consid-
ered very serious. A high altitude burnup is considered serious,
with other events following in severity (sea, land, and space
lunar type accidents). The assessment of effects to man and
ecosystems as a result of these events is extremely difficult.
However, order of magnitude projections can be made.
"In the case of the catastrophic on- or near-pad launch vehicle
failure, assuming Mix No. 3, the local human population exposed
to the'cloud resulting from the fireball containing the radio-
active material could receive an inhalation radiation dose ex-
ceeding 500 times background (background 0.1 rem/year). At
downwind distances of 100 km, exposures could exceed 100 times
background. Public radiation standards vary from 5 to 15 times
background, depending upon the organ or part of the body ex-
posed. If gravitational settling of radioactive particles were
a predominant effect, the area downwind of the event would be-
come severely contaminated, many life forms would be destroyed,
and the land area would have to be isolated indefinitely.
"An upper atmospheric burnup of the payload could result in
similar effects, depending upon the particle size distribution
of the radioactive material, and the longitude and latitude of
the event. Chronic toxicity effects would be expected for the
case of worldwide distribution of the material. The amount of
strontium-90 which could be released by one accident involving
6-40
-------
Mix No. 3 amounts to 40 percent of that released from all nu-
clear devices through 1962.
"For accidents at sea resulting in the release of radioactive
material, exposure to man would primarily be by bioaccumula-
tion of nuclides in aquatic food chains. Reduced productivity
of aquatic organisms could limit food supplies to man.
"Effects caused by a nuclear waste package crashing on land,
in a populated area, followed by dispersal of the waste
(calcine powder) could be significant. If the waste mate-
rial is characterized by a fine particle distribution, then
the chance for resuspension in the air becomes likely, thus
causing severe impacts to local human, plant and animal pop-
ulations.
"Accidents in space, followed by radioactive releases are not
expected to impact the earth's biosphere; however, contamina-
tion of orbital regions or other celestial bodies (especially
the moon) could preclude the use of an orbit or as a future
resource. Strong opposition would be expected from the scien-
tific community, if it were likely to contaminate the moon or
other planets by a waste disposal accident."
The worst case analysis is not, of course, representative of the
expected consequences of an accident and can be considered improbable.
The analysis does indicate, however, that a detailed risk and conse-
quence study is required to assess the acceptability of extraterres-
trial disposal. It is not possible at this time to specify allowable
or critical dosage levels for the undetermined waste disposal frac-
tions to be launched. Acute toxicity at dose levels in the range of
50 rems per year will result in some deaths. Chronic toxicity for
dose levels of a few mrem per year may introduce genetic and fertility
effects.
All of the dose magnitudes (except the short term airborne respi-
ratory particles and the vaporized fraction) may be greatly affected
by recovery and corrective measures. In the case of launch area
6-41
-------
accidents, recovery and corrective action can probably be completed in
a few days. Outside the launch area the recovery time is probably a
function of population density ranging from a few days to longer pe-
. *
riods, depending upon the location of the waste capsule.' Recovery
will be important in reducing the hazard and the long-term risk to
future populations.
In contrast to the individual launches of the past decade, space
waste disposal will require many routine, repetitive space operations.
By the year 2000, or shortly after, as many as several thousand radio-
active payloads may be launched. It might be reasonably projected
that the probability of a prompt release accident for a single flight
could be in the range 10"^ to 10"^. The corresponding long-term
probability for several thousand launches could be in the range of
10~2 to 10~^. It would therefore be anticipated that a small
number of accidents would occur with some release.
A small number of accidents and the minor releases that would
occur in a program in which vehicles, capsules, and operational pro-
cedures are optimized in the manner previously discussed would be
highly unlikely to serious affect the ecology as a whole. Local im-
pacts could be significant, however. System designs would therefore
be required which would limit ecological impacts to inconvenience
rather than injury. These system designsxwould, of course, be es-
tablished in the context of the weighing of the overall long-term
benefits to the public of permanent waste disposal against the degree
of risk and cost.
6-42
-------
6.4 Economic Impacts
The cost of space disposal estimated in reference 2 (and simi-
larly in reference 1) in 1973-74 dollars is as follows:
Cost, $/Kg waste
Partitioning 14,000
Encapsulation 4,700
Space launch 150,000
Total 168,700
The cost in mills per kw hr generated is 0.5 mills/kWh.
The cost in percent of cost of generated electricity
(early 1970's) is 5 percent.
These costs do not include the cost of disposal of the separated
wastes remaining on the earth. A more modern estimate is required
to account for escalation of the past few years. It will then be
instructive to identify and add the cost of disposal of the wastes
left behind and finally to determine the incremental cost for space
disposal so that its benefits may be weighed against cost.
6.4.1 Partitioning
For the purposes of this draft and pending an accepted number, an
escalation update is assumed as follows:
$14,000 $/MT escalated 6 percent per year for 3 years
14,000 x 1.19 - 16,674 $/MT
6.4.2 Encapsulation
Encapsulation costs given in reference 2 are based on existing
designs scaled to the capacity required to meet waste disposal needs
and are broken down approximately as follows:
6-43
-------
Labor, materials and labor related $2,000/kg
Construction (capital) $2,700/kg
Escalation of labor and materials by six
percent per year and capital costs by 12
percent per year for three years $6,200/kg
Because this cost is small relative to the other costs involved, high
precision is not needed.
6.4.3 Space Launch Costs
Shuttle and tug costs used in the references are broken down as
follows:
Each shuttle flight $10.5 M
Each reusable tug flight $ 1.75 M
Each expendable tug $ 5.5 M
One complete payload launch $28.25 M plus $.5M per flight
for new launch facilities or total = $28.75M.
Actual operational costs are not known at this time. The most
recent published estimates available have been given in a September
1976 statement by NASA Associate Administrator, J. F. Yardley, to
the U.S. House of Representatives Committee on Science and Technology
as follows:
Cost of each shuttle flight in 1975 $, Million
Commercial and Foreign 19.0 - 20.9
Other U.S. Government 16.1 - 18.0
DOD 12.7 - 14.1
Assuming that waste disposal can be considered to be a government
activity, a cost of $17M per flight is assumed.
6-44
-------
The new cost per shuttle flight escalated to 1977 dollars is:
17 x (1.06)2 -• $19. 1M
19 1
The growth ratio is *'. =• 1.82. .
l u. j
Applying this same ratio to tug costs yields the following vehicle
costs for one pay load:
$M (1977)
Two shuttles 38.2
One reusable tug 3.2
One expendable tug 10.0
Escalated facility costs .7
Total $ 52. 1M
or 52.1/113 kg $450,000/kg.
The new total cost breakdown is as follows:
Separation 17,000 $/kg
Encapsulation 6,000 v/kg
Vehicle 450,000 $/kg
Total $483,000 $/kg
The new cost ratio is ,'-, nn = 2.8.
loo , /UU
The cost per kWh estimated in references 1 and 2 was 0.5 mills/
kWh. Assuming constant electric energy generation per kilogram of
waste, then the cost per kWh is 0.5 x 2.8 = 1.4 mills/kWh.
The above is provided only as an estimate of cost. The cost of
disposal of separated wastes remaining on earth must also be added.
Although the cost is higher than other disposal methods, it is not
necessarily the limiting factor. The consequences of potential acci-
dents is the more important consideration.
6-45
-------
REFERENCES
1. "Feasibility of Space Disposal or Radioactive Waste" I, Executive
Summary, NASA TM X-2911, Lewis Research Center, Cleveland, Ohio
44135, December 1973.
2. "High Level Radioactive Waste Management Alternatives" BNWL-1900
Volume 4, Battelle Pacific Northwest Laboratories, Richland,
Washington 99352,, May 1974.
3. "Preliminary Evaluation of the Space Disposal of Nuclear Waste"
Report to NASA, Battelle Columbus Laboratories, August 30, 1977.
4. "Technical Support for the Radiation Standards for High-Level
Radioactive Waste Management," Task A, Draft, Arthur D. Little,
Inc.
5. "Final Safety Analysis Report for the MTS Mission," General
Electric, Doc. No. 775054206, January 1977.
6. "SNS Source Term Evaluation Program," BNWL-975, Battelle Pacific
Northwest Laboratories, Richland, WA, January 1969.
7. "Final Environmental Statement, Barwell Nuclear Fuel Plant, USAEC
Docket No. 50-332.
8. Draft Environmental Impact Statement for Space Shuttle Program,
NASA, Washington, D.C., August 1977.
9. "Overall Safety Manual" USAEC Space Nuclear Systems Division,
NUS Corporation, Rockville, Maryland 20850, June 1974.
10. "Contingency Operational Plan for S Map 27" Manned Space Craft
Center, Houston, Texas, October 1969.
6-46
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7.0 SEABED DISPOSAL
In this section, the emplacement of radioactive wastes in deep-
sea sediments is discussed relative to the technical feasibility and
environmental acceptability of seabed disposal.* The technical
feasibility of the concept depends upon demonstrating that seabed
disposal can contain radioactive waste long enough for it to decay
to innocuous levels. The time required for some long-lived actinides
and fission products to decay to innocuous levels is several million
years, a time period for which long-range predictions are somewhat
tenuous at best. The environmental acceptability must therefore be
assessed as to the degree of long-term isolation and the potential
radiological impacts of seabed disposal on the marine environment and
to man. A discussion of the environmental impact assessment of seabed
disposal will be made by dividing the high-level radioactive'and
transuranic contaminated wastes into distinctive components
(actinides, select fission products, volatiles, etc.). The
effectiveness of seabed disposal for each component can be compared
and will help identify potential'environmental problems.
Seabed disposal has been explored by several countries as a
means of permanent disposal of high-level radioactive and transuranic
wastes. Currently, there are no accepted international criteria or
standards to guide individual national efforts. The International
*Seabed disposal is the emplacement of waste within the seabed
sediment or geologic formations in such a way as to ensure long-
term containment. It is not to be confused with ocean-dumping.
7-1
-------
Atomic Energy Agency (IAEA) has recently expanded its waste management
programs to evaluate several proposed high-level waste disposal
options including seabed disposal. However, waste management programs
in the nations producing nuclear power are still in very early stages
of development, and serious efforts by the IAEA to solve the waste
problem on an international level are just beginning. A series of
three advisory group meetings have been held by the IAEA with the
2
task of developing definitions and guidelines for seabed disposal.
The public concern today over the radiological consequences of
seabed disposal, in part, is based on past marine disposal practices
of the U.S. and other industrial nations. Between 1946 and 1970, for
example, the U.S. Atomic Energy Commission (AEC) licensed the dis-
posal of more than 86,000 containers of low-level wastes (totaling
94,000 curies) into the Atlantic and Pacific Oceans. Britain dis-
posed about 45,000 curies of low-level radioactive wastes into the
Atlantic from 1951-1966.1
From a scientific point of view, it is very difficult to
determine if damage has occurred or if a real hazard exists as a
result of international radioactive waste disposal practices. In
this regard, the U.S. has taken a leading role to protect the
marine environment from pollution including disposal and dumping of
radioactive wastes into the oceans.
Under the Marine Protection, Research and Sanctuaries Act of
1972, EPA was given authority to issue permits for disposal of
7-2
-------
low- and medium-level radioactive wastes into the ocean, but EPA
has no similar control over high-level wastes. Congress would have
to amend the Act, if the government decided to implement any form of
sub-seabed disposal of high-level wastes.
The Nuclear Regulatory Commission (NRC) presently has jurisdic-
tion over the licensing of radioactive waste repositories while the
EPA has authority over the establishment of standards and regulations
for the placement of radioactive waste into the ocean. The Department
of Energy (DOE) is responsible under the National Environmental Policy
Act (NEPA) for the environmental assessment of planned high-level
waste disposal techniques, including seabed disposal. While specific
criteria and standards for new regulations for waste management
are still to be developed, recently established NRC goals include the
following:^
• Isolation of radioactive waste from man and his environment
for specific periods to assure public health and safety and
preservation of environmental values;
• Reduction to as low a level as is reasonably achievable of
(a) the risk to public health both from chronic expo-
sure associated with waste management operations and
possible accidental releases of radioactive materi-
als from waste storage, processing, handling, or
disposal;
(b) long-term commitments such as land-use withdrawal,
resource commitment, and surveillance requirements.
Thus, the ultimate evaluation of the potential DOE seabed disposal
concept by the NRC and EPA will have to be made with an established
'• • ! ' .
set of technical, social, and environmental criteria and standards.
This study will discuss the present state of knowledge on
seabed disposal and will assess the radiological impact to man and
7-3
-------
possible damage to the marine ecosystem from emplaced wastes. Sea-
bed disposal, as defined in this study, is the controlled emplacement
of radioactive waste in deep-sea sediments or rock formations under
the ocean. The evaluation will carefully identify the transport
processes by which radionuclides could migrate from the emplacement
site through the metal canisters and the deep-sea sediment and the
ocean column to the biosphere.
Physical and environmental barriers that may prevent migration
of radionuclides exist. On the other handj several mechanisms may
act singularly or in combination to compromise the integrity of these
barriers. Included among these mechanisms are the following:
• corrosion of the-canister;
• leaching of the waste material;
• upward transport through the upper sediment layers to the
lowest water layers;
• advection and diffusion through the water column;
• thermal effects on sediment or the water column;
• biological transport of incorporated isotopes across the
seabed or upward through the water column.
In principle, the rates of all these processes are measurable
or can be estimated. Regardless of the method chosen for emplacement
of wastes in the seabed, calculation of breakthrough times (migration
times) for each of these barriers must demonstrate that the waste
will be contained for long periods of time.
The chapter will be organized in the following manner:
7-4
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• Section 7.1, Ocean Characteristics
This section describes the ocean environment and selects
ocean regions which will be most suitable for waste reposi-
tories. A comparison of the relative merits of.alternative
ocean sites is made based on generic site selection criteria.
• Section 7.2, Emplacement Techniques
This section/discusses the possible methods of placing
canisters at a proper depth in a sediment or rock layer.
• Section 7.3, Environmental and Health Considerations
This section discusses the environmental and health aspects
of seabed disposal.
• Section 7.3.1, Engineering and Environmental Barriers Against
Waste Intrusion into Biosphere
This section discusses migration mechanisms by which man
may become exposed to radiation after its release from the
deep sea emplacement site.
• Section 7.3.2, Research Needs
This section identifies data required to understand the
entire ocean-sediment waste system in order to adequately
assess the. feasibility of the seabed'disposal concept.
• Section 7.3.3, Radiological Impact Assessment
This section discusses the potential radiological impact to
man and possible damage to the marine ecosystem from emplaced
waste.
• Section 7.4, Economics
This section provides data on costs for seabed disposal.
7.1 Ocean Characteristics
Several ocean provinces may contain possible locations for
controlled emplacement of high-level radioactive waste under the
sediments of the ocean floor. High-level wastes are the most diffi-
cult wastes to dispose of because of the combination of intense
7-5
-------
radiation and heat from the 'relatively short-lived isotopes and the
great length of time required for the transuranic nuclides to decay.
If one considers the ocean provinces on the basis of their overall
suitability as disposal sites, it is possible to compare the relative
characteristics of each province and apply the results of the compari-
son to select a potential disposal site. The criteria which have
3 4
been used to evaluate disposal sites are as follows: '
• Temporal and Geological Stability: This may be estimated by
observing the record of the past geological events held in
the sediments;
• Inaccessibility; The areas selected should be as far removed
as possible from the normal and expected activities of
mankind;
• Lack of Resources; Waste disposal should not seriously
interfere with the exploitation of resources;
• Permanence; Recovery of the waste material at a later
date need not be a requirement;
• International Acceptability; If agreeable to all affected
nations, seabed disposal may provide an international solu-
tion to nuclear w^ste disposal. As such, areas should be
selected outside of direct national jurisdiction.
The ocean floors are divided into three principal physiographic
provinces, each occupying about a third of the world's ocean area:
• Continental Margin, which includes continental shelf, inland
seas, marginal plateaus, continental slope, and continental
rises;
• Midoceanic Ridge, a global plate boundary which includes
fracture zones, ridge flank and crest, and rift valley and
mountains;
• Ocean Basin Floors, which include abyssal plains, abyssal
hills, oceanic rises, and deep sea trenches (global plate
boundary).
7-6
-------
Characteristics of these three ocean provinces are summarized In
Table 7-1.
A number of geological media have been considered for disposal
beneath the seabed. Clay, shale, crystalline rocks of several kinds,
and similar deep sea sediments are under consideration as prime
disposal candidates or alternatives.
7.1.1 Continental Margin
The continental margins, located on the perimeter of the
continents, represent the most dynamic environment of the ocean.
Seasonal temperature changes in the water are high, chemical and
biological processes are most variable, and the geology is most
complex and unpredictable. Continental margins contain pools of
hydrocarbons accessible with todays technology as well as most of
the world's great fishing grounds. Surface sediments of these
provinces change radically over short distances, ranging from hard
rock to gravel to clay within only a few miles.
Continental margins may be characterized by:
• high resource value including food, mineral, hydrocarbons
• shallow water depth
• low geologic stability
• very strong and variable currents
• high sedimentation and erosion
• variable conditions (temporally and geographically)
• biological activity
7-7
-------
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These dynamic elements are enough to rule out the margins because
they do not meet the criteria of stability and isolation.
7.1.2 Mid-Oceanic Ridge
The Mid-Oceanic Ridge (MOR) forms the "construction" plate
4
boundary of the ocean floor. The center of/ the MOR is a hot,
seismically active rift valley which continually extrudes new crust.
Sediment thickness is typically too small to be detected. The center
of the MOR may be characterized as follows:
• seismically and volcanically unstable, almost constant
earthquakes
• without sedimentation
• topographically rough
• shallow in water depth
• having hot, molten basalt near surface
It is unlikely that the MOR center would be chosen as a suitable
location for the disposal of large quantities of potentially hazar-
dous waste.
The MOR is a global plate boundary and includes fracture
zones, flanks, crests and rift valleys, and mountains. The flank
areas are characterized by:
• high stability
• low resources potential
• inaccessibility
7-9
-------
Based on these factors and the comparison provided in Table 7-1, the
flanks of the MOR meet the criteria for acceptable waste disposal
sites.
7.1.3 Ocean Basin Floor
The ocean basin floor is the deepest of the three provinces
and includes the flat abyssal plains, abyssal hills, and deep-sea
trenches. The flat abyssal plains have been created through deposi-
tion of sediments and debris from continental margins by strong
currents. Sediments recovered from abyssal plains are typically
silty clays mixed with coarsely graded layers of sand and gravel.
The abyssal hills were originally formed as extrusions of
basalt from the MOR center. These regions are generally covered with
50 to 100 m of brown zeolitic clay overlying a few tens of meters of
limestone. The concept of disposal under the ocean floor in the
abyssal hills is attractive for several reasons:
• high geologic stability (seismically passive),
• invariant conditions (temporally and geographically),
• slow currents,
• low bio-productivity (low on surface, very low on bottom),
• limited resource potential
Deep-sea trenches and subduction zones are areas where, accord-
ing to crustal global plate tectonics, theory, one edge of a crustal
7-10
-------
lithospheric plate is moving under the other plate and down the
earth's mantle into the asthenosphere (plastic zone) of the mantle.
