c/EPA
United States
Environmental Protection
Agency
Office of
Radiation Programs
Washington DC 20460
EPA 520/4-79-007B
Radiation
Technical Support of
Standards for High-Level
Radioactive Waste
Management
Volume B
Engineering Controls
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TECHNICAL REPORT DATA
(Please read Instructions on the reverse before completing)
1, REPORT NO.
EPA 52(1/4-79-0078
^-RECIPIENT'S ACCESSION NO.
4. TITLE AND SUBTITLE
Technical Support of Standards for High-level
Radioactive Waste Management, Volume B, Engineering
Controls ,
5. REPORT DATE
1979
6. PERFORMING ORGANIZATION CODE
7. AUTHOFMS!
8, PERFORMING ORGANIZATION REPORT NO
9. PERFORMING ORGANIZATION NAME AND ADDRESS
Arthur D. Little, Inc.
Cambridge, Massachusetts 02140
10. PROGRAM ELEMENT NO.
11. CONTRACT/GRANT NO.
68-01-4470
12. SPONSORING AGENCY NAME AND ADDRESS
Office of Radiation Programs
Environmental Protection Agency
Washington, B.C. 20460
13. TYPE OF REPORT AND PERIOD COVERED
March - August1977
14. SPONSORING AGENCY CODE
15. SUPPLEMENTARY NOTES
16, ABSTRACT
This report is the result of work performed under the second part (Task B) of a
four-part contract to gather technical information to evaluate environmental
acceptability of various options for disposing high-level wastes. The other tasks
are: A-Source Term Characterization; C-Assessment of Migration Pathways; B-Assessment
of Accidental Pathways.
The scope of work for Task B was divided into 2 major steps: analysis of
technology for engineering control of high-level wastes, and projections of costs for
various alternative disposal technologies. Primary emphasis has been placed in
non-reprocessed spent fuel. Brief consideration has been given to lower-level
transuranic contaminated wastes which arise primarily from plutonium fuel fabrication
and reprocessing plants.
The objective of this limited review is to provide some insight into the state
of the art for applying engineering controls to such wastes, including, where
necessary, direct burial with high-level wastes in a geologic disposal facility.
KEY WORDS AND DOCUMENT ANALYSIS
DESCRIPTORS
b,IDENTIFIERS/OPEN ENDED TERMS C. COSATI Field/Group
high-level wastes
radioactive waste disposal
radioactive waste management
engineering controls for high-level wst di
waste disposal technologies
18. DISTRIBUTION STATEMENT
19, SECURITY CLASS (This Report)
21. NO. OF PAGES
20. SECURITY CLASS (Thispage}
22. PRICE
EPA Form 222G-1 (Rev. 4-77) PREVIOUS EDITION is OBSOLETE
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TECHNICAL SUPPORT OF STANDARDS FOR
HIGH-LEVEL RADIOACTIVE WASTE MANAGEMENT
TASK B REPORT
EPA Contract No. 68-01-4470
Prepared by
Arthur D. Little, Inc.
Cambridge, Massachusetts 02140
March-August 1977
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DISCLAIMER
This report was prepared as an account of work sponsored by the
Environmental Protection Agency of the United States Government under
Contract No. 68-01-4470. 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 accu-
racy, completeness, or usefulness of any information, apparatus, product,
or process disclosed, or represents that its use would not infringe
privately owned rights.
il
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ACKNOWLEDGMENTS
Many individuals contributed to the work done under the direction
of Arthur B, Little, Inc., for the U.S. Environmental Protection Agency
under Contract So. 68-01-4470. John L. Russell and Daniel Egan of the
Office of Radiation Programs at EPA served as constant guides in the
process of our work. Dr. Bruce S. Old, James I. Stevens, and David I.
Hellstrom of Arthur D. Little, Inc., were Program Director, Program
Manager, and Assistant Program Manager, respectively, of the overall
project* Key individuals involved in each of the reports prepared
under the four tasks were:
TASK A
TASK B
Donald Korn
"Arthur D. Little.
Task Director
Inc.
Robert McWhorter,
Michael Raudenbush,
and Lester Goldstein
S.M. Stoller Corp.
_Edwin L. Field
Arthur D. Little, Inc.
Task Director
TASK C
TASK D
Robert McWhorter and
Michael Raudenbush
S.M. Stoller Corp.
P.J. O'Brien
'Arthur D. Little, Inc.
Task Director
Dr. Ronald B. Lantz
Intera Environmental
Consultants, Inc.
Dr. John Gormley
D'Appolonia Consulting
Eng ineer s, Inc.
Donald S. Allan
Arthur D. Little^ Inc.
Task Director
Ajit Bhattacharyya and
Charles R. Hadlock
Arthur D. Little, Inc.
ill
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FOREWORD' / ' ' '
A major Federal effort is underway to develop methods for disposal
of high-level radioactive waste in deep geologic repositories. An impor-
tant element of this program is the development and promulgation by the
U.S. Environmental Protection Agency (EPA) of environmental standards
for the management of these wastes.
In anticipation of its efforts to develop these standards, EPA
recognized that it would be necessary to estimate the expected and
potential environmental impacts from potential geologic repositories
using modeling techniques based upon as thorough an understanding as
.possible of the uncertainties involved in the quantities and charac-
teristics of the wastes to be managed, the effectiveness of engineering
controls, and the potential migration and accidental, pathways that might
result in radioactive materials entering the biosphere. Consequently,
in March 1977, the EPA contracted with Arthur D. Little, Inc.,for a study
to provide technical support for its development of environmental regula-
tions for high-level radioactive wastes. This study was divided into
the following four tasks:
Task A - Source Term Characterization/Definition
B - Effectiveness of Engineering Controls
C - Assessment of Migration Pathways
D - Assessment of Accidental Pathways
The information presented in the reports on these tasks was developed
principally during the period March 1977 to February 1978. In the case of
this report, Task B, the information contained in it was prepared during
the period March-August 1977. There are many national and international
programs underway to develop additional data, especially in the fields
of waste forms, knowledge of geology and geohydrology, and risk assess-
ment. The information presented in these reports has been developed
on conceptual bases and is not intended to be specific to particular
conditions at geologic repositories.
iv
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TABLE OF CONTENTS
Page
Acknowledgements ill
Foreword iv
List of Tables viii
List of Figures x
B-l.0 INTRODUCTION 1
1.1 BACKGROUND AND PURPOSE OF STUDY 1
1.2 SCOPE OF TASK B EFFORT 4
B-2.0 SUMMARY 5
2.1 GENERAL 5
2.2 EVALUATION OF PROCESSING AND PACKAGING TECHNOLOGY 5
2.2.1 High-Level Liquid Wastes 6
2.2.2 Cladding Hulls and Fuel Bundle Residues ' 7
2.2.3 Spent Fuel Elements 8
2.3 SELECTION OF REFERENCE CASES FOR TASK D RISK
EVALUATIONS , 8
2.4 ANALYSIS OF ALTERNATIVE GEOLOGIC DISPOSAL
TECHNIQUES 9
2,5 EFFECTIVENESS OF ENGINEERING CONTROLS 14
2.5.1 Canister Integrity 14
2.5.2 Matrix Leach Resistance 14
2.5.3 Solubility Characteristics of Leached Waste Forms 16
2.6 COST CONSIDERATIONS 16
B-3.0 ANALYSIS OF TECHNOLOGY 17
3.1 GENERAL , 17
3.2 ALTERNATIVE DISPOSAL FORMS FOR HIGH-LEVEL WASTES 18
3.2.1 High-Level Liquid Wastes 18
3.2.2 Cladding Hulls and Fuel Bundle Residues 52
3.2.3 Storage Canisters 63
3.2.4 Off-Gas Treatment 66
3.2.5 Spent Fuel Assembly Disposal 70
3.3 SELECTION OF REFERENCE TECHNOLOGIES 74
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TABLE OF CONTENTS (continued)
Pace
B-3.3.1 Calcination 75
3.3.2 Classification 75
3.3.3 Spent Fuel Disposal 75
3.3.4 Canisters 76
3.3.5 Off-Gas Residue Disposal 76
3.3.6 Cladding Hulls and Fuel Bundle Residues 77
3,3.7 Summary of Reference Cases 78
3.4 DISPOSAL OF OTHER TRU-CONTAMINATED WASTE 79
3.4.1 Combustibles 81
3.4.2 Non-Combustibles 84
3.4.3 Wet Wastes (Liquids and Sludges) 88
3.4.4 Final Packaging 88
3.5 ALTERNATIVE GEOLOGIC DISPOSAL TECHNIQUES 88
3.5.1 Design Characteristics for Disposal in Bedded Salt 90
3.5.2 Disposal in Other Media 106
3.5.3 Thermal Analysis 111
3.6 EFFECTIVENESS OF ENGINEERING CONTROLS 142
3.6,1 Reference Technologies in Bedded Salt 142
3.6.2 Reference Technologies in Other Media 156
B-4.0 COST CONSIDERATIONS 159
4.1 GENERAL 159
4.2 SPENT FUEL STORAGE COSTS 160
4.3 POST STORAGE/PRE-BURIAL COSTS 160
4.3.1 Solid HLW Disposal Path 160
4.3.2 Spent Fuel Disposal Path 164
4.4 REPOSITORY COSTS 165
4.4.1 Costs for Geologic Disposal in Salt 165
4.4.2 Cost Versus Depth 171
4.5 COST SUMMARY FOR HLW OR SPENT FUEL DISPOSAL IN SALT 171
4.6 COSTS FOR GEOLOGIC DISPOSAL IN HARD ROCK 171
vi
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TABLE OF CONTENTS' (continued)
B-4.7
NOTE ON THE DISPOSAL COSTS FOR LOW-LEVEL
TRU WASTES
Page
171
References
Appendices
Appendix B-I
Appendix B-II
Appendix B-III
Appendix B-IV
Appendix B-V
Glossary and List of Abbreviations
Solidification Processes in Early
Stages of Development
Off-Gas Treatment Technology
Details of Reference Disposal-
Site Facilities
Corrosion of Metal Containers
175'
179
B-I-1
B-II-1
B-III-1
B-IV-1
B-V-1
vii
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LIST OF TABLES _• ,
Table No. Page No.
B-l Reference Cases for Various Haste Forms 10
B-2 Waste Calcine Characteristics 20
B-3 Waste Glass Characteristics 20
B-4 Principal U. S. Solidification Programs 22
B-5 Principal Foreign Solidification Programs 24
B-6 Solidification Product Characteristics 25
B-7 Characteristics of Borosillcate Glass 39
B-8 Fuel Bundle Properties 53
B-9 Fuel Assembly Disposal Summary 73
B-10 Incinerators Used in the United States for the
Combustion of Radioactive Solid Wastes 85
B-ll Processes Under Development In the United States for
the Combustion of Radioactive Solid Wastes 87
B-12 Volumes of Waste to be Stored in Reference Facility 93
B-13 HLW/Spent Fuel Retrievable Storage Area Data 99
B-14 Reference Facility Dimensions for TRU Waste iQl
B-15 Total Facility Area Requirements 102
B-16 Conservative Estimates of Gross Vertical Expansion
("Thermal Uplift") for Different Repository Media no
B-17 Thermal Properties of Geologic Media 115
B-18 Geologic Formations at Depths in Model 117
B-19 Parameters Affecting the Time-Dependent Heat
Generation Rate 119
viii
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LIST OF TABLES'-(ContM)
Table No. PageNo.