Sea trenches are among the less stable areas on the earth and undergo
extensive changes in relatively short times. Deep sea trenches may be
characterized as follows:
• seismically active
/
• volcanically active
• containing unstable sediments including slumping, sliding, and
strong currents
These conditions do not meet the criterion of stability.
7.1.4 Criteria for Site Selection of Oceanic Provinces
Site selection criteria have been applied to the three ocean
provinces to determine feasible locations for waste disposal. Follow-
ing are the most important considerations:
• frequency of catastrophic events
• rates of natural processes
• -predictability
Most of the data necessary to compare the three ocean provinces
come from interpretations of past events by examining the properties
of deep-sea sediment. The Deep Sea Drilling Project of the National
Science Foundation and The Seabed Disposal Program of Sandia Labora-
tories and Woods Hole Oceanographic Institution have been instrumental
in obtaining sediment data from numerous drilling experiments,
seismic profiles (seismographs), and bottom sediment photographs.
Interpretation of these data yields significant insight into'the
7-11
-------
geologic stability and predictability over periods of million of
years.
Comparing the major ocean provinces using the above criteria
(see Table 7-rl), it can be concluded that two ocean provinces are best
suited for waste disposal. They are (in order of suitability) the
3 4
abyssal hills and the flanks of the Mid-Oceanic Ridges. ' Those
areas which occur in the middle of the great oceanic gyres are
especially attractive because of their low biological productivity.
Thus, the areas in the middle of the tectonic plates and the middle
of the gyres (mid-plate/mid-gyre) are best suited for waste disposal
and have been the targets for further analysis. The mid-plate/mid-
gyre region of the Pacific Ocean has been investigated as a potential
site to perform further experiments and analysis of sediment samples.
Core sample data have indicated '•hat this region has a continuous
record of millions of years of tranquility and geological stability.
The Pacific Ocean mid-plate/ mid-gyre region is also characterized by
unconsolidated clay sediments which make good sites for waste emplace-
ment.
The Department of Energy (DOE) has supported the Seabed Emplace-
ment Program to determine if any submarine geologic formation can
contain radioactive wastes long enough for it to decay to innocuous
level. More specific geological, geophysical, and oceanographic data
are currently being obtained from site-specific studies at mid-plate/
mid-gyre areas, such as MPG-1 in the middle of the central North
7-12
-------
Pacific, about 600 miles north of Hawaii. The Pacific mid-plate/
mid-gyre region is a more suitable location for a disposal site than
the mid-plate/mid-gyre region of the Atlantic for several reasons:
• greater water depth in the Pacific Ocean;
• the Pacific has steady, deep, stable, and cold ocean currents
capable of maintaining non-mixing conditions for perhaps a
thousand years;
• the Pacific generally is believed to have geologically older
sediments with a mineral composition (montmorillite and zeo-
lite) containing higher distribution coefficients (Kd) than
the Atlantic sediment (kaolinite and illite);
• the Pacific has greater distance from global plate boundaries
and remoteness from man.
To extend the data base and further assess the mid-plate/mid-gyre
environment, a second area (MPG-2) has been selected for sediment
sampling and measuring. MPG-2 is located 700 miles northeast of
MPG-1.5
Experiments are currently underway to establish the adequacy
of the sediment to waste migration, especially with respect to the
retention of radionuclides. Sediments that have adequate containment
properties, such as brown oxidized clays, still have to be studied at
sea to determine whether they can be found in sufficient thickness in
MPG-type settings. Finally, it is necessary to determine in situ the
physical and dynamic response of the sediments to emplacement and to
4 5
establish sediment hole closure properties. '
7-13
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678
7.2 Emplacement Techniques ' '
Many possible methods of placing canisters at a specified depth
in a deep sea sediment have been investigated (see Figure 7-1). These
methods include controlled drilling from a surface ship and free-fall
penetration (with a high velocity as driving force). Radioactive
wastes could be emplaced either in the unconsolidated sediments such
as oxydized red clays or in the underlying bedrock. The free-fall
penetration technique would require a sediment with plastic properties
which will collapse to fill the hole made by canister entry in a
reasonable time.
An exact procedure for emplacement will not be chosen until
it has been demonstrated that seabed disposal is feasible. However,
it is necessary to consider one technique in order to assess the
effects of emplacement on sediments. The technique chosen for
analysis in this assessment is the free-fall penetration technique.
The full spectrum of possible techniques should be studied, however,
before a total emplacement system can be designated. A description
of three emplacement methods is given below.
7.2.1 Free-Fall Penetration
In this emplacement method, the waste container would be dropped
from a ship through the water column. A terminal velocity of 70
miles/hr would be reached before the canister would penetrate the
clay sediments. Since the clay sediment is soft, it is expected
7-14
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WINCH-EQUIPPED SHP
(COST FACTOR • I
ORLUNG SHP
(COST nCTOft • 3)
•' CRUSTAL ROCK' (BASALT)
IKU (I WMW».S«»
£_gp*-«WUT S8AL5-
SOURCE: Alternatives for Managing Wastes from Reactors and
Post-Fission Operations in LWR Fuel Cycle, Volume 4,
Waste Disposal, ERDA-76-43, May 1976.
FIGURE 7-1
ENGINEERING CONCEPTS FOR EMPLACEMENT OF
RADIOACTIVE WASTE CANISTERS IN THE SEABED
7-15
-------
that penetration could exceed 30 meters. Monitoring instruments
would be placed on the seabed floor to detect leaks. Canisters
could be retrieved from the sediments, but this is not a goal Of
sediment emplacement. Laboratory studies indicate that closure of
the emplacement cavity would occur immediately following canister
penetration.
7.2.2 Winch-controlled Emplacement
In this option, the waste canister is attached to a drilling
device designed to penetrate into the clay sediment. This device
would either use momentum or some driving mechanism, such as vibration,
to achieve penetration. One advantage of this method is that the
canister could be immediately recovered in the event of a malfunc-
tion. However, laboratory studies indicate that there may be some
hole closure problem associated with this method. If necessary, it
may be possible to provide a sealant that could be left to fill the
cavity above the canister wheii the drilling device is pulled out.
7.2.3 Drilled Holes
The technique for deep-sea drilling from a surface ship has
been demonstrated by several marine research centers. This emplace-
ment technique has the advantage that many canisters could be placed
in a single bedded area at greater depths (100-500-meters) than other
emplacement methods. As such, it will be necessary to develop a seal-
ant which would fill the drilled cavity above and between the canisters,
7-16
-------
To date, drilling techniques using sealant for seabed disposal have
not been demonstrated.
7.3 Environmental and Health Considerations
7.3.1 Engineering and Environmental Barriers Against Waste
Intrusion into the Biosphere
This section discusses the mechanisms by which radionuclides
are transported from the emplacement through engineering and environ-
mental barriers which retard migration to parts of the ocean of
immediate significance to mankind. Because specific disposal sites
have not been designated and because data on the rates of transport
for all radionuclides are insufficient in some cases, the analysis of
engineering and environmental barriers against waste intrusion into
the biosphere contains many uncertainties. Until site specific
data on transport mechanisms in deep-sea sediments and thermal
effects on and by the canister are obtained and better understood,
analysis of mechanisms by which radionuclides are transported will
have to rely on generalized information on the ocean environment.
There are several mechanisms by which radionuclides are trans-
ported:3,6,9,10
• corrosion of the canister
• leaching of the waste material
• upward transport through the upper sediment layers to the
lowest water layers
• advection and diffusion through the water column
• thermally driven transport through the sediment or the
water column
7-17
-------
• biological transport of incorporated isotopes across the
seabed or upward through the water column
In principle, the rates of all of these processes are tneasureable
or capable of being estimated. Regardless of the method chosen
for emplacement of wastes in the seabed, calculation of breakthrough
times (migration times) for each of these barriers must demonstrate
that the wastes will be contained for periods approaching geologic
time scales. A diagram of the transport processes of radionuclides
in the ocean which will be considered in this assessment is illustrated
in Figure 7-2. This methodology forms the basis for discussion, the
radiological impact to man, and ecological damage to the marine
environment from seabed disposal.
349
7.3.1.1 Waste Form. * * There are several considerations in
providing engineering barriers against dispersion of radionuclides to
the ocean environment. The first consideration is the specific waste
form which is designed to prevent leaching of the waste material.
The exact forms in which high-level radioactive and transuranic
wastes will be packaged for seabed disposal are sensitive to the
choice of fuel cycle, the physical characteristics, and the radio-
logical properties of the waste material.
If the reprocessing option is implemented, the liquid waste
produced during reprocessing of reactor fuel rods is basically a
solution of radioactive and nonradioactive elements in nitric acid.
The solution is very corrosive, generates large amounts of heat,
and is highly radioactive. For waste disposal, these wastes have to
7-18
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STRONG CONTACT
WITH MAN
WEAK CONTACT
WITH MAN
LAND
ASSOCIATED
WATERS
FLOW OF
RADIONUCLIDES
Reference: Alternatives for Managing Wastes From Reactors and Post-
Fission Operations in the LWR Fuel Cycle, Volume 4:
Alternatives for Waste Isolation and Disposal, ERDA-76-43.
FIGURE 7-2
TRANSPORT PROCESSES OF RADIONUCLIDES FROM SEABED DISPOSAL
7-19
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be in suitable chemical forms which are stable even at the high
temperatures caused by the heat from radioactive decay. The solu-
bility of the chemical compounds in water must be as low as possible,
so that even after final disposal, if there is any contact with
water, the leach rate would be low. Present plans call for the
solidification of the liquid waste by evaporation of the acid fol-
lowed by incorporation in some stable material of high integrity such
as concrete, glass, or zeolites. The percentage of radioactive waste
that can be incorporated in the stable material depends on the
chemical composition and nature of both materials. Not all of the
fission products, particularly volatile radionuclides, can be incor-
pprated into available types of material. For example, there is no
technique currently available to fuse iodine compounds into glass.
The problem of disposal of krypton-85 is difficult because krypton-85
(a noble gas) does not form a stable chemical compound. The only
possible methods of disposal are storage at high pressure in cylinders
and adsorption in some suitable porous material. In a companion
study*, specific waste forms associated with the volatile radio-
nuclides iodine-129, tritum, krypton-85, and carbon-14 are discussed
in great detail.
The waste form itself forms the first barrier to migration.
Several questions about the properties of these waste forms and their
^Assessment of Waste Management of Volatile Radionuclides,
The MITRE Corporation.
7-20
-------
effectiveness in preventing dispersion are still unanswered. Exact
leach rates for many of the fission products are not known because a
final decision on the best types of waste forms has not been made.
Few, if any, leach experiments have been carried out using solutions
resembling sediment pore waters or at temperatures and pressure anti-
cipated in the seabed after emplacement. If glass is used as the
waste form, another question of concern is the long-term stability of
the glass. The heat produced by the fission products during decay may
convert the waste from a glass to a mass to tiny crystals. Devitri-
fication may have the effect of speeding up the rate at which elements
are released from the glass. Thus, the effectiveness of a glassy
waste form may be very different if devitrification occurs in a few
years rather than a few centuries.
Because of the possibility of devitrification, a glass waste
form may not confine radioactive elements for more than a thousand
years. This duration is far less.than the time period that is re-
quired for the longer-lived actinides to decay to innocuous levels.
This period of time, however, may be long enough to allow the waste
to dissipate most of its heat before the waste begins to interact
with the surrounding sediments. Therefore, it is important to
determine how effective the waste forms are in preventing isotope
migration for the first several thousand years.
349
7.3.1.2 Canister. ' ' High-leveJ. radioactive and transuranic
contaminated waste, whether in solidified form from reprocessing or
7-21
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spent fuel form, will most likely be sealed in metal canisters.
Glass or ceramic canisters are possible options, but may be less
suitable for seabed disposal because of strength requirements for
handling and shipment. Figure 7-3 shows a proposed standard canister
for high-level, low-level, and intermediate-level wastes. The waste
canister is the second barrier against dispersion of radionuclides
to the ocean environment. The canisters will be designed to meet
the following requirements:
• ability to dissipate the heat from newly packaged waste;
• long-term integrity of canister;
• ability to resist corrosion and leaching at high pressure
and temperatures.
If seabed disposal is implemented, canisters will need to be
designed to resist corrosion for a long time. Seawater (which is
much like sediment pore water) is an extremely corrosive fluid.
The only candidates for canister materials that appear suitable at
present are titanium and zirconium alloys. Research to better
understand the behavior of these materials in seawater and ocean
sediments is being carried out at Sandia Laboratories. Several in
situ corrosion experiments have also been conducted.
The corrosion of metals in marine environments limits their
useful life and precludes the use of some materials which are
attractive because of their low cost. The rates of corrosive attack'
have been documented for a large number of systems, even though the
basic corrosion processes which occur are not well understood.
7-22
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•12" STANDARD PIPE
12J5" O.D. (32.4cm)
12" I.D. (30.5 cm)
CARBON STEEL OR STAINLESS STEEL
HEMISPHERE HEAD
SOURCE: High-Level Nuclear Wastes in the Seabed?
Oceanus, Volume 20, Number 1, Winter 1977
FIGURE 7-3
THE PROPOSED STANDARD CANISTER
7-23
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Experiments have shown that for many canister systems, the rates of
localized corrosion (e.g., pitting, crevice corrosion) were high
and may present a serious problem in the search for candidate materials
which have extended lives (1000 years).
The best estimate at present is that materials such as zirconium
and titanium alloys are capable of confining radioactive materials for
a few thousand years. Again, this breakthrpugh time estimate is
insufficient for the total containment that is required. Nevertheless
as was the case for the waste forms, a thousand-year time period is
long enough to allow the waste to dissipate most of its heat before
it begins to interact with the surrounding sediment. This may have
the net effect of reducing the possibility of rapid upward transport
in convection currents that could be produced by heat dissipated
from the canister.
To illustrate this point, a newly filled canister containing a
mix of radionuclides, some with short-half lives and some with
long-half lives, will give off 10 to 30 kilowatts of heat to the
9
surrounding sediment. A typical canister will radiate enough
heat to raise the temperature to 600 C in the immediately
surrounding sediment. This temperature will decrease to an
undisturbed sediment temperature of about 0 C at a distance of 30
meters from the canister. At 600°C, strong thermal gradients are
7-24
-------
created which may cause rapid upward transport of leached radio-
nuclides if these nuclides would breach containment immediately
following emplacement. After about 1000 years, most of the short-
lived radionuclides will have decayed and thus the total heat emitted
from the canister will be reduced significantly. Based on the
physical properties of deep-sea sediments, the temperature of the
sediment immediately at the canister may be reduced from 600 C to
about 200 C after 1000 years. This temperature reduction may have
the net effect of reducing the possibility of rapid upward transport
because of the reductions in the thermal gradients created at this
temperature. This fact points out the importance of developing
suitable waste forms and canisters which will be effective as barriers
for approximately 1000 years.
The high temperatures in the vicinity'of the waste canister could
have the effect of fluidizing the entire sediment/pore water. The
canister could then sink through this viscous fluid to greater depths.
Present knowledge is not adequate to predict whether this process would
actually occur.
7.3.1.3 Sediment.4'5'6'9'10'11'12
Physical Properties
Information on the physical properties of deep sea sediments pro-
vides a very crucial part of the data necessary to evaluate transport
mechanisms for radionuclides. A combination of spot sampling by
drilling or coring and sub-bottom acoustic profiling techniques is
used to obtain information on the physical properties of sediments.
7-25
-------
Under the Seabed Disposal Program, several unconsolidated sedi-
ments have been sampled and examined to determine their physical
properties and their appropriateness as a barrier to radionuclide
migration. These sediments are as follows:
• oxidized red clay sediments
• calcium carbonate sediments
• silica sediments
• continental margin sediments
Oxidized red clay sediments have a number of physical properties
that make them attractive as emplacement sites and have been the sub-
ject of studies on transport mechanisms in sediments. Oxidized red
clays are extremely fine grain sediments with most particles less than
1 micron in diameter. As a result, they have low permeabilities
—8 —7 9
(10 to 10 centimeters/second). Oxidized red clays also have
very large surface areas'per unit volume of sediment. This is an im-
portant attribute in reactions between dissolved waste elements and
clays, and in their ability to extract (sorb) metals from solutions.
To examine the barrier properties of oxidized red clays, data
are being gathered in the following areas:
• distribution coefficients of sediments
• effects of heat on sediments (heat transfer properties)
s
• dynamic response of sediment to canister emplacement
• hole closure properties of clays
• biological and ecological implications of thermal waste
heat on sediments
7-26
-------
Distribution Coefficient and Retardation Factors ' ' '
Some fission products may react little or not at all with
deep-sea sediments. These are expected to include tritium, krypton,
technetium, iodine, and radon. The time it takes for these isotopes
to migrate from the canister through the clay sediment to the
sediment surface can be represented by the following:
T = d2/C_,
d
where T = time [ sec ]
d = sediment depth [ cm]
C = diffusion coefficient of element in sediment
[cm /sec]
As an illustration, it would take iodine and tritium buried 100
meters below the deep-sea sediments approximately a million years to
migrate to the ocean sediment interface. This is based on a diffusion
coefficient of 3 x 10 square centimeters/second (which is an
average value for deep sea sediments). For tritum, deep sea clays
will certainly act as an effective barrier to migration because of
its short half-life. Since iodine-129 has a half-life of 1.7 x 10
years, however, the clay sediments may not act as an adequate barrier
to iodine migration. It will, however, reduce the cumulative time
that the iodine exists in man's environment. Therefore, the sediment
properties are a big factor for those radionuclides such as iodine-129
which have half-lives such that a significant quantity of the isotope
still remains after canister and specific waste forms are no longer
intact.
7-27
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Most fission products, however, will enter into complex physio-
chemical reactions with the deep-sea sediments by phenomena such
as adsorption, ion exchange, and colloid filtration. These mechanisms
are usually combined into one general term called sorption. Many
waste elements with long half-lives such as plutonium, react with the
clay sediment so that some of the element is sorbed to the sediment
and some remains dissolved in the pore water. Sorption is expressed
in terms of distribution coefficients, K,. and is the ratio of the
d
sorbed and dissolved concentration of isotope in the sediment. Because
only the dissolved fraction diffuses through the sediment, the rate
of diffusion of a reactive isotope is much slower than the rates for
nonreactive elements such as iodine and tritium. The K, values, are
d
dependent on such parameters as pH of the water, the specific nuclide
present, the concentration and type of dissolved ions, and temperature.
The effectiveness of deep sea sediments to act as a retarder
for a particular condition is expressed as the retardation factor, R,.
For a particular radionuclide, R is defined as the ratio of the
water velocity to the nuclide migration velocity (dimensionless
term). The retardation factor is related to the distribution
coefficients by the following relationships:
R - 1 + K. P/E
d d
where P = bulk density of the sediment
E = porosity (ratio of the volume occupied by pores
to the total volume of the sediment).
7-28
-------
The magnitude of radionuclide migration retardation that can be rea-
lized may be expressed by relating the velocity of ions moving through
the sediment to the interstitial velocity of water flow by the follow-
ing equation:
V
v - -Z
1 Rd
where V - velocity of the ionic isotopes
V - interstitial velocity of water flow
R, » retardation factor
d
Estimated distribution coefficients (K,), retardation factors (R.), and
u a
relative transport rates of elements in soil to that in water (V./V )
in a typical desert soil are shown in Table 7-II.