B-20 Glass Leach Rate Constants 146
B-21 Solubility Limits of Principal Species from
Actinide Dissolution 157
/
B-22 Approximate Costs, Reference Repository in Bedded
Salt—Non-Retrievable Storage for Equivalent of
107,000 MTHM 166
B-23 Approximate Costs, Reference Repository in Bedded
Salt—Fully Retrievable Storage for Equivalent of
107,000 MTHM 167
B-24 Plant Staff Estimate for Reference Repository 169
B-25 Summary of High-Level Waste Disposal Costs 170
B-26 Reference Facility Costs (Bedded Salt)—Non-Retrievable
Storage of Low-Level TRU Waste 172
APPENDIX
B-III-1 Production of Carbon-14 in Light-Water Reactors B-III-15
B-III-2 Description of Possible Flowsheet Variations B-III-27
B-III-3 Analysis of Flowsheet Variations B-III-28
B-V-1 Corrosion of Titanium Alloys in Sodium Chloride
Solutions B-V-3
B-V-2 Corrosion of Inconel Alloys in Sodium Chloride
Solutions B-V-6
B-V-3 Corrosion of Stainless Steels in Sodium Chloride
Solutions B-V-7
B-V-4 Corrosion of Mild Steel in Sodium Chloride Solutions B-V-9
B-V-5 Comparison of the Corrosion Rate of Zirconium with
other Alloys in Chloride Solutions B-V-10
ix
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LIST OF FIGURES
Figure No. Page
B-l Fluidized-Bed Calcination 29
B-2 General Calcination Flowsheet
for the NWCF 30
B-3 Continuous Inert Bed Fluidized-Bed Calciner 33
B-4 Spray Calciner for 5 MT/Day Reprocessing
Facility 35
B-5 Rotary Kiln Calcination 37
B-6 In-Can Melting 40
B-7 Process Summary - Continuous Ceramic
Melter . 43
B-8 French Rotary Kiln - Continuous
Metallic Melter 45
B-9 Original VERA Process 47
B-10 VERA Process Systems 48
B-ll Simplified Flowsheet of FINGAL Process 50
B-12 New HARVEST Pilot Plant Flow Diagram 51
B-13 Decision Areas in the Management of
Fuel Bundle Residues 56
B-14 Treatment of Fuel Hulls - Process-Product
Characteristics and Status 62
B-15 Typical Post-Fill Treatment of Solidifed
HLLW Canisters 67
B-16 Management Methods for Transuranic-
Contaminated Wastes 80
B-17 Development of Incineration Methods for
Transuranlc-Contaminated Wastes 82
B-18 Artist's Concept - Federal Repository 97
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LIST OF FIGURES (continued)
Figure No. Page
B-19 Geometrical Array Used for Thermal
Analysis 114
B-20 Assumed Values for Thermal Conductivity
vs. Temperature 116
B-21 Vertical Temperature Distribution in
Typical Bedded Salt Formation 120
B-22 Vertical Temperature Distribution in
Typical Granite Formation (150 k¥/Acre) 121
B-23 Vertical Temperature Distribution in
Typical Salt Dome 122
B-24 Maximum Mine Temperature vs. Time in
Three Geologic Formations (Reprocessed
Waste) 123
B-25 Maximum Mine Temperature vs. Time for
10-, 20-, and 50-Year-Old Waste in
Bedded Salt 124
B-26 Maximum Mine Temperature vs. Time for
10- and 20-Year-Old Waste in Granite 125
B-27 Vertical Temperature Distrubition in
Typical Granite Formation (60 kW/Acre) 126
B-28 Maximum Mine Temperature vs. Time for
10-Year-Old Waste in Bedded Salt
Formation 128
B-29 Horizontal Temperature Distribution in
Salt for Various Times After Burial
(3-D Calculation) 129
B-30 Vertical Temperature Distribution in Salt
for Various Times After Burial
(3-D Calculation) , 130
B-31 Isotherms in Salt Around Waste Canister
Midplane (Z=0) at 2-Year Point After Burial
(Q=3.95 kW/Canister) 131
xi
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LIST OF FIGUREg (continued)
FigureNo. Page
B-32 Horizontal. Temperature Distribution in
Granite for Various limes After Burial
(3-D Calculations) (150 kW/Acre) 132
B-33 Vertical Temperature Distribution in
Granite for Various Times After Burial
(3-D Calculation) (150 kW/Acre) 133
B-34 Isotherms in Granite Around Waste
Canister Midplane (Z=0) at the 2-Year
Point After Burial (Q=3.95 kW/Canister) 134
B-35 Horizontal Temperature Distribution in
Granite for Various Times After Burial
(3-D Calculation) (60 kW/Aere) 136
B-36 Vertical Temperature Distribution in
Granite for Various Times After Burial
(3-D Calculation) (60 kW/Aere) 137
B-37 Isotherms in Granite Around Waste
Canister at Canister Midplane (Z=0)
at 2-Year Point After Burial (Q=1.58
kW/Canister) 138
B-38 Comparison of Predicted Maximum Mine
Temperature vs. Time for a One- and
Two-Dimenslonal Analysis in Salt Dome 139
B-39 Temperature Rise vs. Radius at Canister
Midplane (Z=0) for a 1100-Acre Storage
Facility in a Salt Dome (Result of 2-D
Analysis) 140
B-40 Maximum Vertical Uplift vs. Radius for a
1100-Acre Storage Facility in Three Geo-
logic Formations (Results of 2-D Analysis) 1*1
B-41 Percent of Glass Remaining (Unleached) vs.
Time, k= 10"%/cm -day (Conservative Value) 151
—6 9
B-42 Annual Leach Rate vs. Time, k«10 g/cm -day
(Conservative Value) 152
B-43 Percent of Glass Remaining (Unleached) vs.
Time, k=10~8g/cm -day (Possibly Attainable
Value) 153
xii
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LIST OF FIGURES (continued)
Figure No.
B-44
APPENDICES
B-II-1
B-II-2
B-II-3
B-II-4
B-II-5
B-III-1
B-III-2
B-III-3
B-III-4
Page
—A n
Annual Leach Rate vs. Tine, k=10 g/cm -day
(Possibly Attainable Value) 154
Supercalcine Processes B-II-2
Sintering Processes B-II-2
Metal Matrix Formation B-II-4
Glass Ceramic Process B-II-4
Ion Exchange Fixation B-II-7
Schematic Diagram of the ORGDP
Selective Absorption Pilot Plant B-II1-5
Alternative Technologies for the
Management of C-14 B-III-19
Tritium Equilibrium Model B-III-26
One Proposed Tritium Separation
Scheme B-III-30
xiii
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B-l.Q INTRODUCTION
B-l.l BACKGROUND AND PURPOSE OF STUDY
One of the major environmental concerns associated with the pro-
jected increase in nuclear power generation is the handling and disposal
of radioactive waste. Highly radioactive wastes must be placed in secure
repositories so that their entry into the environment can be' prevented
for a long period of time.
Furthermore, these wastes must be managed in a fashion that assures
acceptafile risk to the environment from the detrimental effects of radio-
active contamination. In March 1977, the Environmental Protection Agency
(EPA) contracted with Arthur D. Little, Inc., for a study to provide
technical support for the development of standards for the management of
high-level radioactive waste (HLW).
In the context of this study, HLW sources include (1) irradiated
or "spent" nuclear fuel elements discharged from a nuclear reactor for
disposal; (2) those aqueous wastes containing the bulk of the fission
products, resulting from reprocessing of spent fuel for recovery and
recycle of uranium and plutonium; (3) the fuel cladding and structural
materials associated with the fuel. In addition, specific isotopes
produced as a result of nuclear reactions associated with irradiated
fuel constitute items of special interest from the viewpoint of
environmental protection, e.g., earbon-14, iodine-129, krypton-85, and
the longer-lived heavy elements.
The existence of HLW, mainly from defense programs, has already led to
concern about temporary storage of relatively large volumes of liquid
wastes, especially since there have been leaks of such wastes from under-
ground storage tanks at the Hanford reservation, albeit with no demon-
strable effects on the environment. The lack of fully-demonstrated
processes and technologies for long-term disposal has also caused concern.
The problems associated with management of HLW have been recognized
for some time, and considerable research and development work has been
done on many aspects of these problems. Although this report will review
the problems in detail, it is important to stress at the outset that
considerable progress has been made toward solving them. Indeed, a
preferred methodology for handling high-level wastes has evolved;
this method involves isolation of solidified forms of these wastes in
deep geologic repositories.
Briefly, the approach is to impose multiple barriers in series be-
tween the solidified radioactive waste material and man's environment.
If the probabilities of failure for each of these barriers are small and
reasonably independent, the overall probability of environmental
contamination can be made quite small. The multiple barrier concept has
been used successfully in other areas, e.g., control of radiation releases
from nuclear power plants.
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As for the barriers themselves, the principal ones arei
(1) Waste Matrix: Incorporation of the wastes into a solid
matrix material resistant to attack by groundwater~the major
vehicle for possible transport of wastes from the repository
to the environment.
(2) Containment; Packaging of solidified wastes in a suitable
container to reduce the probability of radioactive contamination
during handling and during disposal operations, and to delay
the onset of attack of the matrix material.
(3) Geologic Isolation; Deposit in a deep site selected for its
low probability of (a) groundwater penetration and uJtimate
transport to the environment, (b) massive faulting or uplifts
that would cause loss of isolation, and (c) accidental
penetration by man.
(4) Adsorption: Reducing radioactive contamination by using an
effective natural barrier. Substrate materials tend to retain
many radioactive elements temporarily. Therefore, even in the
event of transport'by groundwater, adsorption will serve as a
mechanism to decrease the transport rate of radioactive
material and thereby allow time for additional radioactive
decay to less-hazardous levels.
To characterize these barriers more specifically, a case being
given serious consideration as a reference concept in the United States
is incorporation of wastes in a glass matrix contained in a protective
metal canister, buried in a deep mine located In a geologically ancient
salt bed. The glass has excellent resistance to leaching by groundwater,
and the ancient salt strata, by their very existence, have demonstrated
long-term isolation from the surface and from contact with groundwater.
The preceding discussion has focused on geologic disposal of
high-level radioactive wastes from reprocessing operations. Within the
past year questions have arisen about disposal of another waste form—
spent fuel elements. The question arose because of an indefinite
moratorium on fuel reprocessing imposed by the present Administration,
concerned over possible diversion of plutonium for weapons production.
At the present time, it is not clear when reprocessing might be undertaken
in the United States; there is at least a question as to whether it will
ever be undertaken. If reprocessing is indeed not permitted, disposal
of non-reprocessed spent fuel would be required; i.e., spent fuel would
be the ultimate waste form.
In view of the very large fuel values (uranium and plutonium)
remaining in spent fuel elements, many people believe it unlikely that
spent fuel would actually be "thrown away" or permanently buried as
waste; instead, there would be an interim storage of spent fuel
elements, with ultimate reprocessing at such time as suitably
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stringent controls are placed on use of plutonium fuel. Nevertheless,
spent fuel elements have been considered as a possible waste form to
cover the possibility that the "throwaway" fuel cycle might be adopted.
Although much has already been accomplished in solving the problems
of managing radioactive waste, the primary thrust to date has been in
the research and development area, with only limited demonstration and
prototype projects. Accordingly, the Federal Government has substantially
increased the scope of its waste management programs. As part of the newly
expanded program, several avenues are being intensively pursued. These
include demonstration projects on engineering controls, such as fixation
in glass matrices, hydrogeological and geochemical investigations to
gain better understanding of potential pathways to the environment, and a
detailed terminal storage design program leading to construction and
operation of a disposal facility in at least two deep geologic forma-
tions.
EPA has already agreed to publish proposed generally applicable
environmental standards for the ultimate disposal of radioactive waste.
To help establish the technical bases for HLW standards, EPA has embarked
on this technical support project.
This technical information will be used by EPA in evaluating the
environmental acceptability of various options in the EKDA waste manage-
ment program and of presently-operating and proposed disposal sites for
high-level radioactive wastes. Furthermore, a major Environmental Impact
Statement (EIS) is being prepared by ERDA to assess the total U.S. waste
management program. The current schedule for issuing this EIS is during
calendar year 1977. Clearly, public understanding would be enhanced
if the proposed EPA environmental standards for high-level waste were
available during the review phase of this major EIS.
The schedule for this study was set by the original target date of
December 1977 for publication of proposed standards. According to the
terms of the contract, the study is limited to reviewing and assessing
published data from programs related to irretrievable deep geologic
disposal that are now being conducted under ERDA and NRC sponsorship,
in order to provide EPA with an independent assessment of the state of
the art in management of radioactive wastes.
The effort is divided into four tasks:
Task A - Source Term Characterization/Definition
Task B - Effectiveness of Engineering Controls
Task C - Assessment of Migration Pathways
Task D - Assessment of Accidental Pathways
This report represents the results of the work done under Task 'B
above.
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B-1.2 SCOPE OF TASK B EFFORT . ! >-
The scope of work for this task is divided into two major steps:
(1) analysis of technology for engineering control of high-level wastes
and (2) projections of costs for various alternative disposal technologies.