Although the'distribution coefficients and retardation factors
shown in Table 7-II are estimates of migration of radionuclides in
typical desert soils, they do give a perspective and order of magni-
tude for K, values for ocean sediments. Actual values for ocean sedi-
Q
ments may differ substantially. The collection of data on the solu-
tions formed by reactions between pore waters and specific radioactive
wastes, and on the distribution coefficients for elements in such
solutions, are task areas presently being undertaken at Sandia Labora-
tory. Distribution coefficients must be determined for all long-lived
radionuclides as a function of sediment type. Some distribution co-
efficients are known for several candidate deep-sea sediments. These
are summarized in Table 7-III.
From Tables 7-II and 7-III, the following is apparent:
• soil has greater retention for most of the long life radio-
nuclides (actinides)than the short lived radionuclides,
7-29
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TABU 7-tI
ESTIMATED DISTRIBUTION COET7ICIZNTS (I.) AND
UTAUAIION FACTORS (»d) IN A H7ICAL OESZW SOIL
asmtT
Tritium
Chlorine
Argon
Krypton
Taehaetium
Iodine
Aatatine
Radon
Carbon
Thallium
Molybdenum
Sodium
Bismuth
Calcium
Antimony
Neptunium
Selenium
Strontium
Polonium
Potassium
Beryllium
Cobalt
Nickel
Radium
Rubidium
Iron
Cesium
Francium
Palladium
Tin
Promethium
Samarium
Europium
Hrtlnvi iim
Curium
Berkelium
Actinium
Yttrium
Zirconium
Niobium
Plutonium
Amarlcium
Lead
Protactinium
Thorium
*d «/,)•
0
0
0
0
0
0
0
0
2
2
5
"10
10
13
15
15
20
20
23
35
73
75
30
100
125
130
200
200
250
250
600
600
600
600
600
700
1,000
2.000
2,000
2,000
2,000
2,000
2,000
4,000
4,000
15,000
V
1
1
I
1
1
1
1
1
10
10
25
50
50
100
100
100
100
100
110
170
330
330
330
500
500
3,300
1,000
1,000
1,100
1,100
2,500
2.500
2.500
2.500
3,300
3,300 <
5,000
10,000
10.000
10,000
10,000
10, 000
10,000
16,700
16,700
50,000
Vt/Vw t
1
1
1
1
1
1
1
1 .
txlO"1
1x10" l
4xlO"2
2x10' 2
2xiO-2
IxlO-2
IxlO'2
IxlO-2
IxlO-2 •
IxlO-2
9x10" 3
6x10" 3
3x10-3
3x10-3
3x10-3
2x10-3
2x10-3
3xlO~4
1x10-3
1x10-3
9xlO"4
9xiO"4
4xlO-4
4xlO"4
4x10
4xlO"4
3x10
3xlO'4
2x10
IxlO"4
IxlO-4
1x10
IxlO-4
IxlO-4
IxlO"4
6x10-5
6x10-5
2X10'5
NOTES:
•Equilibrium distribution coefficients betveen vatar and soil.
•Hletardaeion Factor (Rd) - Vw/Vi
^Relative transport rate of elements in assumed soil to that
in water.
Reference: Assessment of Geologic Site Selection Factors, Subtask C-
Report, Arthur D. Little, Inc. Movember 1977.
7-30
-------
TABLE 7-III
ESTIH4TED DISTBI3CFTI0U C0EFF1CIOIS
19 DEZP-SEA.
ELEMENT
Sr
Cs
Pu
U
SZDDffiHT I £j
i i
Montaorillite
Kaolin! te
1111 t*
104
15 i
'150 '
Calcice i 1 i
j j
Montaorillite j 4,400 j
Kaoliaite i 45 |
Illlte
Montoorlllite
400 |
630
Kaoliaite i 352
Illite 129
Illlte
139
Source: L.L. Ames, D. Pal, "Sadionuclides Interaction with
Soil and Rock Media, "Vol. 1, EPA 520/6-78-007, 1978
7-31
-------
• Considering Cs and Sr, clay minerals may have higher
retention (higher K
-------
sediments as a barrier to radionuclide migration. Due to low thermal
conductivity, high temperatures will exist around the canisters. After
initial emplacement, the temperature of the surrounding sediment may
be as high as 600 C. Substantial thermal gradients may exist around
each container with temperatures declining to that of the surrounding
sediments 10-20 meters away, (0 C). Such gradients give rise to up-
ward pressure gradients which will cause water to migrate. This may
well produce an upward flow of pore water away from the waste canister
that will tend to carry the radionuclides toward the sea floor.
As previously noted, the high temperatures surrounding the waste
canister could also have the effect of fluidizing the sediment/pore
water. The canister, assuming a greater density than surrounding sedi-
ments, could sink to greater depths. In the event that failure of the
canister released a sufficient quantity of heat-producing radioactive
waste, such that the sediment/pore water was maintained in the fluidized
state, it is conceivable that convective upward transport could occur.
Present information on the physical behavior of the sediment is not
available to determine if this process is possible.
Fortunately, the heat released by the radioactive waste will be
reduced after several hundred years. For example, after a thousand
years, the temperature of the surrounding sediment may be reduced to
about 200°C because of radioactive decay of short-lived fission pro-
ducts. Thus, containment of the radioactive elements by the solid
waste form and by the canister is very important to minimize any
7-33
-------
dispersion of these isotopes due to thermal effects for several
hundred years after burial.
The effects of temperature on the distribution coefficient and
retardation factors for the radioisotopes and their chemical com-
pounds is also important in determining the isolation capabability
of the deep sea sediments.
Rock Emplacement
Disposal in the deeper lithified sediments (at a depth of greater
than 500 meters) is also being considered under the Seabed Disposal
Program. Unlike the deep-sea sediments, the bedrock layers are suscep-
tible to fracturing that could lead to fast migration of fluids along
cracks. The fracturing is due primarily to higher shear strength and
reduced plastic properties of these sediments. The transition down
from soft deep-sea (clay) sediments to lithified deposits may be
gradual or abrupt, and sometimes alternating layers of bedrock and
soft clays are found. Data on lithified sediments below the sea floor
have been obtained from Deep-Sea Drilling Project experiments.
Disposal within igneous rock beneath the ocean sediments has
been considered, but only limited experimentation has been conducted.
To date only a few holes have been drilled 500 meters or more into
igneous rock by the Deep-Sea Drilling Project. From the few experi-
ments conducted, the basement rock is comprised of the following:
a layer of basaltic pillow lavas resulting from underwater
eruption and rapid chilling of molten lava
7-34
-------
• fractured blocks and breccia
• sediment-filled cavities and inter-layered sediments over-
lying quantities of basalt
• basaltic dykes at greater depths
The whole basement complex is cut by fractures and fissures at
depths of 1.00 meters or more. Because the exact nature and predict-
ability of these rocks is poorly known, neither b4sement rock nor
the overlying lithified sediments are being considered as disposal
sites at the present time.
3 4
7.3.1.A Ocean. ' The ocean water is likely to be a poor
barrier for large quantities of released nuclides but provides some
protection against inadvertent release of smaller amounts such f>s
might be released from a single canister. Transport and dispersion
through the ocean can occur due to a number, of conditions:
• deep horizontal advection
• deep vertical mixing
• surface currents
• biological transport both horizontal and vertical
• thermal plume
• adsorption onto falling debris
• turbulent eddies
Material balance arguments and the age of the bottom water in the
mid-plate/mid-gyre regions of the ocean indicate that the movement of
dense water from the ocean bottom to areas where this water is re-
turned to the surface layers takes from 1000 to 2000 years. Studies
7-35
-------
have indicated Chat the mixing time for the Pacific Ocean waters is
1000 to 1600 years or nearly twice as long as that of the Atlantic.
Knowledge of transport mechanisms of radionuclides through ocean
water is far from complete. Data needs to be gathered in che
following areas:
• bulk diffusion and advection coefficients
• effects of eddies and currents
• radionuclide scavaging by particulates in ocean columns
• biological transport through the food chain
Studies have indicated that the biological community either in the
surface waters or on the bottom may provide a path for both horizontal
and vertical transport.
7.3.1.5 Summary - Barrier Effectiveness for Waste Isolation. In
Section 7.3.1, the emplacement of high-level wastes in geologic
formations underlying the ocean floors was discussed relative to the
technical feasibility of seabed disposal. The technical feasibility
depends upon demonstrating that seabed disposal can contain radioactive
waste long enough for it to decay to innocuous levels or not to
exceed established radiation standards.
Physical and environmental barriers exist which may prevent
migration of radionuclides to ocean areas of immediate significance to
mankind. These mechanisms of breaching these barriers include the
following:
• corrosion of the canister
• leaching of the waste material
7-36
-------
• upward transport through the deep-sea sediments
• transport through the ocean columns
The rates of radionuclides migration for all of these processes have
been estimated in Section 7.3.1. Because data on the rates of trans-
port for all radionuclides in a varied sample of deep-sea sediments
is insufficient in many cases, the estimates of mechanisms by which
radionuclides are transported is based on generalized information of
the ocean environment and will contain many uncertainties. The
potential effectiveness of the barriers for waste isolation for
several radionuclides is provided in Table 7-IV. For purposes of this
estimate, it is assumed that canisters will provide an effective
barrier for 1000 years, the waste form will exist for 1000 years, the
sediment will delay radionuclide release to the ocean for 10 years, and the
ocean will delay radionuclide.entry to the human environment for 1000
years. Further research is obviously required to support these
assumptions. Table 7-IV, therefore, only represents the potential
barrier effectiveness.
7.3.2 Research Needs
The investment required to develop the necessary baseline infor-
mation regarding ocean characteristics, emplacement techniques, and
engineering and environmental barriers against waste intrusion into
the biosphere from seabed disposal may be significant. There are
large gaps in information required to understand the entire ocean-
sedimenc waste system that is necessary to adequately assess the
technical feasibility of seabed disposal.
7-37
-------
TABLE 7-IV
POTENTIAL BARRIER EFFECTIVENESS FOR WASTE ISOLATION
Retardation Factor**
Barriers Adequate to
Allow Nuclide to Decay to
Nuclide
Cs-134
Co-60
Kr-85
H-3
Pu-241
-g Eu-154
i
CO
00 Sr-90
Cs-137
Cm-243
Pu-238
Sm-151
Am-242M
Am-24l
Ra-226
'1/2
2.05y
5.24y
10. 8y
12. 3y
13. 2y
16y
27. 7y
30y
32y
86y
87y
1.5 x 102y
4.58 x 102y
1.6 x 103y
10tl/2*
20. 5y
52. 4y
1.08 x 102
1.23 x 102
1.32 x 102
1.6 x 102
2.77 x 102
3.0 x 102
3.2 x 102
8.6 x 102
8.7 x 102
1.5 x 103
4.58 x 103
1.6 x 104
y
y
y
y
y
y
y
y
y
y
y
y
y
y
Rd
1,000
1
1
10,000
2,500
100
1,000
10,000
10,000
2,500
10,000
10,000
500
Innocuous Levels***
A
A
A
A
A
A
A
A
A
B
B
B
C
C
-------
TABLE 7-IV (Continued)
Retardation Factor**
Barriers Adequate to
Allow Nuclide to Decay to
LO
VO
Nuclide
Cm- 246
C-14
Pu-240
Th-229
Am- 24 3
Cm- 24 5
Pu-239
Th-230
U-233
U-234
Pu-242
Cm- 248
Np-237
Cm- 24 7
1-129
U-236
S/2
3
4.7 x 10 y
•»
5.7 x 10 y
6.58 x 103y
7.34 x 103y
7.4 x 103y
9.3 x 103y
2.44 x 104y
8 x 104y
1.62 x 105y
2.47 x 105y
3.79 x 105y
4.7 x 105y
2.14 x 106y
1.6 x 107y
1.7 x 107y
2.39 x 10?y
10tl/2*
4
4.7 x 10
L
5.7 x IQ
6.58 x 104
7.34 x 104
7.4 x 104
9.3 x 104
2.44 x 105
8 x 105
1.62 x 106
2.47 x 106
3.79 x 106
4.7 x 106
2.14 x 107
1.6 x 108
1.7 x 108
2.39 x 108
y
y
y
y
y
y
y
y
y
y
y
y
y
y
y
y
Rd
3,300
10
10,000
50,000
10,000
3,300
10,000
50,000
14,300
14,300
10,000
3,300
100
3,300
1
14,300
Innocuous Levels***
C
C
C
C
C
C
C
C
E
E
E
E
E
E
E
E
-------
TABLE 7-IV (Concluded)
Retardation Factor**
Barriers Adequate to
Allow Nuclide to Decay to
Nuclide t /n lOt ,„*
LI z ii z
Pu-244 8 x 107y 8 x 108 y
U-235 7.1 x 108y 7.1 x 109 y
Q 10
U-238 4.5 x 10 y 4.5 x 10 y
10 11
Th-232 1.4 x 10 y 1.4 x 10 y
*99.9 percent decayed.
**Retardation Factors (Rj) represent estimates
R_, Innocuous Levels***
Q '
10,000 E
14,300 E ...•<-
14,300 E
50,000 E
for each isotope in soils based on Analysis
of Migration Potential, Subtask C-2 Report, Arthur D. Little, December 1977.
***A = canister
B = canister + waste form
C = canister + waste form + sediment
D = canister + waste form + sediment + ocean
E = the retardation factor will be significant in preventing the escape of the radionuclide
Innocuous levels mean less than 0.1 percent of the original activity remains.
-------
7.3.2.1 Ecological Implications of Thermal Waste Heat. By
affecting the physical/chemical conditions in its surroundings, the
placement of radioactive wastes may induce ecological changes. Since
waste disposal sites are areas of 'low biological productivity, the
major effect of thermal waste heat is likely to be one of increased
biological activity. Three major factors must be examined to assess
the ecological implications of thermal waste heat:
• increase in biological activity may increase the rate at
which the canisters are decomposed
• increase in biological activity may increase the rate at
which radionuclides are transported through the sediments to
the surface waters
• higher biological productivity which may result from increased
temperatures may be counteracted by the biologically deleter-
ious effects of ionizing radiation
7.3.2.2 Hole Closure. ' ' Any emplacement procedures will
disrupt the sediment layer of the ocean floor. In order to ensure
safe emplacement, it is necessary to examine the response of clay
sediments to canister emplacement, particularly the hole closure
properties of clays. To prevent a decreased migration time of the
clay barrier, it is essential that the hole created by emplacement of
canisters be filled either with the same type of sediment or with' a
suitable sealant.
Laboratory and field experiments are underway at Sandia
Laboratories to examine sediment behavior during and subsequent to
penetration by waste canisters. These initial experiments indicate
that closure of a completely penetrating projectile (such as the free
fall emplacement method) would be immediate and total, while closure
of a hole left open by an emplacement rod would be gradual.
7-41
-------
7.3.2.3 Summary of Other Data Requirements. Areas which
require further information to adequately assess the technical
•' - " ' ' ! J
feasibility of seabed disposal, particularly its ability to act as a
barrier to radionculide migration, include the following:
• information on the characteristics of ocean provinces to
determine and establish.-their overall suitability as potential
seabed disposal sites
• technological capabilities including transportation, ship-
ment, and emplacement of wastes
• corrosion properties of canister materials at high tempera-
tures and pressures
• leach rates for all radionuclides in proposed waste forms
• physical properties of deep-sea sediments
• sorption and distribution coefficients of deep-sea sediments
• retardation factors of sediments
• effects of thermal gradients on sediments (heat transfer
properties)
• dynamic response of sediment to canister emplacement;
• transport processes of radionuclide in deep sea sediments
including structural and chemical properties and driving
forces
• transport processes in the. water column, including diffusion
currents, advection, biological (food web), and thermal
plume
7.3.3 Radiological Impact Assessment-' ' '
This section will assess -the- potential radiological consequences
to man of solidified high-level radioactive vns-t* which is emplaced
in deep-ocean sediments. The principal route of return to man that
is considered in this assessment is via dispersion in the deep ocean,
physical transport to the productive surface layers, incorporation
in marine food chains, and consumption of contaminated seafoods by
7-42
-------
man. The consequent radiation exposure to man will be assessed in
terms of both individual and collective doses. Radiation doses
arising from concentration of beach sediments are also considered.
In addition, operational and transportation accident risks will be
discussed. The discussion presented here will rely heavily on the
information provided in the previous sections.
It is intended that only broad conclusions be drawn from this
section. In the course of discussion, those subject areas where
more study or information is required to complete a radiological
impact analysis will be highlighted. Most of the information
contained in this section has been abstracted from two reports:
Assessment of the Radiological Protection Aspects of Disposal of
High Level Waste on the Ocean Floor, Grimwood and Webb, National
Radiological Protection Board NRPB-R 48 (1976); and Consultants
Meeting to Review the Radiological Basis of the Agency's Provisional
Definition and Recommendations for the London Convention, Inter-
national Atomic Energy Agency (IAEA), June 1977, London, England.
These reports have attempted to assess radiation dose to man and
possible damage to the marine ecosystem based on models which evaluate
release rates and pathways of radionuclides to man. The reports
are preliminary and contain large gaps in the information that would
be necessary to complete an Environmental Impact Statement (EIS) on
the radiological impacts of Seabed Disposal. No attempt has been
made in either report to establish radiation protection standards
although the criteria for such assessments have been addressed.
7-43
-------
7.3.3.1 Source Term. In the context of a rapidly expanding
commercial nuclear program, cdncern is often expressed with regard to
final disposal and potential for release of long-lived radionuclides
to the environment. In the case of seabed disposal, Section 7.3.1
discussed the effectiveness of engineering and environmental barriers
against waste intrusion into the oceans. A summary of the effective-
ness of these barriers for waste isolation for several radionuclides
was illustrated in Table 7-IV. The conclusion drawn from Table 7-IV
is that the combination of environmental and engineering barriers may
be inadequate to allow the radionuclides with long half-lives to decay
to innocuous levels before spreading to productive surface layers of
the ocean. Further, the assumptions upon which the barrier effective-
ness is estimated are unproven. Therefore, earlier, though gradual,
or later release of radionuclides may be expected.
When account is taken of the quantities of various nuclides
emplaced in the seabed, their half-lives and their dispersibility,
those radionuclides likely to be most significant in terms of radia-
tion exposure to man and potential damage to the marine ecosystem are
the volatile radionuclides (C-14, and 1-129), and the long-lived
actinides. To illustrate this point, the amount of each nuclide
which would initially be present in a seabed repository is listed in
Table 7-V for three cases, the throwaway fuel cycle, U02 recycle, and
mixed oxide recycle. These amounts (expressed in grams) are based on
50,000 MTHM charged to the reactor, and a 10 year cool-off period.
If all radioactive wastes from U.S. nuclear power production were
buried in the sea, these initial quantities would be much larger,
particularly if the current backlog of stored waste was buried in the
7-44
-------
TABLE 7-V . ' •
RADIONUCLIDE AMOUNTS IN INITIAL SEABED REPOSITORY•
Initial Mass in Place **
(g)
Nuclide
Cs-134
H-3
Pu-241
Eu-154
Sr-90
... 37
Cm-243
Pu-238
Sm-151
Am-242m
Am-241
Ra-226
Cm-246
C-14
Pu-240
Th-229
Am-243
Cm- 24 5
Throwaway
3.53 x IO5
2
3
1
2
4
3
6
2
2
2
1
1
1
1
2
4
1
.5
.91
.79
.12
.96
.6
.50
.14
.26
.53
.48
.48
.71
.11
.10
.46
.25
x IO3
x IO7
x IO7
x IO7
x IO7
x IO3
x IO6
x IO6
x IO1
x IO7
x IO"2
x IO3
x IO2
x IO8
x 10~2
x IO6
x IO4
U00 Recycle
3
1
1
1
2
4
3
2
2
2
3
5
1
1
9
9
4
1
.53 x
.7 x
.96 x
.79 x
.•12 x
.96 x
.6 x
.14 x
.14 x
.26 x
.28 x
.50 x
.48 x
.71 x
.10 x
.60 x
.43 x
.25 x
io5
io2
10
10
10
.10
5
6
7
7
io3
10
k
10
10
10
10
10
10
10
10
10
5
6
1
6
-3
3
2
5
~3
6
io4
Mixed
Oxide Recycle
1
2
1
1
1
5
3
3
3
4
5
1
3
2
1
1
1
4
.55
.27
.67
.24
.23
.10
.62
.87
.62
.95
.25
.97
.18
.17
.68
.35
.27
.46
x
x
X
X
X
X
X
X
.X
X
X
X
X
X
X
X
X
X
io5
io2
io6
io6
io7
io7
io4 .
IO6
IO6.
io2
io7
io-11
io5
IO1
io7
io'3
io8
IO6
7-45
-------
TABLE.7-V (Concluded)
RADIONUCLIDE AMOUNTS IN INITIAL SEABED REPOSITORY*
Initial Mass in Place **
(g)
Nuclide
Pu-239
Th-230
U-233
U-234
Pu-242
Cm-248
Np-237
Cm-247
1-129
U-236
Pu-244 .
U-235
U-238
Th-232
information
Throwaway
2
2
2
9
2
1
2
1
1
2
2
4
4
6
.70
.91
.60
.15
.24
.41
.35
.99
.16
.05
.56
.02
.72
.70
based
X
X
X
X
X
X
X
X
X
X
X
X
X
X
on
10
10
10
10
10
10
10
10
10
10
10
10
10
10
8
2
^
2
6
7
o
\J
7
/
1
7
8
-5
8
\J
10
1
Analysis
UO,, Recycle
1
5
7
5
1
1
2
1
1
1
2
2
2
1
of
.36
.85
.20
.90
.12
.41
.33
.99
.46
.03
.38
.01
.36
.17
X
X
X
X
X
X
X
X
X
X
X
X
X
X
10
10
10
10
10
10
10
10
10
10
10
10
10
10
6
1
X
1
4
5
o
VJ
7
/
1
4
6
-5
6
\J
8
1
Migration
Mixed
Oxide Recycle
3.76 x IO6
-7
2.19 x 10
2.59 x IO2
2.57 x IO6
2
2.44 x 10
6
6.90 x 10
4.20 x IO3
1.86 x IO4
1.54 x IO5
4.12 x 10~3
7
9.50 x 10
2.26 x IO8
1.47 x 10°
Potential, Subtask C-:
Report, Arthur D. Little, December-1977.
**Based on 50,000 MTHM charged to the 'reactors, and a 10-year cool
off period. .
7-46
-------
sea. The Grlmwood and Webb model, for example, assessed the potential
radiological consequences of seabed disposal based on the total
high-level waste which would be generated by a postulated world
nuclear program of nuclear power production to the year 2000.
The initial quantities listed in Table 7-V could easily be scaled
to represent sowce terms which reflect the quantities of radioactive
wastes from U.S. power production to the year 2000.
CJ A I I rt
7.3.3.2 Environmental Paciiways to Man. ' ' ' After radioactive
wastes migrate through the environmental and engineering barriers
discussed in Section 7.3.2, the principal mechanisms by which
radionuclides reach man are dispersion of waste material in the
deep ocean, physical transport to productive surface layers; incorpora-
tion in marine food chains; and consumption of contaminated seafoods
by man or exposure of man to contaminated beach sediments.
The lowest trophic level of the marine food chains is plankton,
Phytoplankton constitutes the largest single source of biomass in the
oceans and accumulates nutrient elements directly from the water.
Light is necessary for photosynthesis by phytoplankton. If they are
carried by currents to deeper waters, the lack of illumination will
eventually cause their death. The major portion of the oceans in
which incorporation of elements into the food chains occur is,
therefore, the surface layers to a"depth of 200 meters.
Zooplankton, the next higher trophic level, includes groups
which are omnivores as well as carnivores. They derive most of their
7-A7
-------
food either directly or indirectly from the phytoplankton layer.
Zooplankton are found at all depths in the oceans but the extent of
their vertical migrations is usually a few hundred meters and the
biomass per unit volume is much lower at depths below a few hundred
meters than in the surface layers.
The present marine food sources utilized directly by man come
from higher nektonic trophic levels than the plankton. Both pelagic
and benthic animals constitute important food sources, the most
important both in terms of numbers and availability being the near-
shore pelagic and benthic groups; the open-ocean pelagic groups being
of intermediate value and the open-ocean benthic groups being by far
the least important both at present and in future potential.
Marine organisms can accumulate radionuclides from food, water
and suspended or deposited sediments. For phytoplankton, accumula-
tion of activity occurs via direct uptake from the water in a similar
manner to their uptake of nutrients. For zooplankton, the major
source of radionuclides is the water but considering the relative
quantities involved, it seems most of.the uptake occurs via food,
except for those nuclides which are only slightly concentrated in
food. For other nekton, the majority of the activity is taken in via
food rather than water.
The concentration of a radionuclide in a given organism may be
greater or less than the concentration in the surrounding water, the
ratio being known as the concentration factor. Although the uptake
7-48
-------
of radionuclides by organisms is a dynamic process which depends on many
variables (including the,physio-chemical state of the activity.
temperature and salinity of the water, growth rate and physiological
state of the organism), the concept of the concentration factor is
meaningful in an environment such as ocean transport which changes
slowly compared with the turnover rates of activity in the organisms
comprising the food chain.
In order to calculate the eventual return of radioactivity to
man via the marine food chains, it is necessary to estimate values
for appropriate concentration factors and to define pathways and
associated modes of exposure to man. For mixed marine plankton, a
4
concentration factor of 10 'is typical for many of the radionuclides,
although concentration factors for many individual radionuclides are
not available. Concentration factors for marine molluscs, Crustacea,
and fish are, in general, better known although, for some nuclides,
there still -is considerable uncertainty. Table 7-VI lists concentra-
tion factors for some of the major radionuclides. They have been
taken from values given in several review documents and are thought
to represent realistic values for edible flesh of these organisms. A
list of pathways and modes of exposure to man for various radionu-
clides is shown in Table 7-VII.
7.3.3.3 Nuclides of Importance if Barriers Maintain Expected
Integrity. As discussed in Section 7.3.1, engineering and environ-
mental barriers exist which may prevent migration of radionuclides to
7-49
-------
TABLE 7-VI
CONCENTRATION FACTORS
NUCLIDE
H-3
Se
Sr
Zr
Nb
Tc
Pd
Sn
Sb
Te
I
Cs
Pm
Sm
Eu
Pb
Po
Ra
Ac
Th
Pa
U
Np
Pu
Am
Cm
Source: Assessment
High-Level
CONCENTRATION
FACTOR FOR FISH
1
1000
1
30
30
10
10
1000
300
10
10
30
30
30
30
300
300
100
30
10000
10
10
10
10
10
10
of Radiological Protection
Waste on the Ocean Floor,
CONCENTRATION
FACTOR FOR
MOLLUSCS OR CRUSTACEA
1
300
3
100
100
100
300
300
300
1000
100
30
1000
1000
1000
100
3000
1000
1000
1000
10
10
10
300
1000
. 1000
Aspects of Disposal of
Grimwood and Webb,
NRPB-R 48 (1976).
7-50
-------
TABLE 7-VJI
PATHWAYS TO MAN AND MODES OF EXPOSURES
NUCLIDE
PATHWAY
MODE OF EXPOSURE
H-3
C-14
Co-60
Sr-90
Ru-106
1-129
1-131
Cs-134
Cs-135
Cs-137
Eu-154
Ra-226
Th-229
Th-230
Th-232
U-233
U-234
U-235
U-238
Np-237
Np-238
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Am-241
Am-242
Am-243
Cm-242
Cm-243
Cm-244
Cm-245
Cm-248
Miscellaneous
Fish Consumption
Crustacea Consumption
Mollusk Consumption
Seaweed Consumption
Beach Dwellers
Seaweed Consumption
Seaweed Consumption
Seaweed Consumption
Seaweed Consumption
Beach Dwellers
Fish Consumption
Fish Consumption
Beach Dwellers
Fish Consumption
Fish Consumption
Fish Consumption
Fish Consumption
Seaweed Consumption
Seaweed Consumption
Seaweed Consumption
Seawaed Consumption
Seaweed Consumption
Fish Consumption
Seaweed Consumption
Seaweed Consumption
Seaweed Consumption
Seaweed Consumption
Seaweed Consumption
Seaweed Consumption
Seaweed Consumption
Seaweed Consumption
Seaweed Consumption
Seaweed Consumption
Seaweed Consumption
Seaweed Consumption
Seaweed Consumption
All
Ingestion
Ingestion
Ingestion
Ingestion
External Irradiation
Ingestion
Ingestion
Ingestion
Ingestion
External Irradiation
Ingestion
Ingestion
External Irradiation
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Ingestion
Source: Consultants Meeting to Review the Radiological Basis of The
Agency's Provisional Definition and Recommendations For The
London Convention, IAEA June 13-17, 1977, IMCO Headquarters,
London, England.
7-51
-------
ocean areas long enough for them to decay to innocuous levels. These
barriers included containment in metal canisters, solidified waste
forms, deep-sea sediments, and the ocean column. The potential of
the barriers' effectiveness for waste isolation for several fission
products and actinides was illustrated in Table 7-IV.
If engineei'ing and environmental barriers are assumed to be
effective for 10 years, the elements identified by "E" on
Table 7-IV are the waste isotopes which pose the greatest environmental
impact. However, if these barriers are ineffective in preventing
migration and dispersion in the deep ocean for 10 years, several
other fission products with intermediate half-lives may escape con-
tainment and become dispersed into the deep ocean before they will
have decayed to innocuous levels. This will have the net effect
of increasing the radiological impacts to man.
Using the quantities of fission production and actinides ini-
tially present in a seabed repository (see Table 7-V), and assuming
a barrier effectiveness of 10 years, the amount of actinides and
fission products which could potentially be dispersed into the deep
ocean is calculated in Table 7-VIII.
As shown in Table 7-VIII, several radionuclides will still be pre-
sent in large quantities after 10 years of decay. These radionuclides
include: U-234, U-235, U-236, U-238, Pu-242, Np-237, and 1-129. If
the initial quantities of wastes listed in Table 7-V were scaled to
represent all radioactive wastes from U.S. nuclear power production,
7-52
-------
TABLE 7-VIII
RADIONUCLIDE AMOUNTS AFTER 10 YEARS OF DECAY
en
Co
Nuclide
Cs-134
H-3
Pu-241
Eu-154
Sr-90
Cs-137
Cm-243
Pu-238
Sm-151
Am-242m
Am-241
Ra-226
tl/2
2.05y
12. 3y
13. 2y
16y
27. 7y
30y
32y
86y
87y
1.5 x 102y
4.58 x 102y
1.6 x 103y
Throwaway
0
0
0
0
0
0
0
0
0
0
0
0
Mass in place after 10 years (g)*
U02 Recycle
0
0
0
0
0
0
0
0
0
0
0
0
Mixed Oxide Recycle
0
0
0
0
0
0
0
0
0
0
0
0
-------
TABLE 7-VIII (Continued)
RADIONUCLIDE AMOUNTS AFTER 10 YEARS OF DECAY
Mass in place after 10 years (g)*
I
en
Nuclide
Cm-246
C-14
Pu-240
Th-229
Am-243
Cm- 24 5
Pu-239
Th-230
U-233
U-234
Pu-242
Cm-248
4
5
6
7
7
9
2
8
1
2
3
4.
tl/2
.7 x IO3
.7 x IO3
.58 x 10
Throwaway
y
y
3
y
.34 x 103y
.4 x IO3
.3 x IO3
.44 x 10
x 104y
.62 x 10
.47 x 10
.79 x 10
7 x 105y
y
y
4
y
y
5
y
y
i
2
2
2
9
5
1
•5
4
5
3
3
.36 x
.70 x
.02 x
.08 x
.51 x
.43 x
.25 x
.03 ,x
.99 x
.53 x
.60 x
.23 x
io-61
io-51
io-38
io-43
io-35
io-29
io-4
ID'2
10°
IO5
ID6
Hf1
U00 Recycle
1.
2.
1.
9.
9.
5.
6.
1.
9.
3.
1.
3.
36 x
70 x
66 x
52 x
51 x
43 x
29 x
01 x
99 x
57 x
80 x
23 x
io-61
io-51
io-40
io-44
io-35
io-29
io-7
io'2
io-1
io3
IO4
1Q-1
Mixed Oxide
2.93 x
3.43 x
3.06 x
1.33 x
2.71 x
1.94 x
1.74 x
3.79 x
3.59 x
4.13 x
5.59 x
Recycle
1Q-59
io-52
io-39
lO'44
io-33
io-26
io-6
io-11
10°
IO5
io1
-------
TABLE 7-VIII (Concluded)
RADIONUCLIDE AMOUNTS AFTER 10& YEARS OF DECAY
Mass in place after 10 years (g)*
Nuclide
Np-237
Cm-247
1-129
U-236
Pu-244
U-235
7" U-238
Ol
Th-232
tl/2
2.14
1.6
1.7
2.39
8 x
7.1
4.5
1.4
&
x 10"y
x 10
x 10
7
y
i
y
7
x 10 y
7
10 y
x 10
x 10
x 10
8
y
9
7
y
10
\.\9
y
Throwaway
1.
1.
1.
1.
2.
4.
4.
6.
70 x
91 x
11 x
99 x
54 x
02 x
72 x
70 x
7
10
I
10
io7
8
10
-5
10 3
8
10
10
10
*
10
DO Recycle Mixed Oxide Recycle
1
1
1
1
2
2
2
1
7
.69 x 10
1
.91 x 10
.40 x 10
4
.00 x 10
_c
.36 x 10
&
.01 x 10
8
.36 x 10
•
.17 x 10
4
4
1
1
4
9
2
1
.99 x
.02 x
.79 x
.50 x
.08 x
.49 x
.26 x
.47 x
6
10
3
10
io4
5
10
_o
10
7
io'
8
10
0
10"
Calculations of the mass in place after 10 years of decay, is based on the decay formula
N = N0e~ . The values for No were taken from the values from initial amounts of radio-
nuclides in a seabed repository from Table 6.1.
-------
particularly projections of accumulated waste through the year 2000,
then the quantities of radioactive materials remaining after 10 years
of decay would be significantly greater than the amounts shown in
Table 7-VIII. If the radionuclides become widely dispersed in the
/leep ocean, then the radiological impacts on marine organisms may be
less significant. However, if dispersion and physical transport of
these wastes is localized, marine organisms as well as suspended
sediments may receive large doses of radioactivity which, in turn,
will be incorporated in marine food chains.
If engineering and environment barrier integrity is not main-
tained for -10 years, significant quantities of radionuclides with
intermediate half-lives (i.e., 10 -10 years) may be dispersed in the
deep ocean and will undergo similar physical transport to productive
layers of the ocean and, in turn, incorporated in marine food
chains. The integrity of environmental barriers depends heavily on
the transport mechanisms of radionuclides through the deep-sea sedi-
ments (retardation factors). As discussed in Section 7.3.1, research
and experimentation on retardation factors for radionuclides in deep-
sea sediments is being conducted but established data on these coef-
ficients is currently not available.
7.3.3.4 Dose Assessment. This section discusses radiation
exposure to man from seabed disposal in terms of both individual and
collective doses. The data and results contained herein are abstracted
7-56
-------
from Assessment of the Radiological Protection Aspects of Disposal of
High Level Waste on the Ocean Floor, Grimwood and Webb, NRPB-R 48
(1976). Two models were developed in this report which characterize
the physical transport and mixing processes in the ocean, as well as
incorporation in marine food chains and ultimate consumption of
seafoods and radiation exposures to man. These models contain many
assumptions and input data which will not be discussed here.
The following is a brief summary of the most significant findings
of NRPB-R 48 and other conclusions from previous sections concerning
the radiological implications to man from seabed disposal:
ICRP Recommendations
• In order to provide a basis for comparison with individual
and collective dose estimates from seabed disposal, the maxi-
mum permissible annual intakes (MPAI) of activity by ingestion
for individual members have been calculated for the principal
radionuclides;
• ICRP recommended maximum permissible dose rates for external
exposure are 0.5 rems y-1 for whole body irradiation and
3 rems y~l for skin;
• ICRP have made no specific recommendations on collective dose
limits.
Doses to Individuals via Critical Pathways
• The highest ratios of individual doses to the appropriate dose
limit (or intake (I) to the MPAI) are for the potential routes
involving consumption of deep-ocean fish or plankton. The
maximum values of I/MPAI are of the order of 10~2 for both
routes. The times at which these maximum values occur tend
to be either short (50-100 years) or intermediate (500-2000
years). Critical organs are usually bone for Sr-90 and the
actinides, and whole body for Cs-137.
7-57
-------
• The highest predicted intakes by individuals in the critical
group due to consumption of surface fish are of the order of
10~3 to 10"^ of the MPAI for fission products at 50 years, and
may reach 10~5 of the MPAI for the actinides at 105 to 106 years,
• Similar types of results are obtained from the consumption of
deep ocean fish except that the predicted intakes are one to
two orders of magnitude higher than surface fish.
• For consumption of plankton, only Sr-90 has a significant
predicted intake with a ratio of 4 X 10~2 at 100 years. Two
actinides of comparable importance are Am-241 and Am-243.
• Postulated intakes from consumption of molluscs or Crustacea
are less than via the routes already mentioned.
• Intakes from drinking desalinated water are low.
• External doses from contamination of coastal sediments are
comparable fractions of the dose limit for both skin & whole
body irradiation. The highest doses in both categories are
given by Cs-137 which would deliver 3 x 10~3 of the whole body
dose limit and 7 x 10"^ of the skin dose limit. The calculated
doses are at a maximum after only 100 years, and it is most
unlikely in practice that the coastal sediments would become
contaminated so quickly.
Collective Doses
• The only intake route actually established for collective
doses is via consumption of surface fish. The nuclides that
are responsible for the maximum individual doses give rise to
the maximum collective doses and the same limitations on the
accuracy of the available information also apply.
• The largest annual collective dose to the whole body due to
consumption of surface fish is about 4 x 1Q4 man rems at 10
years from Cs-137 and Sr-90 taken together. Collective doses
to the whole body at longer times will be of the order of 102
to 10^ man rems per year. Nuclides which contribute include
Am-241, Am-243, Pb-210, Ra-225, Ra-226, and Sn-126.
7-58
-------
• Collective doses to the critical organ, which is bone for most
of the important radionuclides, are of the order of 105 man
rems in the early stages due mainly to Sr-90, decreasing to
103 to 10^ man rems at longer times from a number of different
radionuclides.