Most of the Task B effort has been devoted to (1), primarily because
the information available for (2) is limited. The technological assess-
ment starts with a comprehensive review of alternative disposal forms for
high-level wastes, including cladding hulls, fuel bundle residues, and
spent fuel elements themselves. Based on this review, reference disposal
forms have been selected to span a range of isolation effectiveness.
These reference cases can then be used in the Task C and D studies to test
the sensitivity of environmental risk to changes in waste form and engineer-
ing controls. For example, for HLW from reprocessing plants, both calcine
(high-leaehability) and glass (low-leachability) matrices are considered,
while for spent fuel both a high-integrity titanium canister and a low-
integrity steel canister are considered.
The overall measure of effectiveness of engineering controls, as
defined in Task B, is the rate at which radioactive material becomes avail-
able for further transport out of the repository. The rate of such
transport depends principally upon two factors: the rate of removal from
the matrix material, and the solubility of the radioactive elements in
the surrounding groundwater. Data are summarized herein for each of these
factors and for several alternative waste forms; these data serve as
basic inputs to the Task C and Task D efforts.
Primary emphasis has been placed on high-level radioactive waste
from reprocessing or present in non-reprocessed spent fuel (throwaway
cycle), Brief consideration has been given to lower-level transuranic-
contaminated wastes, which arise primarily from plutonium fuel fabrication
and reprocessing plants. The objective of this limited review is to
provide some insight into the state of the art for applying engineering
controls to such wastes, including, where necessary, direct burial with
high-level wastes in a geologic disposal facility.
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B-2.0 SUMMARY
B-2,1 GENERAL
The effectiveness of engineering controls for isolation of high-
level radioactive wastes hinges on two key factors: (1) the form of the
waste material and Its resistance to transport, and (2) the location and
design of the geologic disposal facility to achieve maximum isolation from
the environment. Item (1) includes the form of the waste product, the
type of containment, the resistance of the waste matrix to leaching, and
the solubility of the leached radioactive elements in groundwater since
all of these affect the rate at which water might transport radioactivity
from the repository. It Includes certain aspects of the facility design
that might conceivably influence the degree of isolation afforded by the
facility itself. In particular, the engineering design of the facility
has been characterized in sufficient detail to identify potential design
uncertainties that might impact on the Task D risk analysis. Other aspects
of facility design, such as characteristics of model sites, selection of
alternative geologic media, etc., are not considered herein, but are
discussed in the Task C report.
In assessing the effectiveness of engineering controls, consideration
has been given in this study to a spectrum of waste forms with varying
degrees of transport resistance and differing amounts of radioactive
material—-e.g., spent fuel with relatively higher actinide content and
higher leachability versus reprocessing plant wastes incorporated in
glass with lower actinide content and lower leachability. All wastes,
whatever the form, are assumed to be at least ten years old when processed
and placed in the repository. This assumption is consistent with the
reality of today's reprocessing moratorium as well as the probability
that, even if reprocessing were started, spent fuel will continue to be
stored for at least a ten-year period.
The major thrust of the analysis of engineering controls has been
aimed at salt deposits, because this is the only type of geologic medium
for which available design data are sufficiently detailed to permit mean-
ingful design reviews. For other media, such as shale, basalt, or granite,
only limited engineering data are available, and the development of addi-
tional data is beyond the scope of this study. Detailed characterizations
of repository designs have not been made for these alternative media.
Instead, this review has been limited to some generalized comparisons
between certain salt repository design parameters and those for other
geologic media.
B-2.2 EVALUATION OF PROCESSING AND PACKAGING TECHNOLOGY
A number of possibilities have been considered for processing
(solidification) and packaging (containment) of high-level wastes for
permanent disposal in a Federal repository. This study has considered
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various HLW sources: high-level liquid wastes, cladding hulls and
fuel bundle residues, and spent fuel elements. The preferred tech-
nologies for each of those waste sources will be briefly described In
the following paragraphs. Possible processing and packaging approaches
for low- to intermediate-level, transuranic (TRU) contaminated wastes as
well as for certain other waste products from reprocessing and waste-
processing plants, e.g., iodine-129, carbon-14, krypton-85, tritium,
and ruthenium-106, have also been reviewed. In general, the processing
problems associated with those other waste products seem solvable, although
refinements may be needed in the specification of regulations for disposal
of these wastes.
B-2.2.1 High-Level Liquid Wastes
High-level liquid wastes (HLLW) are generated by a spent fuel re-
processing plant. Present NRC regulations (10 CFR 50, Appendix F)requlre
solidification of HLLW within five years after reprocessing and transfer
to a Federal repository within ten years after reprocessing.
Extensive research and development effort has taken place throughout
the world on possible alternatives for processing of HLLW, Some techniques
have been successfully demonstrated on a near-commercial, or commercial
scale; most of this work has focused on converting HLLW to solid form.
In the present regulatory framework it seems unlikely that liquid waste
disposal would be permitted in the near term on a large-scale commercial
basis. Direct disposal of liquid wastes has therefore not been evaluated
in this study.
Turning specifically to the question of solidification of HLLW, two
technologies are preferred — calcination and glassification. Calcination
(roasting at high temperature) involves the production of a granular
powder that is stable and less mobile than liquid wastes but still highly
leachable. Extensive practical operating experience has been obtained
with the fluid-bed calciner for defense wastes, but the simpler spray-
calclner seems preferable for commercial wastes. (Because the spray-
calciner product is a fine particulate, Battelle Northwest Laboratories
(BNWL) has pointed out that a fluidized-bed calciner might actually be
preferable if calcine were to be the final product.)
Glassification has a distinct advantage over calcination In that a
vitrified solid with excellent leach resistance Is produced. Glassifi-
cation is accomplished by melting calcined waste together with glassmaking
"frits" to form a relatively homogeneous solid solution of waste products
in glass. The continuous ceramic melter process using resistance heating
is "an attractive option for glassmaking and has the advantage of having
been extensively used in the glassmaking industry. An alternative process,
which has been much more extensively demonstrated with simulated and
actual radioactive wastes, is the in-can melter process, where the glass
is formed by heating in a metal canister. A disadvantage of the in-can
melter is that it subjects the canister material to higher temperatures
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than does the separate, pot-type, continuous melter, and therefore may
require different canister'materials.