• If plankton were to become established as a major direct food
source comparable with fish, then the predicted whole body
collective doses could be larger than those from consumption
of surface fish. The maximum annual value of collective whole
body dose is 2 x 10^ man rems after 100 years due to Sr-90.
• The maximum, annual whole body collective doses from consump-
tion of desalinated water are small.
• External collective doses from contaminated sediments are of
the order of 103 to 10^ man rems for both skin and whole body
in the early stages due to Sr-90 and Cs-137.
Comparison with Natural Levels of Activity & Levels Due to
Fallout
• As an attempt to provide a further yardstick against which to
compare the results of the calculations of water concentrations,
and therefore the consequent doses, Table 7-IX lists the levels
of natural and fallout activities for some of those nuclides
known to be present in seawater. The levels of the same
nuclides predicted by the modeling for the assumed input are
also given. It can be seen that in no case does the prediction
from the model exceed the natural level of the nuclide, and
that in most cases the model predictions are orders of magni-
tude lower. Even for those short-lived nuclides such as Ra-225
which do not occur to a significant extent in nature, the model
concentrations are less than the natural concentrations of any
of the radionuclides listed. The highest concentration of any
actinide predicted is comparable with the natural level of
Ra-226. Most fission products do not- occur in nature but are
present in seawater as a result of fallout from nuclear
weapons testing. The levels predicted by the models are
comparable with these fallout levels.
• These comparisons are not intended as a justification of the
introduction of high-level waste in the ocean, merely as an
indication that although the numerical results predicted for
individual or collective doses may appear high, they are con-
siderably less than the current doses from natural activity in
seawater would appear to be if calculated on the same basis.
7-59
-------
TABLE 7-IX
LEVELS OF NATURAL AND FALLOUT RADIONUCLIDES IN SEA WATER
Nuclide
Actinides:
Pb-210
Po-210
Ea-226
Th-230
Th-23U
U-23U
U-238
Pu-239
Fission
products:
H-3
Sr-90
1-129
Cs-137
Natural activity
in sea water
(jiCi cnr3)
(1-9) x1
-------
7.3.3.5 Operational & Transportation Risks. Seabed disposal
involves the loading and shipment of high-level radioactive waste by
sea to the emplacement site. Such shipments give rise to operational
and transportation risks such as the loss of a canister into the sea.
The potential radiological impacts arising from accidents during
operation and transport of high-level waste to seabed disposal sites
represent an integral part of the overall radiological impact
assessment of the seabed disposal concept.
This section will briefly summarize possible operational and
transportation accidents and risks from seabed disposal. The infor-
mation presented has been abstracted from Evaluating The Loss Of A
LWR Spent Fuel or Plutonium Shipping Package Into The Sea, Heaberlin
& Baker, BNWL-SA-5744.
A more detailed description of the radiological impacts of
transportation may be found in Final Environmental Statement On The
Transportation of Radioactive Material By Air And Other Modes, NUREG-
0170. Although this report addressed the environmental impacts re-
sulting from the transport of radioactive material by air, many of
the conclusions concerning transportation risks, particularly the
assumptions and methodologies used, may be applicable to seabed
disposal.
Pre-loss Conditions
0 Two initial states for the shipping packages were considered
prior to loss into the sea
7-61
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(1) An undamaged package assumed Co have its full design
integrity
(2) Package damaged by a shipboard fire
The fire environment associated with commercial freighters is
not well defined but data from Sandia indicates that fire
temperatures in hydrocarbon fires (the type of fire most likely
to occur) may reach averages of 1000°C. Other types of pre-loss
damage, such as a collision by two vessels in a harbor, have not
been considered.
• Since plutonium is not volatile and will not evolve as a gas
even at high temperatures, no distinction has been made between
a fire damaged and an undamaged package. An extended fire at
1000°C could, however, cause the canisters to rupture, but no
significant release of plutonium is expected.
• In the case of spent fuel casks, after approximately 4 hours
of high temperature fire, some fuel elements would begin to
fail. This may lead to unanticipated releases at the ship
fire.
Failure Mechanisms in the Sea
• Once the shipping package (damaged or undamaged) is lost into
the sea, two failure mechanisms may take place:
(1) hydrostatic pressure
(2) corrosion
• Since it was assumed in Section 7.3.1 that canisters would be
designed to withstand high pressures, only under the case of
a damaged canister will there be any potential for canister
collapse by hydrostatic pressure.
• Similarly, corrosion rates to canisters lost at sea will ex-
perience the same leakage rates as described in Section 4.2.
However, the canister does not have the sediment barrier to
protect against radionuclide migration. If the canister was
damaged by fire prior to loss at sea, then the corrosion
rates for canisters will increase.
Radiological Impact
• Radioactive materials released into the sea environment
would disperse into a large, volume of the ocean. Most of
the radionuclides such as cesium and plutonium will be
7-62
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reconcentrated through the food chain to fish and inverte-
brates which could be eaten by man. The dose to a man from
the consumption of fish, Crustacea, and molluscs is highly
dependent upon the concentration of radionuclides in the
individual fish consumed.
• Table 7-X gives the population and average individual doses
as the dose received over the period of intake and 50-year
dose commitment for the plutonium package loss.
• Table 7-XI gives the doses for loss of a spent fuel cask.
• Only in the most severe case, that of a spent fuel cask in
an extended fire, are the calculated radiation doses for the
average exposed individual as high as natural background.
All other cases had much smaller doses.
7.4 Economics
In Section 7.2, two basic emplacement techniques were described
in detail:
• Free fall penetration
• Controlled drilling from a surface ship.
In the free-fall penetration method, high-level waste canisters
of the types discussed in Section 7.3.1 would be dropped from a ship
through the water column. A terminal velocity of 70 miles/hr would
be reached at impact. This technique assumes that the medium for
emplacement would be soft deep-sea (clay) sediments. It is projected
from sample extraction experiments that these clays would be soft
enough to allow a canister to penetrate from 30 m and more. Clearly,
this method would be inappropriate if emplacement site surface
layers are to be composed of underlying bedrock. If bedrock is the
chosen medium, then the controlled drilling technique(s) from a
surface ship would need to be employed.
7-63
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TABLE 7-X
ESTIMATED DOSE AND DOSE COMMITMENT FROM MARINE FOOD
CHAIN FOR LOSS OF PLUTONIUM PACKAGE AT SEA*
Population Dose
(man-rem)
Average Individual
Dose (rem)
DOSE DURING 50-YEAR DOSE
INTAKE COMMITMENT
5.0 100
5.7 x 10~6 1.1 x 10~4
238 239 240
*2.55 kg Pu per package - 1.5 wt% Pu. 58 wt% Pu, 24 wt% Pu,
11 wt% 241Pu, 4.9 wt% 242Pu, 1.0 wt% 241Am, typical recycle
plutonium.
SOURCE: Evaluating The Loss of An LWR Spent Fuel or Plutonium
Package into The Sea, Heaberlin & Baker, BNWL-St-5744.
7-64
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TABLE 7-XI
ESTIMATED DOSE COMMITMENT FROM MARINE FOOD CHAIN FOR LOSS OF
A SPENT FUEL SHIPPING CASK CONTAINING 3.1 MT OF URANIUM
LOCATION OF
LOSS
Continental
Shelf
Deep Ocean
'INITIAL
CONDITION
Population
(man-rem)
Average Individual
(rem)
Population
(man-rem)
Average Individual
(rem)
SOURCE: Evaluating The Loss of An
UNDAMAGED
MINOR FIRE
INTERMEDIATE FIRE
510
5.9 x 10"4
<100
<1.1 x 10"4
LWR Spent Fuel or
EXTENDED
FIRE
1 x 105
0.11
<100
<1.1 x 10"4
Plutonium
Package into The Sea, Heaberlin & Baker, BNWL-St-5744.
7-65
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In SecCion 7.3.1, the sediment chosen as the barrier against
waste intrusion into the biosphere was soft deep-sea (clay) sediments.
The reasons are three-fold:
(1) Several studies (previously mentioned in Section* 7.3)
have indicated that deep-sea (clay) sediments will act as
effective barriers to radionuclide migration. Experi-
mentation on distribution coefficients and retardation
factors of radionuclides have been conducted for deep-sea
sediments.
(2) Drilling techniques in several types of bedrock will
create h'ole closure problems (see Section 7.3.2).
Development of suitable sealants has not yet begun.
(3) The drilling techniques have not been demonstrated.
Because the current policy is to dispose of high-level
wastes in land-based repositories, funds have not been
appropriated which would be adequate to test the accuracy
and effectiveness of the drilling concepts.
Because of these facts, free-fall penetration is soft deep-sea
sediments in the most likely form of emplacement to receive continued
funding at this time. Therefore, the economics of seabed disposal
will be presented using this concept as the base case (most likely
case) for cost estimates. Cost estimates will also be provided for
controlled drilling techniques, but these methods are less likely
to be implemented.
7.4.1 Cost Estimates
Cost estimates for the free-fall penetration and for controlled
drilling are given in Table 7-XII. As shown on Table 7-XII, the total
costs for controlled drilling are more thaja twice as much as that
shown for free-fall penetration.
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TABLE 7-XII
SUMMARY OF COST DATA FOR SEABED DISPOSAL*
REFERENCE PLANT CAPITAL COSTS** FREE-FALL CONTROLLED
($ MILLION) PENETRATION DRILLING
1. Port of Embarkation 20 20
2. Sea Transport Vessel 100 100
3. Sea Drilling Platform 0 300
4. Platform for Free Fall 50 0
5. Drill Pipe and Casing 0 5
6. Monitoring Equipment 3 3
7. Shipping Cask (300) 45 45
TOTAL CAPITAL COSTS 200 475
(rounded)
REFERENCE PLANT OPERATING COSTS***
($ MILLION/YR)
1. Port Operation 1 1
2. Sea Vessel Operation 8 8
3. Sea Platform Operation 5 8
(either drilling or free-fall)
4. Drilling and Support Maintenance 0 7
Operations
TOTAL OPERATING COSTS 14 24
*A11 costs are expressed in 1973 dollars.
**Capital costs are based on a 25-year plant lifetime, and a total
capacity for storage of 45,625 MTHM.
***Plant operating costs are based on emplacing 1,825 MTHM/yr.
SOURCE: High Level Radioactive Waste Alternatives. Section 6: Sea-
bed Disposal, BNWL-1900, Volume 3, May 1974.
Note: Cost estimates for free-fall penetration were changed
slightly by MITRE staff to be consistent with other dis-
cussions in the report.
7-67
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REFERENCES
1. David A. Deese, "Seabed Emplacement & Political Reality," Oceanus,
Volume 20, Number 1, 1977.
2. "Consultants' Meeting to Review the Radiological Basis of the
Agency's Provisional Definition and Recommendations for the
London Convention," International Atomic Energy Agency (IAEA),
June 1977, London, England.
3. "Release Pathways for Deep Seabed Disposal of Radioactive Wastes,"
Sandia Laboratories, IAEA-SM-198/34.
4. "Seabed Disposal Program - Annual Reports," Sandia Laboratories,
SAND 74-0410, SAND 76-0256, SAND 77-1270, 1974, 1976, 1977,
respectively.
5. Charles D. Hollister, "Seabed Disposal Option," Oceanus. Volume 20,
Number 1, 1977.
6. Armand J. Silva, "Physical Process in Deep-Sea Clays," Oceanus,
Volume 20, Number 1, 1977.
7. "High-Level Radioactive Waste Management Alternatives," Section 6,
Seabed Disposal, BNWL-1900, Volume 3.
8. "Alternatives for Managing Wastes from Reactors and Post-Fission
Operations in the LWR Fuel Cycle," Volume 4: Alternatives for
Waste Isolation and Disposal, ERDA-76-43, 1976.
9. G. Ross Heath, "Barriers to Radionuclide Waste Migration," Oceanus,
Volume 20, Number 1, Winter 1977.
10. P. D. Grimwood and G. A. M. Webb, "Assessment of the Radiological
Protection Aspects of Disposal of High-Level Waste on the Ocean
Floor," Natinal Radiological Protection Board, NRPB-R48, Oct. 1976.
11. "Technical Support for the Radiation Standards for High-Level Radio-
active Waste Management," Subtask C-2, Draft, Arthur D. Little, Inc.
12. "Technical Support for the Radiation Standards for High-Level Radio-
active Waste Management," Subtask C-l, Draft, Arthur D. Little, Inc.
13. "Technical Support for the Radiation Standards for High-Level Radio-
active Waste Management, Subtask C-3, Draft, Arthur D. Little, Inc.
14. S. W. Heaberlin and D. A. Baker, "Evaluating the Loss of a LWR Spent
Fuel or Plutonium Shipping Package into the Sea," BNWL-SA-5744,
Battelle, 1976.
7-68
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8.0 ICE SHEET DISPOSAL
Continental ice sheets have been considered an alternative
approach to the international solution for the final disposal of
high-level radioactive waste. Theoretically, they could provide
means for adequate geographic isolation of high-level radioactive
waste from man's environment. However, the feasibility of ice
sheets' long-term containment capailities is presently uncertain.
These uncertainties exist in areas that have recently been reviewed
by three international groups of glaciologists.1'2*3 T^16^ findings
concluded that before ice sheets could be considered for waste dis-
posal applications, certain areas of limited knowledge require
further investigation:
• the evolutionary processes in ice sheets
• the relationships of ice sheets with climatic changes
• the nature of future climatic changes on the stability
of the ice sheets
The following sections are a summary of the ice sheet disposal
concept reported in references 4>nd 5.
8.1 Descriptions of Ice Sheet Disposal Concepts
The ice sheet disposal alternative is considered in terms of the
feasibility of three concepts discussed in the literature. >^
Waste disposal by any of the three concepts, if established, would be
either in the Antarctica or Greenland ice sheets. A generalized
schematic of the waste management operational requirements is shown
in Figure 8-1. This schematic includes the basic system operations:
8-1
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CD
I
N>
REPROCESSING
PIJVNT
SHIELDED CELL
PORT J^ACILITY SHIPPING V
CASK * -
TRANSPORT SHIP
CANISTER IN
TRANSPORT
CASK
DRILLING
RIG
Source: High-level Radioactive Waste Management Alternatives,
BNWL-1900, Volume 3, Section 5, Ice Sheet Disposal,
Richland, WA, May 1974.
FIGURE 8-1
SCHEMATIC OF OPERATIONS IN ICE SHEET DISPOSAL SYSTEMS
FOR HIGH-LEVEL RADIOACTIVE WASTES
-------
• transportation of solidified waste from the reprocessing
plant or interim retrievable surface storage facility by
truck, rail, or barge to embarkation ports;
• marine transport by specially designed ships during one-
to-three-month periods of each year, with ice-breaker
escorts near t)ie ice sheets;
• a debarkation facility for unloading the waste canisters
near the edge of the continent;
• the use of surface vehicles or aircraft for over-ice
transport on a year-round basis;
• unloading and emplacing the waste canisters at the
disposal site.
Ice sheet disposal of high-level radioactive waste would be done
using one of the three concepts described in the following sections.
8.1.1 Meltdown or Free Flow Concept
The meltdown or free flow concept is shown in Figure 8-2.^ In
this concept, waste disposal is accomplished by selecting a suitable
location in the ice sheets, predrilling a shallow hole, and eventu-
ally lowering the canister into the hole where it is allowed to melt
down or free flow to the ice sheet basal.
Surface holes, predrilled to depths from 50 to 100 meters, serve
as protective shielding from radiation during the initial operation
phase of canister emplacement. To avoid individual canisters
interfering with each other during descent and possible concentration
at the ice sheet basal, it has been suggested that a spacing of about
one kilometer apart will be required. Figure 8-2 shows a schematic
of the meltdown or free flow concept.
8-3
-------
oo
I
HEAT
1!
DRILLING
RIG
SURFACE ANCHORS
AND SITE MARKERS
UP TO
4000 METERS
1
If
ICE SURFACE
MELT
WATER
MELT DOWN
ANCHORED
EMPLACEMENT
ICE
EXTENDED LEGS
SURFACE
FACILITY
BEDROCK
^^S^^T?5^^ '
SOURCE: U.S. Energy Research and Development Administration,
AJternatives for Managing Wastes from Reactors and
Post-Fission Operations in the LWR Fue-1 Cycle,
Volume 4 of 5, "Alternatives for Waste Isolation
and Disposal," ERDA 76-43, Washington, D.C., May 1976.
FIGURE 8-2
ICE SHEET DISPOSAL CONCEPTS
-------
The canister meltdown rate is based on calculations from the
penetration rates of thermal ice probes. It is estimated that the
rate of descent for each canister would be on the order of 1.0 to 1.5
meters per day." Assuming only vertical movement and an ice sheet
3000 meters thick, a period of 5 to 10 years is required for meltdown
to the bedrock.
Another important factor in this concept is the design and shape
of the canister. Adequate design and shape is important to assure a
vertical path from surface to bedrock. In addition to the canister
design and shape, the type of construction materials is important.
Considerations for these materials should meet requirements for dif-
ferences in ice sheet pressure and the possibility of saline water
present at the basal.
There are also other options to this straightforward meltdown
concept. Some appear more attractive from some viewpoints than
others. For example, if the canister were so designed such that the
contained waste and its the density was intermediate between those of
water and ice, the rate of descent could be greatly decreased. The
melt-down time would then approach that of the residence time of ice
particles and by that time the canister would have become thermally
inert.
8.1.2 Anchored Emplacement Concept
The anchored emplacement concept requires similar technology to
the meltdown or free flow concept described above, the difference
8-5
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being that this concept allows for interim retrieval of the waste.
Canister emplacement is accomplished also by drilling a hole in the
ice sheet at a depth from 50 to 100 meters; cables 200 to 500 meters
are attached to the canister before lowering it into the ice sheet.
After meltdown, the canister is anchored at a depth of 200 to 500
meters by the anchor plates on or near the surface. The advantage of
this concept, over the melt-down or free-flow concept, is that
Instrument leads attached to the lead cable could be used to monitor
the condition of the canister during descent and emplacement. A
period of 6 to 18 months is required for emplacement based on calcula-
tions from thermal ice probe rates.
Following emplacement, new snow and ice accumulating on the
surface could eventually cover the anchor markers and present diffi-
/
culties for their future recovery. The average height of snow and
ice accumulating In the Antarctica and Greenland is about 5 to 10
cm/year and 20 cm/year, respectively. Future'recovery of canisters
for periods up to 200 to 400 years may be possible by using 20-meter
high anchor markers. The approximate time for the entire system to
reach bedrock at a typical site is estimated to be 30,000 years.
During that time, the canisters and anchors would tend to follow the
flow pattern of the ice.
8.1.3 Surface Storage Facility Concept
This concept requires the use of large surface storage units
constructed above the snow surface. The facilities will be supported
8-6
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by jack-up pilings or piers resting on load-bearing plates. Waste
disposal would be accomplished by initially placing the waste canis-
ters in cubicles inside the facility. Cooling of the canisters would
be by air from natural draft. Elevation of the facility above the
ice surface for as long as possible would provide for reduced snow
drifting and heat dissipation. During this period the waste canis-
ters would be retrievable. However, when the limit of-the jack-up
pilings is reached, the entire facility would act as a heat source
and begin to meltdown through the ice sheet. It is estimated that
such a facility could be maintained above the ice for a maximum of
400 years after construction.^
8.2 Status of Ice Sheet Technology Development
Current technology appears adequate for waste canister emplace-
ment using the concepts previously described. Some uncertainties /
still exist in the technology and additional research is required.