For the canister itself, several materials are possible, depending
upon the specific process and geologic medium. Carbon steel is suitable
for the ceramic melter process, but stainless steel or Inconel is prob-
ably required for in-can melting. All of these metals have relatively
high corrosion rates in a hot brine environment. For glass, this is
not considered to be a serious problem because the glass itself, with
its low leach rate, is really the effective containment barrier.* For
calcine, however, more corrosion-resistant metals might be considered,
e.g., titanium, which has excellent corrosion resistance in hot brine
(see Appendix B-?),
Borosilicate glass appears to be the most suitable type of glass
because of its favorable leaching characteristics and because it does
not react with certain ferrous metals as do some other types of glass,
e.g., phosphate glass.
Alternative solidification systems with potential for improved pro-
cess or product characteristics are in various stages of development, but
improvements may not be necessary or even desirable if they lead to more
complex process requirements. The evidence from the technology assess-
ment of already-developed waste solidification processes supports the
conclusion that a selection could be made today from several processes
that would function satisfactorily, assuming competent- design,
installation and operation.
B-2.2.2 Cladding Hulls and Fuel Bundle Residues
Cladding hulls and fuel bundle residues arise during mechanical
chopping and eventual dissolution of fuel bundles or fuel rods at the
initial stages of reprocessing. The dissolution step leaves these
hulls slightly contaminated with fission products and TRU elements.
From the standpoint of waste management, the preferred process is
mechanical separation of relatively uncontaminated fuel-bundle hardware
prior to chopping of the fuel rods and subsequent dissolution. The
undissolved hulls could then be washed, compacted, and packaged in steel
drums. If necessary, although it may not be cost effective, further
decontamination could be accomplished or the hulls could be placed in
some form of solid matrix, e.g., concrete or even glass. There is no
proven technology for such packaging at present.
Selection of the best technology for disposal of cladding hulls
will depend upon the evaluation of more refined guidelines for disposal
*Recent studiesW question the validity of this assumption at higher
temperatures with water present, and may thus impose further restrictions
on the allowable design temperatures within repositories.
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of TRU-contaminated wastes. The current proposed guidelines, which
have a single cut-off point of 10 nanocurles per gram (10"^ Ci/g),do not
recognize the relative risks associated with a wide spectrum of low- to
intermediate-level TRU-contamlnated wastes.
B-2.2.3 Spent Fue1 Element s
Because spent fuel elements contain approximately the'same amount of
fission products as HLW from reprocessed fuel, both waste forms have com-
parable heat generation rates and radioactivity for the first 500 years.
Beyond that point, however, the higher plutonium content of spent fuel
leads to higher radioactivity and greater heat generation.
The simplest disposal method for spent fuel would be containment
in an outer steel canister purged and backfilled with dry nitrogen and
then seal-welded. Depending upon the leach rate for spent fuel elements,
a quantity that cannot be estimated now with any accuracy, It may be
worthwhile to consider a high-integrity containment, such as titanium,
for spent fuel elements. Corrosion rate data for titanium in hot brine
indicate that containment lifetimes of the order of 500-1000 years may
be obtainable.
Some further elaboration on spent fuel leach rates is appropriate.
The principal uncertainty concerns the chemical environment after the
containment and Zircaloy cladding have corroded away. On the one hand,
in an oxidizing environment at 200°C, UC>2 rapidly oxidizes to U^Og;
corresponding physical changes cause the dense ceramic material to
disintegrate, leading to extremely high surface/mass ratios and, hence,
high leach rates. On the other hand in a reducing environment, it is
possible that the UOj could retain its ceramic form; some laboratory
tests show very low leach rates for UO- under these conditions. Much
more data and evaluation are needed to characterize spent fuel leach
rates properly.
B-2.3 SELECTION OF REFERENCE CASES FOR TASK D RISK EVALUATIONS
Given the wide range of possibilities, the varied state of technolo-
gical development, and the differing degrees of effectiveness for the
alternatives, some weeding out of less suitable alternatives is essential,
in order to focus the risk analysis effort (Task D) on the most suitable
techniques.
The approach used for this selection process has been to choose a
set of reference cases that span a reasonable range of possibilities in
terms of (a) waste form, (b) processing (solidification) approach, (c)
packaging (containment) method, and (d) isolation effectiveness. In
selecting reference cases, an attempt has been made to span a wide enough
range of possibilities to permit interpolation and/or extrapolation to
-------
other combinations of variables not specifically included in the
reference cases.
Additional criteria used in selection of reference technologies for
treatment and disposal of HLLW are as follows t
1. The reference case waste characteristics should span the
reasonable range of possiblities available in the near term
(within ten years).
2. The reference case technologies should be available for full-
scale deployment for treatment of reprocessing wastes.
3. Reference cases are based on U.S. technology, since there are
several processes for solidification available in the United
States.
Choice of the following three reference waste forms for HLW seems
appropriate: calcine, borosilicate glass, and spent fuel. (These span
a reasonable range of possibilities in terms of likely waste forms and
leach resistance.) For containment, calcine wastes are assumed to be
placed in carbon steel or titanium canisters, to bracket the range of
containment effectiveness. For glass, a stainless steel canister is
assumed (to accommodate in—can melting process temperatures). Higher-
integrity containment (e.g., titanium) was not considered because of
the inherently high leach resistance of glass. Spent fuel is assumed to
be placed in either carbon steel or titanium canisters with dry nitrogen
gas backfill. Because spent fuel could be as leachable as calcine, ther.e
two canister types provide a further test of the effect of canister .
corrosion rates on transport of radioactive material out of the repository,
For cladding hulls and fuel bundle residues, the reference case is
based on available technology: mechanical removal of fuel bundle hard-
ware from clad fuel rods, compaction of chopped hulls, and packaging in
carbon steel canisters.
Reference cases have also been selected for other waste forms; these
will be described in detail in the body of the report. Table B-l
- summarizes all the reference cases for various waste forms.
B-2.4 ANALYSIS OF ALTERNATIVE GEOLOGIC DISPOSAL TECHNIQUES
Disposal of radioactive wastes in deep, stable geologic formations
has long been considered the preferred method for isolation of these
wastes from contact with man's environment. A number of possible geologic
media have been considered for such disposal, including salt beds, salt
domes, crystalline rock forms such as granite or basalt, shales, limestones,
certain types of clay beds, and others. To date, salt deposits have
received the most attention, especially in the United States, because
of their demonstrated natural stability over extremely long time periods.
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TABLE B-l
REFERENCE CASES FOR VARIOUS WASTE FORMS
Source of
Waste
Converted
Waste Form
Assumed Mode
of Disposal
Canister Material
Notes
HLLW
calcine deep geologic burial
borosili- deep geologic burial
cate glass
Spent Fuel none
Cladding Hulls* none
Krypton-85
Iodine-129
Carbon-14
Tritium
none
none
none
deep geologic burial
deep geologic burial
storage at reprocessing
plant
deep geologic burial
deep geologic burial
release
(a) carbon steel
(b) titanium
stainless steel
(a) carbon steel
(b) titanium
carbon steel
steel (pressure
cylinders)
carbon steel
carbon steel
spray calcination process
spray calcination process
in-can^melt process
dry nitrogen backfill
dry nitrogen backfill
compacted to 1/3 original
volume
stored for *v»100 years for
decay
as silver zeolite and
mercuric iodate
as calcium carbonate
For this study fuel bundle hardware is assumed to be separated from cladding hulls and treated as low-level waste.
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Moreover, salt deposits are homogeneous and are capable of plastic
flow. The self-healing characteristics of salt effectively eliminate
the possibility of extensive cracking, thereby preventing the opening
up of pathways to the environment. For salt to remain intact, it roust
continue to remain dry, and thus must be protected from uncontrolled
drilling activities.
An alternative to salt is a stable crystalline rock such as basalt
or granite. There are abundant examples of such rock of suitable depth
and-age with demonstrated seismic stability. Crystalline rock does not
have the self-healing characteristics of salt, but possesses other
advantages, e.g., resistance to water intrusion, that make it .a desirable
medium for geologic disposal of radioactive waste.
In addition to salt and crystalline rock, other geologic media have
been considered, e.g., shales and clay deposits. In general, these have
both desirable and undesirable characteristics, depending upon the
specific medium and type of waste. For example, the laminar structure
of horizontally-bedded shale reduces water permeability in the normal
(vertical) direction but the presence of water in shale could lead to
high stresses and possible disintegration if the shale ,is subjected to
high temperatures. In a similar vein, certain types of. clay-till deposits
have the advantage of low water permeability, but the disadvantage of
indeterminate long-term stability characteristics.
In light of the foregoing considerations, the following cases
have been selected as reference cases for the purposes of this study;
Case 1. Disposal in bedded salt.
Case 2. Disposal in granite.
Case 3. Disposal in salt domes.
A uniform depth of 460 meters (1500 ft) has been assumed for all
three cases.
Until recently, nearly all of the work performed on engineering
design of disposal facilities has been for salt deposits; only limited
information is available for media other than salt. Therefore, the
major focus of this report is on salt deposits.
The design of waste disposal facilities generally follows con-
ventional mining practices, with two important exceptions: (1) special
provisions are required for safe transport and cooling of radioactive
material; and (2) waste canister spacing and arrangement must be in
accordance with thermal design criteria for the specific medium. The
thermal criteria are basic to the repository design.
The radioactive wastes release heat to the repository at a rate
that decreases with time. In the earlier years, heat is n >t transferred
out of the repository as fast as it is generated and the stored thermal
energy causes repository temperatures to rise. As temperatures rise,
11
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heat begins to flow to the surface, but It Is a few hundred years before
the heat flow from the surface to the atmosphere significantly influences
the temperatures in a deep repository. Ultimately, as heat generation
rates decline, heat losses to the surface exceed heat generation rates
and repository temperatures go down. Elevated temperatures are present
for about 1000-10,000 years, depending upon the medium, the repository
depth and the overall heat generation rates; peak temperatures are
reached in 100-500 years for a 460-m (1500-ft) deep repository. Basically,
the problem can be simply summarized as a heat storage phenomenon
concurrent with slow transfer of heat to the surface, and the temperature
and temperature gradients must be limited to values that do not have an
adverse impact on the integrity of the repository. The actual limits
depend upon the type of geologic medium, e.g., salt versus granite.
The only way to limit aggregate (as opposed to localized) repository
temperatures is to control the amount of heat produced per unit of
horizontal area over an extended time period, since most of the heat
transfer must occur in the vertical direction through a thickness that is
large compared with that of the repository itself. The waste heat output
is described per unit time and per unit horizontal area at some fixed
point in time; it is called "planar heat density" in this report. The
usual units are in kilowatts per acre and, unless otherwise rioted, all
values in this report are for ten-year-old wastes. (Historically, values
of about 150 kW/acre have been used in the engineering designs considered
for salt repositories at 305-m (1000-ft) depth.) Clearly, the planar
heat density is significant, since the size of the mined area for HLW
disposal is inversely proportional to planar heat density.
The allowable planar heat density may depend upon such considerations
as allowable temperatures within the waste canisters, in the salt or
rock adjacent to the canisters, throughout the mined area, in neighboring
aquifers, on the surface, and in the ground outside the repository
boundaries. These factors may also limit the allowable vertical uplift
arising from thermal expansion of the entire repository. These thermal
considerations are discussed in detail within this report; in conjunction
with other engineering data on waste canister dimensions, shielding
requirements, etc., they are sufficient to permit preliminary design of
repositories in salt deposits. Considerable conceptual design work has
already been performed under the direction of ERDA's Office of Waste
Isolation for disposal of HLW in bedded salt. The reference facility
design, reviewed in detail herein, is assumed to be located in a salt
bed at 460-m depth and with a. thickness of 50-100 m. The mined area
is roughly a square 3000 m on a side, covering an area of about 2000
acres. It is large enough to store all wastes produced over the lifetime
of 140 GW of nuclear plant capacity. For an assumed addition of 700 GW
to total Light Water Reactor (LWR) capacity (see Task A report), five
such facilities would eventually be required, but these need not
necessarily be on separate sites.
12
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A detailed description of the facility design will not be given
here, but some of its important physical characteristics will be noted.