Further evaluation of transportation, logistics, and support facili-
ties is needed to determine the feasibility of the technology.
Improved means of inland transport of the waste over difficult and
hazardous inland routes, and development of an efficient transporta-
tion system to carry the 20- to 25-ton casks require further evalua-
tion. Areas of specific concern to a transportation system are fuel
depots along the route and the means of fuel supplies.
8.2.1 Emplacement
Because the meltdown and anchored emplacement concepts are
self-emplacing, little developmental research is needed for actual
8-7
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operation after the wastes arrive on site. Predrilled holes of 50 to
100 meters depth would be needed for initial emplacement. At this
depth, the interconnecting air spaces in the ice have been sealed off
into bubbles. Experimental holes up to 400 meters below the ice
surface have been drilled using existing drilling equipment. These
holes were "dry bored" and compressed air/served as the drilling
fluid.
Because the surface storage concept would not require drilling,
emplacement of the waste canisters would be accomplished by surface
handling equipment on site. There is currently equipment available
to handle casks without difficulty.
8.2.2 Transportation
Waste transportation from the embarkation ports to the areas of
the disposal sites would be very difficult but not impossible. The
ports would be designed for maximum safety, utilization, and accept-
ability. Consideration of docking facilities for large ships would
be considered during dock design. The transport ships considered
would be modifications of existing vessels. The ships would be
equipped with the necessary safety features during construction.
Current crude oil tankers are being built in the 400,000-dead weight
ton class. Tankers of this capacity are larger than the ships re-
quired to transport the annual waste generated by a 5 MT/day repro-
cessing plant.
Although transportation appears adequate for transport from the
embarkation port to the ice sheet margin, inland transport to the
8-8
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disposal site does present problems. These problems include slow
travel, severe weather conditions, refueling, equipment maintenance,
etc. Inland transportation would be necessary for the within-ice
sheet concept to reach the most suitable location to gain access to
areas of maximum thickness, stability, and as much isolation as
possible. The distance inland that must be traveled (e.g., in the
Antarctica) could be on the order of 1000 kilometers (600 miles).*
Unloading of casks at the continent margin would probably be
done by crane or helicopter. Inland transport from this point could
be accomplished by several methods. The reference study considered
the use of surface sleds pulled by tracked vehicles, but this method
has been abandoned by the U.S. in favor of aircraft as used to supply
its permanent stations in the Antarctica. The average speed of the
surface tracked vehicles is 3 to 6 kilometers per hour (2 to 4 mph),
and considering trips of 1000 kilometers (600 miles) would require
about 2 weeks travel per roundtrip.
Aircraft have been considered for inland transports, however,
the use of aircraft is subject to limitations. Aircraft carrying
payloads of up to 10 tons have been successfully used for transport-
ing both personnel and supplies to Antarctica. Their use would
involve high fuel consumption, probability of aircraft accidents,
difficulty of navigation in severe weather conditions, and would
require relatively drift-free landing areas at all times.
The final mode of inland transportation considered is Surface
Effects Vehicles (SEV) such as hovercraft. SEV could be a possible
8-9
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means of transport, although they have not been tested in the high
elevations of the Antarctica ice sheets (e.g., typical elevations are
460 meters at 16 kilometers, 1800 meters at 160 kilometers, and 2400
meters at 320 kilometers).^
The use of any type of surface vehicle to transport waste inland
would require the establishment of a chain of fuel depots. Resupply-
ing depots would probably be done using aircraft drop-offs. In this
study, the conservation of fuel is considered a key item for any mode
of shipment in the Antarctica.
8.3 Environmental Considerations
During several periods of the Pleistocene geologic epoch
(approximately the last 2 to 3 million years), ice sheets covered
about 30 percent of the earth's land mass. Only the ice sheets of
the Antarctica and Greenland exist today which, together, cover about
11 percent of the earth's'land mass. Together these two ice masses
constitute the world's largest reservoir of fresh water (approxi-
mately 78 percent of the world's nonoceanic water).
8.3.1 Availability of Ice Sheet Data and Uncertainties
No information is presented in the literature that precludes the
technical feasibility of high-level radioactive waste disposal in the
continental ice sheets. The requirements for all waste management
systems (i.e., transportation, logistics support, and emplacement)
are available or could be made available through existing technology.
However, the limitations of today's knowledge of the physics and
history of ice sheets make the prediction of ice sheets stability
8-10
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uncertain for periods greater than a few thousand years. Verifica-
tions of theories that support ice sheet disposal would require many
years of extensive new data collection and evaluation.
With regard to the limited data presently available, ice sheet
disposal concept could offer potentially favorable features:
o geographic isolation
o relative isolation and containment of wastes by the ice
in the event of leakage or canister failure
o low temperatures and high heat dissipation capacity
o relative safety from damage by storms, sabotage, and
other hazards once the waste is emplaced
There are potentially unfavorable features for ice sheet dis-
posal in general:
o extensive new data on all facets of ice sheet physics
will have to be obtained
o the harsh environment and unpredictability of conditions on
on ice sheets will present severe problems in establishing
safe operations
o ice sheet areas are inaccessible during much of the year
(8 to 11 months) because of storms, long periods of winter
darkness, and freezing of surrounding seas
o monitoring and evaluating waste disposal operations would
be difficult
o recovery from an unforeseen occurrence during transport to
the disposal site would be difficult
8.3.2 Long-Term Containment
The capability of ice sheet to contain radioactive waste for
long periods of time is presently speculative. Containment is highly
dependent on the stability and physical properties of the ice sheet.
8-11
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An analysis of the potential of canister failure upon emplace-
ment in the ice sheets has been considered for the three disposal
concepts. Providing that a canister failure should occur, the radio-
active material contained would be in a potentially mobile system
(i.e., the ice and water that may be present beneath it, either
naturally or melted by the waste canisters). The probability of the
waste eventually reaching man's environment, while in a hazardous
form, depends greatly upon several factors of the system:
o rates of motion within the ice sheet
o the physical state and rates of ice flow
o movement of meltwater at the base of the ice sheet
o the long-term stability of the total ice sheet
8.3.2.1 Motions of Ice Sheets. Over the past few years,
several measurements have been made to measure the surface motion
rates of glacial ice.° Basically, these measurements have been
done in the valley glaciers, ice shelves, and marginal areas of the
ice sheets. The results of the measurements indicate a variation
from centimeters per day to kilometers per year. Although mathema-
tical models and theoretical studies have been made, the interior
rates of ice sheet movement are essentially unknown.
8.3.2.2 Physical State and Rates of Ice Flow. Until recently
the physical conditions at the base of the ice sheets were essen-
tially unknown. Theoretically, some investigators suggest that in
the central areas of the ice sheet which are sufficiently thick,
melting could be occurring as the ice sheet moves as a rigid block
8-12
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sliding over underlying land, creating a bottom melting condi-
tion.9-12
Three general types of ice flow patterns are identified:
o sheet flow—general outward movement of ice over a bed of
low relief
o stream flow—relatively rapid movement of valley glaciers
and ice streams
o ice shelf movement—general seaward movement of an ice
shelf
The velocities of ice surface measured at a number of locations
in various parts of the Antarctica are as follows:
o sheet flow—0.05 to 0.15 meters (2 to 6 inches) per day
o stream flow—0.3 to 2.6 meters (1 to 9 feet) per day
o ice shelf movement—0.9 to 1.2 meters (3 to 4 feet) per day
In Greenland, "measurements of ice surface velocities are gener-
ally lower for sheet flow—as low as 0.1 centimeters (0.04 inch) per
day, and as high as 27 meters (88 feet) per day for outflow gla-
ciers."4
8.3.2.3 Meltwater at Base of Ice Sheet. Within the past few
years, meltwater presence at the base of the ice sheet has been
detected.13 The dimensions of an ice sheet and its movement over
the underlying material are controlled to some extent by the water
layer. Measurements have been limited to a few bore holes which
penetrated the bedrock. Here, meltwater detection has been done
using remote-sensing techniques. It is known that water layers and
under-ice lakes exist beneath parts of the Antarctica ice sheet. But
8-13
-------
the effect of its presence on ice motion theories and ice sheet sta-
bility has not been determined. However, various sources and methods
have been proposed to account for the presence of a water layer at
the ice sheet basal. These sources could be related several factors:
o the geothermal heat flux that may raise the temperature to
the melting point of the ice
o frictional heat caused by the motion of the ice over the
underlying rock may melt some of the ice
o various combinations of geothermal heat flux and frictional
heat may occur
The temperature of the water found at the ice-rock interface in a
core hole drilled through the Antarctica at Byrd Station was esti-
mated to be -1.6°C. Evidence found there indicates that the bottom
surface of the ice was at the pressure melting point. Based on cal-
culations, the water layer present was estimated to be at least 1
millimeter in thickness. A similar analysis was performed at Camp
Century on the Greenland ice sheet. The water temperature found at
the bottom of a hole drilled 1,375 meters was -13.0°C, which was well
below the pressure melting point. Meltwater at the base of the ice
sheet has been proposed as the cause of initiating the (East)
Antarctica surges which were considered to initiate the northern
hemisphere glaciations.
8.3.2.4 Long-Term Stability. The stability of the ice sheet
for long-term containment is essential for waste disposal methods
requiring waste isolation for periods of time of a few thousand years
or longer. This, in turn, depends greatly on future snow accumula-
8-14
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tion rates compared Co ice losses by melting, evaporation, formation
of icebergs, and future world climatic changes. Present scientific
opinion suggests that the Antarctica ice cap is growing or at least
is stable. However, the future of its stability cannot be predicted
from scientific interpretation of past climatic conditions from the
available ice core. It is possible that the occurrence of manmade or
natural climatic changes could affect the long-term stability of the
ice sheets. The magnitude of such abrupt changes that might occur is
presently unknown.
8.3.3 Characteristics of Waste Forms
The reference study considered only solidified waste forms such
as borosilicate glass encapsulated in metal canisters. It may be
stored in the interim for 5 to 10 years to allow some thermal decay,
but will not need any further conditioning for disposal. At this
age, each canister of waste will contain about 1 megacurie of- radio-
active material of a heat generating rate of about 3 kilowatts. This
amount of heat generation is capable of raising the temperature of
the waste to its melting point unless external cooling is provided.
Adequate cooling of the casks would be necessary until the waste
reaches the ice sheet disposal areas. At the disposal areas, the
average ambient temperature is below 0°C and should provide adequate
cooling. Upon emplacement of the canister in the ice sheet, an
initial melt pool of about 70 meters in diameter will result. The
hole will reseal because of the temperature of the surrounding ice
and its plasticity.
8-15
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8.3.4 Site Requirements
Site requirements will vary depending upon which disposal con-
cept is selected. Requirements for the melt-down concept would
require a location where the ice has the greatest thickness and sta-
bility. Such location would be as far from the coast as possible to
assure maximum containment. Some investigators suggest that the best
location for the melt-down concept would be near the top of buried
ridges in the underlying bedrock where the ice thickness is thought
to be one kilometer.^ Here, the ice-rock interface temperature is
considered lower than basin areas and lateral ice movement is
minimal.
8.3.5 Radiological Risks
Only hypothetical dose calculations have been made for radionu-
clides released from an ice sheet disposal site into the ocean off
the coast of Greenland. Based on assumptions that a failure occurs
in the disposal system, the release of radionuclides into Greenland
current of 8 x 10° nrVsec would be 0.3 percent per year of the
total inventory available and complete mixing would occur in the
ocean rapidly. Human pathways are assumed to be mostly via fish
consumption. The maximum dose was considered to be from an indivi-
dual consuming 100 kg/yr of fish caught in these contaminated waters
and is estimated to be 0.2 rarem/yr. (Also refers to Section 7.0 for
discussion of radioactive releases to the ocean.)
8.3.6 Accidental Risks and Consequences
The major accidental risks would be associated with transport at
8-16
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sea. In the event that a ship is sunk, the waste canisters could be
equipped with flotation and other devices for recovery, as shown in
Figure 8-3. This figure shows a typical recovery of a sunken cask at
sea. When the cask sinks, a collar is activated which triggers the
flotation device to raise the cask back to the sea surface. Reloca-
tion of the lost cask is done by radio signals given off after the
cask reaches the surface.
The incident of a ship crashing into an ice pack and sinking
could cause severe problems for canister recovery. During transport,
the canisters would be enclosed in casks to prevent radiation and
high temperature effects on the surrounding environment. Transport
of waste is also discussed in Section 7.
8.3.7 Additional Data Requirements
Additional R&D requirements for ice sheet disposal are discussed
from two perspectives: those related to obtaining basic information
on ice sheets, and those related to the handling, transportation, and
emplacement of the waste. Further studies are needed to adequately
interpret the parts of the ice sheets, where the greatest thickness
occurs. Ice motion measurements are also significant in predicting
ice sheet long-term stability. Several measurements of surface
motion have been made for parts of the surfaces of valley and outlet
glaciers. Measurement of the interior motion is hindered by the lack
of fixed landmarks. In order to obtain more accurate surface motion
measurements, a minimum of 5 to 10 years of R&D would be necessary to
provide meaningful data on the gross motion of ice sheets.
8-17
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RADIO BEACON
RADIATION DOWNWARD
..FLOTATION DEVICE
ACTIVATED
FLOTATION DEVICE
UNACTIVATED
LOWER BALLAST PORTION
Source: High-level Radioactive Waste Management Alternatives,
BNWL-1900, Volume 3, Section 5, Ice Sheet Disposal,
Richland, WA, May 1974.
FIGURE 8-3
POTENTIAL CASK-CANISTER RECOVERY SYSTEM AT SEA
8-18
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The stability of ice sheets (whether they Will continuously
exist in the future or whether they are expanding or shrinking) is
presently unknown. To assure waste isolation for periods of hundreds
of thousands of years, the present trend in the balance must be known
to estimate future climatic conditions.
The estimated time required for expanded R&D programs to lead to
the establishment of a commercial system for waste canister disposal
in ice sheet is summarized in Figure 8-4. It is estimated that about
5 to 10 years would be required to select one of the three disposal
concepts discussed after the program has been initiated. The minimum
time required for the entire program is estimated to be 25 years to
adequately evaluate ice sheets, in general, and conduct detailed
studies necessary for specific site evaluations.
8.3.8 Summary
The ice sheet disposal concepts (assuming that operations are
carried out as visualized) should have negligible environmental
impact. The exception may be the potential impact on the ice sheet
itself. Presently, it is difficult to assess the effects that waste
canisters would have on ice sheets and of the interface conditions on
the waste canisters until the physical conditions within the ice
sheets and the ice-bedrock interface are better defined. In the
meltdown and anchored-emplacement concepts, waste isolation from the
environment can be assured as long as melting at the bottom of the
»•
ice sheets does not occur. The impacts on land, water, air, ecology,
and aesthetics will be considered.
8-19
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REQUIRED RESEARCH AND DEVELOPMENT TASKS
Tee Sheet Geophysical Studies •*••
Site Evaluation
Laboratory Studies
Transport Design and Construction
Embarkation Port Design and Construction
Concept Demonstration •••
Pi lot- Scale Demonstration
Esti.nated Years Required After Start of Program
0 1 5 1 10 1 15 | 20 | 25
^Si /oV_. A
ill \£/ — —/A—
('*> .A
\±/ ' •~r-l\""
(^ A
\^/ /H
W j&
I-J
o
Key Milestones
1 Initial Data Developed - Tentative Site Selections
2 Decision on Disposal Concept - Final Site Selection
3 Start Routine Waste Disposal
Source: Modifications of High-level Radioactive,Waste Management Alternatives,BNWL-1900,
Volume-3, Section 5, Ice Sheet Disposal, Richland, WA, May 197A.
FIGURE 8-4
OVERALL RESEARCH AND DEVELOPMENT SCHEDULE -
WASTE DISPOSAL IN ICE SHEET
-------
Some land impacts would probably be experienced in connection
with the embarkation port facility. An area of about one square
kilometer would be required for the radioactive handling shielded
cell and the loading dock facilities. The port facility would be
equipped with its own separate water, power, and sewer systems to
assure maximum safety.
The over-ice transport routes include an area at the edge of the
ice sheet, ice shelf edge, and ice-free areas on land for unloading
the shipping casks. Approximately six support and fueling stations
will be required along the transport route to the disposal area. An
additional 11,000 square kilometer area would be required for dis-
posal of the output from a reference reprocessing plant of 5 MT/day.
Other possible land impacts considered in the reference study
include accidental spills of fuel and the probability of fuel blad-
ders rupturing during drop-offs. Rupture of the fuel bladders is
considered to be a high risk because the fuel is capable of penetrat-
ing the snow and would reach the underlying ice where it will remain
until evaporated or eventually becomes buried by additional snow.
Accidental spills could reach the ocean if the incident occurred
near the edge of the ice sheet. Few, if any other impacts on water
are expected, except for a marginal increase in temperature of the
water used for once-through cooling of canisters during sea trans-
port. The only other water uses would be for consumption by the 200
operating personnel, which would be obtained by melting the ice.
8-21
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Air impacts would result from the combustion products of over-
ice transport vehicles, support aircraft, and fuel consumed for heat-
ing the facilities at the disposal site. At present, the effects of
these products are not considered a major problem. However, the
accumulation of exhaust fumes and vapors over a long period of time
may lead to temperature inversion and affect the weather pattern over
the ice sheets. Altered weather patterns could conceivably influence
the stability of the ice sheets.
Few, if any, ecological impacts are expected because the plant
and animal life are confined mostly to the coastal areas. The con-
struction of access routes and air traffic lanes could be done to
avoid as much as possible the feeding, nesting, and mating areas "of
the birds and animals that inhabit the coastal areas.
Aesthetic impacts would be nil due to the remoteness of the area
and lack of permanent residence population.
8.4 Capital and Operating Costs
The estimated capital and operating costs (1973 dollars) for the
three ice sheet disposal concepts are summarized in Table 8-1.^
Capital costs are primarily associated with transportation vehicles
and equipment and are essentially the same for all three disposal
concepts.
For the meltdown and anchored emplacement concepts, capital
costs are estimated to be about $410 million to handle the waste from
one reference fuel reprocessing plant. The associated operating
8-22
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costs are estimated at between $27 and $46 million per year. Capital
costs for the surface storage concept are estimated to,be about $415
million, with associated operating costs of about $23 million per
year.
The total system uflit charges are estimated to range between
$19,800/MT for surface storage disposal and $23,500/MT for anchored
emplacement (1973 dollars).^ These charges include: Reprocessing,
5-year interim liquid storage, solidification and containerization,
5-year interim solid storage, transport"to the disposal site, and
final emplacement.
8.5 Policy and Treaty Agreement
Because of treaty agreements, although the concept could be made
feasible through further R&D, ice sheet disposal of radioactive waste
is prohibited in Antarctica. However, Greenland (which is Danish
territory) can be excluded from these restrictions.
8-23
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TABLE 8-1
CAPITAL AND OPERATING COST ITEMS FOR ICE SHEET DISPOSAL
Capital costs, Million Dollars, for Meltdown and Anchored Emplacement:
1. Construction of Embarkation Port Facility 20.0
2. Sea Transport Vessel, Including Fully Equipped Hot Cell,
40-Ton Bridge Crane, etc.
3. Two Ice Breakers @ #60 x 10
4. Over-ice Transport Vehicles
5. Drilling Rigs
6. Monitoring Equipment
'7. Shipping Casks
8. Aircraft
9. Support Maintenance and In-Transit Facilities
Total Capital Costs
•Capital Costs, Million Dollars, for Surface Storage Facility:
1. Construction of Embarkation Port Facility 20.0
2. Sea Transport Vessel, Including Fully Equipped Hot Cell,.
40-Ton Bridge Crane, etc.
3. Two Ice Breakers @ #60 x 10
4. Over-ice Transport Vehicles
5. Surface Facility
6. Monitoring Equipment
7. Shipping Casks
8. Aircraft
9. Support Maintenance and In-Transit Facilities
Total Capital C9Sts
Operating Costs, Per Year, Million Dollars:
Meltdown or Free Flow Concept
1. Operation of Embarkation Facility
2. Operation of Surface Facility with Hot Cell
3. Transport Vehicles Operation
4. Drilling Operations and In-Transit Facilities Operation
Total Operating Cost Per Year
8-24
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TABLE 8-1 (Concluded)
CAPITAL AND OPERATING COST ITEMS FOR ICE SHEET DISPOSAL
Anchored Emplacement
1. Operation of Embarkation Facility 1.0
2. Operation of Surface Facility with Hot Cell 8.1
3. Transport Vehicle Operation 10.5
4. Surface Anchors, Cables, Chains 15.0
5. Drilling Operations and In-Transit facilities Operation 7.0
Total Operating Costs Per Year 41.6
Surface Storage Facility
1. Operation of Embarkation Facility
2. Operation of Surface Facility with Hot Cell
3. Transport Vehicles Operation
4. Maintenance and In-Transit Facilities Operation
Total Operating Costs Per Year
(Cost in 1973 dollars)
SOURCE: High-level Radioactive Waste Management Alternatives, BNWL-1900,
Volume 3, Section 5, Ice Sheet Disposal, Richland, WA, May 1974.
8-25
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REFERENCES'
1. B. Philberth, "Disposal of Atomic Fission Products in Polar Ice
Caps," IAHS Symposium, 1958.
2. E. J. Zeller, D. R. Saunders, and E. E. Angino, "A Suggestion
for a Permanent Polar High-Level Radioactive Waste Repository,"
Bull. At. Scientists, pp. 4-9 and 50-52, January 1973 (or
Reference 1, Appendix 5.A.).
3. K. Philberth, "On the Temperature Response in Ice Sheet to
Radioactive Waste Deposits," Presented at International Symposium
on the Thermal Regime of Glaciers and Ice Sheets, Simon Fraser
University, Burnaby, British Columbia, April 1975.
4. High-Level Radioactive Waste Management Alternatives, BNWL-1900,
Battelle Northwest, Richland, WA, Vols. 1 and 3, May 1974.
5. U.S. Energy Research and Development Administration, Alterna-
tives for Managing Wastes from Reactors and Post-Fission
Operations in the LWR Fuel Cycle, Volume 4 of 5, "Alternatives
for Waste Isolation and Disposal," ERDA 76-43, Washington, D.C.,
May 1976.
6. W. C. Haldor Aamot, The Philberth Probe for Investigating Polar
Ice Caps, AD-661 049, U.S. Army Cold Regions Research and
Engineering Laboratory, Special Report 119, September 1967.
7. M. G. Gross, Oceanography, Merrill Publishing Co., Columbus, OH,
p. 3, 1971.
8. A. J. Grow, "Results of Measurements in the 309 Meter Bore Hole
at Byrd Station, Antarctica," J of Glaciology v ^ no 36, pp. 771-
784, October 1963.
9. J. F. Nye, "The Motion of Ice Sheets and Glaciers," J. Glaciology,
Vol. 3, pp. 493-507, 1959.
10. J. Weertmen, "Stability of Ice Age Ice Sheets," J. Geophys. Res.,
Vol. 66, pp. 3783-3792, 1961.
11. T. Hughes, "Convection in the Antarctic Ice Sheet Leading to a
Surge of the Ice Sheet and Possibly to a New Ice Age," Science,
Vol. 170, pp. 630-633, 1970.
8-26
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12. W. F. Budd, "The Dynamics of Ice Masses," Australian National
Antarctic Research Expeditions, AWARE Scientific Reports,
Series A(IV) Glaciology, Pub. No. 108, Antarctic Division,
Department of Supply, Melbourne, Australia, 1969.
13. A. J. Gow, et al., "Antarctic Ice Sheet: Preliminary Results of
First Core Hole to Bedrock," Science, Vol. 161, pp. 1011-1013,
1968.
8-27
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9.0 CONTINENTAL GEOLOGIC WASTE DISPOSAL
Continental geologic disposal refers to those waste disposal
methods related to interment of the waste in deep geologic forma-
tions on the continents. The deep-mined geological repository is, of
course, included,in this category, but is extensively discussed in
other documents.* Although the deep-mined geologic repository is
the most advanced and most studied concept, many alternative conti-
nental geologic disposal methods have been considered. While these
alternative concepts may offer some advantages to deep-mined reposi-
tories in the form of engineering approach or economy, they also have
a commonality of problems related to the assurance of the isolation of
the waste from the environment. The containment problems, as dis-
cussed in Section 9.2, are sufficiently siiniliar that it might well be
concluded that if the problems of deep-mined geologic repositories
cannot be resolved, they are unlikely to be resolved for alternative
geologic disposal methods. The exception to this might lie in the
ultra-deep disposal methods where the greater depth of waste emplace-
ment could provide an additional time barrier to transport into the
environment. Technology development and cost will, however, be fac-
tors in the feasibility of such concepts. This section of the report
presents the alternative disposal concepts in the following manner:
• Concept Description - A discussion of the engineering
concepts
• Environmental Considerations - The geologic, hydrologic, and
climatic considerations, and the pathways to the environment
9-1
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• Technical Feasibility - A summary of the feasibility of
the alternative concepts
9.1 Concept Description
The alternative concepts considered include the following:
• Solution-Mined Cavities
• Waste Disposal in a Matrix of Drilled Holes
• Waste Disposal in Super-Deep Holes
• Deep-Well Injection
• Hydrofracture
• Rock-Melting Concepts
Most of the alternative disposal concepts require the waste to be
received in solid form. For a few of these concepts interim cooling
may be required prior to final disposal. Figure 9-1 presents a basic
flow diagram for solid waste disposal. If there is interim cooling,
the steam and other off-gases to the condenser are passed through hijh
efficiency filters in prder to trap any radionuclides which may have
escaped. A flow diagram for the liquid disposal process is shown in
Figure 9-2. It must be noted that because of the serious problems
that an accident in transporting high level liquid wastes would cre-
ate, these concepts would most likely require that the reprocessing
plant be at the repository site. The disposal of high-level and
transuranic liquid waste is generally considered unacceptable due to
the safety and containment problems involved. It may, however, be an
acceptable method for lew-level waste.
9-2
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Additional
Potential
Waste
Waste
Receiving
Facility
Waste to
Hole/Cavity
Uncondensable
Gases to Atmosphere
Condenser +
Treatment
Solidified
High Level
Paper, Cloth, Plastic
Wood, Rubber, etc.
Compactor
I
Fuel Clad
J
Steam and
Gases to
Surface
Disposal
Region
FIGURE 9-1
FLOW DIAGRAM FOR EMPLACEMENT OF SOLIDIFIED WASTE
-------
Uncondensable
Gases to Atmosphere
Reprocessing
Plant
i i
U Pu
Waste to
Hole/Cavity
1
I
1
.
r
1
1
i_
Nitric Acid Condenser +
Water Treatment
Steam +
Gases to
„. . , , Surface
t High Level
i
, Papei, Cloth, Plastic, 1 iviA«nlv<»r ~ »
^ Wood, Rubber, etc. | | I
1
1
Low Level and |
Intermediate Level-
Disposal
Region
FIGURE 9-2
FLOW DIAGRAM FOR EMPLACEMENT OF LIQUID WASTE
-------
A brief description of each alternative concept follows. This
will include method of emplacement, type of host rock which can be
used, waste form, sealing from man's environment, depth of emplace-
ment, and technical feasibility.
The Battelle, pacific Northwest Laboratories report, ERDA 76-43,
was a primary source of information contained in the following
sections.
9.1.1 Solution-Mined Cavities
Salt is the only rock type in which solution-mining techniques
can and are being used to construct large caverns. The current usage
is mainly for storage of petroleum products. The technique consists
of washing out the salt by fresh water action. The size and shape of
the cavern can be controlled through manipulation of the fresh water
flow, position of the inlet, location of the brine outflow pipe, the
inert blanket, etc. The cavern can be constructed in salt which is in
a dome, bedded, or anticlinal structure.2 The technology for this
concept is available now and would entail only surface facilities.
However, such a disposal concept may have serious limitations.
The limitations may result from the type of emplacement itself.
In this concept, the waste is received from the reprocessing plant in
a solid form and, upon arrival, is unloaded from the shipment casks by
remotely operated equipment. The waste is then moved into hot cells
for inspection, monitoring, decontamination, repair (patching over-
packing), and, finally, still using remote equipment, loaded into the
9-5
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hoisting facility. After tihe canister is placed in the hoisting
device, it is lowered into the cavern (300 - 3,000 meters-below sur-
face) until it is near the bottom; it is then allowed to fall onto a
random pile of canisters. Figure 9-3 shows a generalized concept of a
solution-mined storage facility. The random placement of canisters
presents a problem if there are high-heat generating materials within
the canisters. The salt host rock may not be able to dissipate the
heat away quickly enough to prevent melting and consequent flow of
salt. This concept is therefore limited to handling only low-heat
generating transuranic wastes. There are additional problems. Little
is known about the stability of the caverns once they are dried out.
There are also questions on the optimal size and shape-of caverns to
assure the greatest stability as well as the best drying method to be
used. There is also the question of retrievability. "Fishing" by
grapple for canisters is not a demonstrated retrieval method and dis-
posal of high-gamma transuranics in the cavern and uncertain cavern
stability would preclude direct .loading of canisters onto the hoist by
men lowered into the cavern.
9.1.2 Waste Disposal in a Matrix of Drilled Holes
In this concept, a matrix of holes about 1 meter in diameter
would be drilled into a thick, tight geologic formation with no
cracks, fractures, faults, etc., to permit water to circulate. These
holes would be drilled to a depth of 30t) to 6000 meters. Salt domes,
bedded salt, argillaceous, intrusive igneous, and metamorphic
9-6
-------
OPERATIONS ANO" •.*'•••'
OFFICE 3UILOINGS
Source: Battele, Pacific Northwest Laboratories, Reference 2.
FIGURE 9-3
GENERALIZED CONCEPT SOLUTION
MINING FINAL STORAGE FACILITY
9-7
-------
formations are examples of the geologic candidates for host rock in
this concept. Solidified (borosilicate glass) waste would be received
and prepared for disposal. It would then be placed on a combination
transporter-hoist vehicle which would move it to the hole and lower it
into position. After the hole has received its maximum amount of
canisters, it is backfilled and sealed. A generalized concept of such
a facility is shown in Figure 9-4. This concept, like most of the
concepts described here, features only surface facilities.
It is assumed that a thick (1000 - 3000m), hydrologically tight,
stable formation can be found. The spacing of the holes and of the
canister within the holes would have to be designed so that heat can
be dissipated without melting.
The problems with this concept lie basically in the many penetra-
ting boreholes which connect the disposal zone with man's environment.
It is feared that these boreholes would increase the probability that
the integrity of the containment provided by the geological formation
could be compromised, with the result that it would be difficult to
satisfy the long-term containment requirements.
9.1.3 Waste Disposal in Superdeep Holes
This concept would place waste far from man's environment by
placing it in holes which range from 10,000 to 20,000 meters in depth.
This great depth would assure that no conceivable climatic or surface
change would expose the waste to the biosphere.
The final storage facility using this concept would consist of
a large number of large diameter holes drilled into a thick and
9-8
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Source: Battele, Pacific Northwest Laboratories, Reference 2.
FIGURE 9-4
SOLID WASTE EMPLACEMENT IN A MATRIX OF DRILLED HOLES
9-9
-------
hydrologically tight sequence of rocks. The waste would be lowered in
canisters into the hole until they fill the bottom one or several
thousand meters of the hole. After filling the hole to a predeter-
mined level, the hole would be sealed.
The concept has many inherent problems. The technology to drill
a large diameter hole to such a great depth, as required by this con-
cept, does not exist today. The time involved in drilling such holes
would be close to six years per hole. It is obvious that an enormous
financial investment would be necessary to drill the number of re-
quired holes and neither the time nor the cost to develop such tech-
niques are known. Another consideration is a limitation on the number
of canisters which can be placed per hole if melting of waste and rock
is not permitted. Temperature problems are greater as you drill
deeper. The rocks may be at a temperature just below melting and the
added heat from the waste may induce melting.
9.1.4 Deep-Well Injection
Industry uses deep-well injection for disposal of liquid wastes
today. The concept is simple: the liquid waste is pumped down the
hole and forced into the geologic formation. Pressures required for
pumping range from zero to 10^ kg/m^. The host formation must
have a porosity of 10-30 percent, a permeability of at least 25 milli-
darcies*, and a depth of at least 1000 m. The formation must be
*1 darcy = the passage of 1 cc-per second of a fluid with 1 centi-
poise viscosity under a pressure difference of 1 atmosphere through
a porous medium with a cross-sectional area 1 sq cm and length 1 cm.
9-10
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bounded by impermeable strata and must be free of•water-transmitting
faults. Such formations occur in the sedimentary basins of the U.S.;
however, it is in these basins that oil and gas companies are, explor-
ing for petroleum and natural gas. This exploration can cause a major
safety problem of connecting waste disposal zones with aquifers.
Other important safety factors are proper casing of the injection well
and monitoring and maintenance of integrity of all pipes and casings.
Technology needed for this concept is available today; however,
its potential for use with liquids containing long-lived or high
levels of radioactivity has not been evaluated.
9.1.5 Hydrofracture
Hydrofracture is a concept which is currently being used by in-
dustry to either stimulate oil and gas production or for the disposal
of wastes. The technology is therefore commercially available.
The concept has three basic steps for the emplacement of waste in
a rock sequence such as shaie:
Step 1. Breakdown of the geological formation. A viscous
fluid which has a gelling and propping agent added
to it is pumped under pressure into the well until
the formation fractures.
Step 2. Preparation for waste injection. A fluid with a
gel breaking agent is pumped in and then drained
out, leaving the propping agent behind to keep
the fractures open.
Step 3. Waste injection. The waste fluids mixed with a
grouting agent are injected Into the fractures.
The grout hardens and fixes the waste in the
formation.
9-11
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Oak Ridge National Laboratory (ORNL) has used this method of
disposal for intermediate level wastes since 1959. The concept has
not, however, been demonstrated for high-level, long-lived wastes.
As with the prior concept of deep well injection, the long-term con-
tainment capability is in question.
9.1.6 Rock-Melting Concepts
The following concepts involve melting of the waste and the sur-
rounding rock. In three of the four concepts, the melted waste and
rock are permitted to mix and resolidify as a rock-waste matrix. In
the fourth concept, the capsule containing the waste remains intact
and melts its way down through the earth's crust. The depth to wnich
the waste penetrates is a function of its aging. Values between 4 and
10 km have been quoted depending on the aging period.
None of the disposal methods involving melting have been exten-
sively investigated, therefore the concepts presented here involving
melting are based on preliminary calculations and experiments and, in
some cases, conjecture.
9.1.6.1 Mined Cavity/Liquid Waste/Interim Cooling. This concept
involves mining a cavity in an isolated, deep (300 to 3000 m) geologic
formation (probably an intrusive igneous rock type such as granite)
under the fuel reprocessing plant. A cavity having a volume of about
6000 m^ (a sphere of about 12 m radius) could dispose of 25 years
waste from a 5 ton/day reprocessing plant.3
After the cavity is formed, waste would be directly injected from
the plant. Cooling water would be necessary as the waste begins to
9-12
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boil because of the heat generation of the radionuclides. The steam
and other gases created would be collected and sent through a conden-
sation and treatment plant to minimize steam transport of radionu-
clides. When the cavity is filled with waste, the cooling water will
be stopped and all access holes will be sealed. Melting would then
follow and would continue for about 65 years, reaching a maximum melt
radius of 96 m.
The problems involved with this concept stem from both the
emplacement of the liquid waste and from the subsequent melt. During
emplacement of the waste, it is necessary to control steam transport
of radionuclides to prevent leakage of the waste into an aquifer con-
taining mobile water and possibly to have design features to mitigate
buildup of silica scale in the steam exhaust line. After sealing the
cavity, steam pressures will build and may cause movement along faults
or cracks that may be unknown at the time of emplacement, or may cause
new ones to form. When melting begins, the surrounding rock may crack
or deform from thermal stress. If this cracking occurs, it may create
a new pathway to man's environment.
9.1.6.2 Mined Cavity/Solid Waste/Interim Cooling. In this
concept, waste is received from the reprocessing plant in solid form.
The waste in canisters is placed in a mined cavity in such a manner as
to require interim cooling to prevent melting of the canisters, their
contents, and the surrounding rock strata. Cooling would be carried
out by filling the cavity containing the waste with water. The cool-
ing water would circulate around the canisters and then to the surface
9-13
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where it would be passed through heat exchangers on the surface. With
this concept there is the capability to retrieve any or all the canis-
ters at any time before final sealing and subsequent melting.
After the cavity is full, the cooling water circulation would be
stopped and the remaining water would boil away. As soon as the water
has boiled away, the waste, canister, and surrounding rock would melt.
The rock melt would dilute the waste to a low concentration.
9.1.6.3 Deep Drilled Hole/Solid Waste/No Interim Cooling. This
concept places solid waste in deep-drilled (several km below the sur-
face) holes. The host rock would probably be an igneous intrusive
type. The waste would be placed in the holes in either expendable
canisters or with no canister at all. The heat of decay of the waste
melts the waste, the canisters, if any, and the surrounding rock. The
waste rock melt mixes by natural convection currents and then resolid-
ifies as it loses heat to the surrounding rock as its heat generation
capability decreases. The top of the cavity is then sealed with glass
which melts at low temperature. After the glass has resolidified, the
remainder of the access hole can be filled with concrete or other
suitable material.
This concept has promise but further study is needed in order to
fully understand the interaction between the waste and the surrounding
rock, both melted and unmelted. Also not fully understood are the
long-term radionuclide migration and transport in the host rock; the
geologic conditions in deep bedrock; the details of heat transfer; and
the transport of volatile and gaseous products from the waste.
9-14
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9'. 1.6.4 Solid Waste/Capsule/Deep Descent. This concept calls
for a capsule of waste to be placed in a drilled hole up to 2 km deep,
which may be partially cased. As the capsule is lowered into position
it is cooled by a retrievable cooling system. When in place, the
cooling system i« shut down and retrieved. The decay heat melts the
waste but not tne capsule. The capsule transfers the heat to the host
rock which melts. Because of its greater density, the capsule settles
to the bottom of the melt chamber in a continuing process. The melt
at the top of the chamber resolidifies, forming a permanent seal.
After a suitable time has elapsed the hole can receive another cap-
sule. This permits one hole to be used for several capsules. The
host rock can range from salt domes to intrusive igneous rocks for
this capsule.
Problems with this concept are in the area of early capsule fail-
ure as well as capsule configuration so 'as to maximize the amount of
waste in each capsule. Capsule size, however, is a tradeoff between
several factors, including handling convenience, safety during loading
and emplacement, borehole diameter, and thermal properties of the
waste.
9.2 Siting (Environmental) Considerations
9.2.1 Geologic, Hydrologic, Climatic, and Other Criteria Which
May Affect Long Term Confinement
Concepts for dispersing of high-level radioactive waste will be
dependent upon many considerations. These considerations must be
dealt with in order to assure safe disposal and effective long-term
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containment of the waste. The areas of primary consideration which
affect the pathways of the radionuclide are as follows:
• Thermal properties of the host rock
• Engineering properties of the host rock
• Water content of the rocks and water movement
• Mineral resources potential
• Geothermal resource potential
• Geographic characteristics
• Seismicity and faulting
• Depth of disposal
• Dimensions of the host rock
• Climate of area and possible changes and their effects
on erosion rate
The most suitable rock types for the concepts discussed are
1) intrusive igneous rocks (e.g., granite) or crystalline metamorphic
rocks (e.g., quartzite) because of their low permeabilities and high
mechanical strengths; 2) salt, either in domes or thick beds because
of its low permeability and self-healing properties; and 3) tuffs and
shales because of their low permeabilities and high ion-exchange capa-
cities. This list does not intend to imply any preference between the
rock types listed above. Sedimentary, except salt and shale, and
volcanic rock, exclusive of tuffs, are considered generally unsuitable
for waste emplacement because of their potential for high permeabil-
ity.
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Waste form is an important consideration, especially for those
concepts which are based upon emplacement of liquid wastes. The geo-
logic restrictions for liquid waste must be more stringent for several
reasons:
• higher mobility of the waste in its interim liquid form
• interim manmade barriers (a canister) are not present
• the concentration of waste and its heat are generally
higher than for initially solidified waste
An important consideration for concepts involving melting of
waste and the surrounding rock is whether or not extensive fractures
will develop as a result of the expansion of molten rock. Such frac-
turing may provide potential pathways to adjacent, possibly permeable,
saturated zones. There is also some potential for geysering resulting
from the buildup of heat after final sealing of the hole.
9.2.1.1 Thermal Propertie.s of the Host Rock
The dissipation of waste-generated heat is important to the
disposal of high-level waste. In order to dissipate heat quickly,
efficiently, and steadily, the host rock must have a high thermal
conductivity. The conductivity is important in order to minimize
surface extent of disposal areas and thereby cost. This is apparent
in these concepts for waste disposal with no interim cooling and no
melting of either waste or the host rock. In such a concept, a high
conductivity would allow more waste per unit area and would thereby
help minimize land area needed for the disposal site.
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The melting point of the host rock may also hold some signifi-
cance* In a concept where no interaction of the host rock and waste
is permitted, a host rock which has a higher melting temperature is
desirable. This would serve to minimize interaction of waste and rock
in the event of canister failure. The opposite would hold where the
concept calls for the formation of a rock-waste matrix. In this case,
a host rock with a lower melting point than the waste is desirable in
order to promote rapid mixing of rock and waste.
9.2.1.2 Engineering Properties of the Host Rock
This consideration deals with the mechanical strengths of the
host rock. It is obvious that the host rock must have sufficient
mechanical strength to allow either mined cavities or drilled holes to
remain open during waste emplacement. Rock can fail in many ways;
however, we are concerned basically with three modes of failure: rock
flow, rock bursting, and rock fracturing. Rock flow occurs when the
pressure of overlying layers causes rock to deform plastically. This
is common in shales and salt. Rock bursts, as the name implies, occur
as sudden releases of stress when the stress becomes greater than the
rock's mechanical strength. Fracturing may be more common in the
hydraulic-waste injection and deep-well injection disposal concepts.
The danger with this mode of failure is the creation of vertical
fractures in the rock which could lead to a breach of the host rock
and also possibly to a break of waterbearing strata.
Rocks with high mechanical strengths are desirable for disposal
of high-level wastes. Rocks which have high mechanical strength and
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still have generally low permeabilities are granites, gabbros, and
quartzites.
9.2.1.3 Water Content of the Host Rock
Groundwater movement is the main pathway by which radionuclides
are released into man's environment from disposal areas. It is,
therefore, very important that the host rock have as little water
content as possible. This includes connate water (water that is
formed at the same time as the rock) and fluid inclusions (water
trapped during crystallization of minerals).
Site selection must evaluate the possibility of over- and/or
underlying aquifers in the vicinity of the host rock under considera-
tion. Where such a situation cannot be avoided, all drilled holes
and shafts which penetrate aquifers must be cased and sealed off to
prevent movement of material either into or out of the aquifers.
9.2.1.4 Mineral Resource Potential
Exploration for minerals and their subsequent production by
future generations can be a potential threat to the long-term con-
finement of the high-level waste. Site selection for the disposal of
these wastes should take into consideration not only the candidate
host rock but also rock strata both above and below the host rock.
Mineral content and future economic value of the minerals should be
determined.
Past mining and/or drilling operations can also jeopardize long-
term containment. When the site has been chosen, all past mining
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and drilling operations must be located so that all mines, shafts, and
bore holes can be inspected and properly sealed.
9.2.1.5 Geothermal Resource Potential
With the current search for new energy sources, geothermal energy
is being sought and brought on line to help meet electricity and pro-
cess heat needs. Geothermal energy exploration and development, as
with mineral exploration and development, poses a threat to long-term
confinement of the waste.
The areas which are thought to be good prospects for geothermal
energy are areas which typically have had recent (< 1 x 10^ yrs)
volcanic activity and/or tectonic stresses. For this reason, these
areas are undesirable for waste disposal. Also, areas which have
above average geothermal gradients are also undesirable because of
future geothermal resource potential.
9.2.1.6 Seismicity and Faulting
Seismic and tectonic stability of the rocks in the disposal site
is of paramount importance. As has been stated before, all avenues
whereby groundwater can penetrate and remove waste must be avoided or
sealed off. Crustal cracking and faulting poses a real and great
threat to the long-term confinement of high-level waste. It does so
by having the potential to rupture the disposal zone and the canis-
ters. In doing so, it can provide excellent pathways for chemical and
groundwater removal of the waste and possible exposure to man's
environment;
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All areas subject to high seismic risk should be eliminated from
consideration as possible disposal sites. Sites of lower seismic risk
should undergo extensive monitoring and detailed mapping to establish
the degree of risk to long-term containment of high level radioactive
waste. Only those sites which have the lowest risk should receive
further consideration.
9.2.1.7 Depth of Disposal
In general, for a given disposal concept with increasing depth
there is greater assurance of long-term containment. There is, how-
ever, a need to set a minimum depth at which high-level waste can be
disposed of. A minimum depth of 300 meters has been proposed.^ In
areas where this depth would conflict with local water, supply aqui-
fers, a greater minimum depth would be required. This would also
apply to areas where excessive erosion may occur. These minima are to
assure isolation and long-term containment of the waste from man's
environment.
£>
As stated earlier, in general, the greater the depth, the greater
the assurance of isolation. There are limitations, however. Mined
cavities can only be mined to depths where the temperature is low
enough to allow man to work. In a typical mine with a geothermal
gradient of 20°C/km (20°C at surface), a temperature of 60°C (140°F)
is reached at a depth of 2000 meters without artificial cooling. For
depths greater than 2000-3000 meters, riethods must be used which do
not require human entry. The limitations which affect using these
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greater depths include the temperature at these depths and its effect
on canister stability and waste-rock interaction. Also, the degree of
difficulty and cost of drilling increase with increasing depth.
9.2.1.8 Dimensions of Host Rock
The dimensions of the host rock should be such that long-term
containment can be obtained. In order to do this, the host rock must
not only have relatively great thickness but great enough lateral or
horizontal extent. Site selection will have to set up minima for
these dimensions. Within this specific site selection, the following
factors will come into play:
• Total size and shape of host rock formation
• Thickness and extent of surrounding formations
• Homogeneity and isotropy of the host rock
• Thermal properties of host rock
• Hydrological characteristics of both the host rock and
surrounding formations
• Waste form
• Chemical properties of host rock and surrounding formations
9.2.1.9 Climate and Possible Changes in Climate
This consideration goes hand in hand with several of the preced-
ing considerations. A dry climate is desirable because it will reduce
the amount of groundwater available to leach waste and also reduce the
rate of erosion. If such an area is chosen, and there is a change in
the climate such that this relatively arid climate becomes a wet rain
forest type of climate, the hydrologic regime of the area will change
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and may pose a threat to the long-term confinement by groundwater
leaching and by increasing the rate of erosion.
On a world-wide scale, if world climate becomes warm enough to
melt the polar ice caps, either partially or totally, a change in sea
level would endanger waste which is disposed of in areas which may be
inundated. The opposite is also true. If a new age of glaciation
began, any waste buried in areas which may become eroded by glacier
movement would have its long-term confinement jeopardized.
A careful analysis of the proposed site must be performed in
order to minimize risks to the long-term confinement of the high level
radioactive wastes.
9.2.2 Pathways and Barriers of Migration of Nuclides
There are several methods by which the radionuclides can be
released from containment and eventually enter the biosphere:
• groundwater intrusion
• faulting
• diapirism
• erosion
• fall of meteorites
• magma intrusion
• change in base drainage levels
These methods of release are minimized before any barriers such as
containment vessel and waste form are considered by careful site
selection prior to waste emplacement.
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There are barriers which can be further used to assure that the
radionuclides in the waste do not reenter man's environment during the
time required for them to naturally decay to innocuous levels:
• waste form
• canister containment
• geologic system of host rock
In some cases, the barriers must be able to contain the waste for
many thousands of years. Such a case is 1-129 (half-life 17 x 10^
years). Therefore, the probable effectiveness of these barriers will
be presented in the following discussion. This will be done as a
comparison of barriers and the methods of release and migration.
9.2.2.1 The Waste Form
The waste form will be an important barrier to the migration of
radionuclides after canister failure. Various solid waste forms have
been considered. These include calcined waste, vitrified (glassi-
fication) waste, and waste incorporated in a metal matrix. A boro-
silicate glass waste form is presently favored both because of its
resistance to leaching and the more advanced state of technical
development. In the case of borosilicate glass, it has been estimated
that "for a cylinder of glass 0.75m high and 0.5 m in diameter, it
would take 20 to 200 million years for 99 percent of the initial load
of radionuclides to be extracted."5 For this, it is assumed that
the integrity of the cylinder is maintained.
Questions have been raised concerning the long term integrity of
the glass form. Heat and radiation range, high pressures, and other
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factors could result in the failure of the glass form. In the event
the glass is fragmented, a greater surface would be exposed and the
leaching rate would increase accordingly. In addition, it has been
observed that interstitial water migrates towards the heat source in a
salt formation. It has also been postulated that the chemical com-
pounds present in a salt formation could form a brine of high leaching
capability. It is possible, therefore, that the waste form would
provide containment for only tens of years rather than hundreds to
thousands of years. Containment would then be dependent upon the host
rock.
Groundwater leaching is the chief method of release and migra-
tion, and for the long time period involved it is prudent to assume
that at sometime groundwater will come into contact with the waste.'
The other methods of release listed above, as well as accidental
access by man, may aid in water contact by providing pathways for
water to follow toward the waste. In the event that the waste form
can maintain its integrity for hundreds to thousands of years even
though the waste is eventually leached out, the time delay will be
long enough to eliminate most of the potentially high levels of fis-
sion product radionuclides which could find their way back to man's
environment.^»° The exceptions are the long half-life fission pro-
ducts and activation radionuclides and the actinides. The potential
for leaching of radionuclides from a rock-waste melt mix has not been
determined.
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9.2.2.2 The Canister
The choice of metal for the canister is likely to be from stain-
less steel, carbon steel, and .titanium. Carbon steel and stainless
steel are not expected to survive more than a few hundred years, how-
ever, it has been suggested that titanium may last for up to 1000
years. It is clear that the canister is not intended to provide con-
tainment in the long term. Its role is one of containment in the
short term when the high-heat generating fission products are in abun-
dance. The canister also aids in handling the waste during emplace-
ment and recovery, if desired. The largest role may, however, be in
preventing rock-waste interaction during the time of highest possible
thermal flux which could cause interactions to occur.
i
The canister will probably be destroyed before about 500 to 1000
years by the geologic environment it is buried in. It is then that
the waste form (contained in glass) will become the important barrier.
By this time most of the fission products will be gone so that the
primary concern is that of migration of long-lived radionuclides.
Following loss of the canister and after leaching from the glass, or
if the glass is destroyed, the final barrier or delaying action comes
into play—the geologic system of the host rock.
9.2.2.3 Geologic System of the Host Rock
The geologic properties of the host rock as stated in Section 8.2
are some of the most important barriers to groundwater leaching of the
waste. Site selection must be carried out with three major criteria:
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• Hydraulic regime
• Geologic stability
• Retention of radionuclides
The first two of these criteria will eliminate areas which would
be highly prone to faulting, diapirism, high erosion rates, magma
intrusion, and changes in base drainage levels in the near geologic
future. These criteria would also address the permeability of the
strata surrounding the host rock as well as the host rock itself. Low
permeabilities, along with mechanisms, e.g., ion-exchange capacity,
form the host rock's ability for retention of radionuclides. The
depth of the waste's emplacement would preclude impact from meteorites
as a threat to the repository's integrity.
The ability to retain nuclides by ion exchange is essential to
long-term confinement of long-lived nuclides. For the length of time
needed to reduce some of the long-lived nuclides to safe levels (e.g.,
1-129, Np-237, Pu-239), ion-exchange capability can be more important
than permeability and depth. Regardless of the host rock's permeabil-
ity and depth (between the 300 and 6000 m considered here), there is
sufficient time for groundwater to penetrate the repository and return
the nuclides to man's environment. "Therefore, a geologic formation
should not be considered a confining barrier for radionuclides with
very long half-lives for which it has no ion-exchange capacity,...""
It has been suggested that it may be possible to artificially set
up this ion-exchange capacity.in the host rock, adding to its natural
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capacity, by burying compounds with the waste which would react with
soluble ions of the radionuclides to form an insoluble precipitant.
This geochemical barrier would provide an additional method of keeping
long-lived radionuclides from man's environment for extremely long
periods of time. "The greater the ion exchange of the surroundings
for a radionuclide, the greater its confinement will be; this con-
finement may even be total."6 The physiochemical reactions which
will retard the transport of radionuclides include phenomena such as
adsorption and colloid filtration as well as ion exchange. The dis-
tribution coefficient and retardation factor which are a measure of
the sorbtion capability of soils, sediments, and geologic formations
were discussed in Section 7 and presented in Table 7-II, for a typical
desert soil. Similiar type information is required for specific sites
for waste disposal in order to assess their capabilities to provide
long-term isolation. Acceptability, however, includes consideration
of the initial quantities, the half-life, and the health hazard of the
radionuclide as well as the retardation capability of the geological
formation. In addition, the various chemical form which the radionu-
clide may take following leaching from containment and interaction
with the host medium must be considered in regard to the sorption
effect.
9.3 Technical Feasibility of Alternative Geological Disposal Concepts
The technical feasibility of the concepts helps set up criteria
which must be met in order for the concepts to be regarded as viable
alternatives to deep mined geologic repositories:
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• Achievability with current technology
• Achievability with technology based on current theory
• Ability to provide long-term confinement
• Ability to meet retrievability requirements
It is felt that all of the concept.0 described earlier can be
implemented using extensions of current technology, with the exception
of supe^deep holes. The technology to drill such deep holes at large
diameters does not currently exist. This does not mean that it is not
feasible with extensions of current technology. No significant break-
throughs are needed in technology and no uncommon construction, min-
ing, drilling, or operational problems are foreseen with the exception
of super-deep holes and with the concepts which call for the formation
of a rock waste matrix. Drilling techniques must be developed which
will allow drilling of large diameter holes to the depths required for
the super-deep concept to become technically feasible. Therefore, all
the concepts described seem to be technically feasible using future
technology based on current theory and technology. The concepts which
involve melting of rock and waste to form a rock-waste matrix need
study in the area of the behavior of the molten rock-waste from the
time of waste emplacement to the time the rock-waste matrix is solidi-
fied in its final disposal form.
Long-term containment is a major concern in the disposal of high-
level waste. It is very important, therefore, that all concepts
assure long-term containment. The major threat to long-term contain-
ment i$vgroundwater. The concepts must preclude contact of the waste
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with groundwater in order to minimize waste migration to the bios-
phere. Several of the concepts may have problems with groundwater
leaching:
• Disposal in a matrix of drilled holes
• Deep well injection
• Hydro fracture
• Rock melting concepts
The matrix of drilled holes may have a problem because of the
many penetrations of the host rock. Each of the drill holes offers a
possible pathway to the biosphere. Development and confirmation of
sealing techniques would be required.
Deep-well injection, as well as hydrofracture, involves pumping
liquid waste into the host rock formation. It is possible that forced
injection may form vertical fractures whic.h may give the waste a path-
way to waterbearing strata. Techniques of monitoring fracture forma-
tion are needed. Although both of these concepts are commercially
available, a study of the feasibility of using these concepts for
disposal of high-level radioactive waste is needed.
The rock-melting concepts are suspect because of the lack of
knowledge of the behavior of the rock-waste melt. Until the uncer-
tainties of its behavior can be resolved, these concepts cannot be
considered to assure long-term containment.
Retrievability in high-level waste disposal is very difficult.
The concepts involving drilling holes for waste emplacement have
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^limited retrievability as does the solution-mined cavern concept. All
of these concepts involve the waste in a solid form at time of
emplacement. Hydrofracture and deep-well injection have no retriev-
ability capabilities. Two of the rock-melting concepts have limited
retrievability only during interim cooling and emplacement, while the
other two have no retrievability. It should be remembered that in
final disposal no retrievability is assume,d, therefore, this criterion
is not of utmost importance.
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REFERENCES
1. "Technical Support for the Radiation Standard for High-Level Radio-
active Waste Management," Tasks A to D, Draft, Arthur D. Little,
Inc.
2. . "Alternatives for Managing Wastes from Reactors and Post-Fission
Operations in the LWR Fuel Cycle," Vol. 4, Battelle, Pacific
Northwest Laboratories, Report #ERDA-76-43, May 1976.
3. Kubo, Arthur S., and Rose, David J., "Disposal of Nuclear Wastes,"
Science, Vol. 182, Number 4118, pp. 1205-1211, 21 December 1973.
4. Schneider, K.J. and Platt, A.M., Editors, "High-Level Radioactive
Waste Management Alternatives," Sections 3 and 4, Battelle,
Pacific Northwest Laboratories, Report No. BNWL-1900, May 1974.
5. Cohen, Bernard L., "The Disposal of Radioactive Wastes from Fis-
sion Reactors," Scientific American, Vol. 236, Number 6, pp. 21-
31, June 1977.
6. de Marsely, G., Ledoux, E., Barbreau, A., and Margot, J., "Nuclear
Waste Disposal: Can the Geologist Guarantee Isolation?" Science,
Vol. 197, Number 4303, pp. 519-527, 5 August 1977.
7. The Study Group on Nuclear Fuel Cycles and Waste Management, "The
Nuclear Fuel Cycle: An Appraisal," Physics Today, October 1977.
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