About 50% of the area is for low- and Intermediate-level TRU wastes and
the remainder Is for HLW canisters. A system of tunnels is used to
ensure adequate support for overburden and to maintain the proper spacing
of Individual cylindrical HLW canisters, each 30,5 cm in diameter by
about 3 m high. Canisters are transported in shielded casks and/or
shielded transport vehicles during handling. Holes are drilled in the
tunnel floors, waste canisters are inserted, and the holes are backfilled
with crushed salt. The assumed reference planar heat density (over the
HLW field) is 126 kW/acre for ten-year-old waste. Additional details on
the preliminary mine design and proposed operating procedures for HLW
disposal may be found In the body of this report.
No detailed engineering work has been reported on facilities for
spent fuel disposal. Because of higher long-term heat generation rates,
disposal of spent fuel may require lower planar heat densities, roughly
40-501 of the value for HLW. The existing conceptual design for an HLW
facility has been modified to accommodate spent fuel. Since the energy
output of a spent fuel bundle in an Individual canister is about l/9th
that of an HLW canister, if the planar heat density must be reduced by
a factor of 2-2.5, then the number of spent fuel canisters per unit area
should be roughly four times that for HLW canisters. This study has
assumed that four spent fuel canisters would be buried in one hole, with
the holes on approximately the same grid spacing as for HLW canisters.
This would result In a 150% increase in the size of the mined area, but
this increase is partially offset by a substantial reduction in the area
required for TRU wastes. The net effect is an increase of about 75%
(from 2000 to 3500 acres) for the mined area.
The preceding discussion has focused on disposal in salt deposits.
Because of the lack of detailed engineering data, it has not been possible
to make a thorough assessment of disposal technology in granite or shale.
A thermal analysis of granite repositories (presented herein) coupled
with limited engineering studies supported by Atomic Energy of Canada
Limited (AECL), suggest that granite repositories are feasible, with
lower planar heat densities than in salt. Changes in canister size
and/or heat loading could also be required for granite repositories.
The need for these changes depends on the limits for cracking and/or
spalling (from thermal stresses) and for vertical uplift (from thermal
expansion), neither of which have yet been developed. Based on the AECL
work, it has been assumed that a granite repository would be roughly
twice the size of a comparable repository in bedded salt.
There appear to be Insufficient data to predict the feasibility of
repositories in rock other than granite.
13
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B-2.5 EFFECTIVENESS OF ENGINEERING CONTROLS
The technology for design of geologic disposal facilities has been
summarized in the preceding section. This section presents a review of
available data on effectiveness of the three major barriers to transport
radioactive waste from the repository: canister integrity, matrix leach
resistance, and solubility characteristics of leached waste forms.
B-2.5.1 Canister Integrity
We assume corrosion to be much more of a threat to canister integrity
than stress caused by rock or salt movement in the short term. Canister
lifetimes can range from as little as several years to more than 1000
years, depending upon canister material and repository characteristics.
Long-term canister integrity may be important only for waste forms with
relatively high leach rates, i.e., calcine and, possibly, spent fuel
elements. The use of high-integrity canisters with expected lifetimes
up to 1000 years appears to be feasible through use of thicker sections
of conventional materials (carbon/stainless steels) or, in salt, through
use of a material, perhaps titanium, with a low corrosion rate. The
technology is available for fabrication of the required canister types,
but use of titanium would substantially increase canister cc'ts.
B-2.5.2 Matrix Leach Resistance
Three types of matrices were considered—glass, calcine, and spent
fuel material.
(a) Glass — Although many years of laboratory work on glass
leaching show it to be a very slow phenomenon, the theoretical models
and physical data are as yet conflicting, imprecise, and scanty. The
effects of temperature, flow, pH, and salinity have not even begun to be
studied in an organized way for the spectrum of suitable glasses.
Moreover, the extrapolation of short-term laboratory tests to very long
time periods is made difficult by the limited amount of long-term
leaching data. It is possible that short-term leach rate data substan-
tially overstate actual long-term leach rates.
On the other hand, there are also concerns about the stability of a
glass matrix over the very long term. There is considerable evidence
that certain types of man-made glass last for several thousand years and
some natural glasses are much older. One obvious difference between
these glasses and HLW glass is the presence of radiation effects in the
latter. Although accelerated life testing has indicated no significant
devitrification, concern about radiation damage persists, since glass
lifetimes beyond 10,000 years probably cannot ever be demonstrated
conclusively, even with accelerated life testing. Recent work at
Pennsylvania State University'D has indicated rapid rates of glass
14
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alteration at hydrothermal conditions (300°C, 300 atmospheres) that appear
to be more severe than would be likely in a properly designed repository.
These findings have not yet been applied in the design of an entire
repository system.
Given the uncertainties in the physical data, along with the
continuing and essentially unanswerable concerns over long-term deteriora-
tion, it appears unreasonable to do more than assume a conservatively
high value for glass leach rates--expressed in terms of the bulk glass
corrosion rate (g of glass per cm -day). A value of 10"^ g/cm^-day has
been chosen for this study because it is at least a factor of 7 and as
much as a factor of 100 larger than any reported data at the one-year
point. By assuming no decrease with time, further conservatism has been
introduced, which should probably more than compensate for any uncertain-
ties concerning long-term deterioration, e.g., devitrification. At the
same time, some experimental data suggest that much smaller leach rates
may occur in practice. These might be in the range of 10~° g/cm^-day or
even less. Accordingly, a leach rate of 10~" g/cm^-day has been adopted
as a reasonably conservative long-term average value, and W~° g/cm -day
as a possibly achievable long-term average value, assuming that the severe
conditions of the Pennsylvania State University work are not approached.
To determine the total release rate of radioactive material, glass
leach rates (g/cm -day) must be combined with surface-to-mass ratios
(cm^/g) to give fractional release per day. An initial surface-to-mass
ratio of about 0,4 cm /g has been assumed, consistent with test data
on actual canisters after normal cooling transients and additional break-
age from normal handling.
The resulting initial release rates (% of initial mass) range from
roughly 0.01%/year down to 0.0001%/year, depending upon the assumed value
of the bulk glass leach rate constant. For the high leach rate, one-half
the glass is dissolved after about 4000 years (with total dissolution
after 20,000 years), while the corresponding figures for the low leach
rate are 400,000 years and 2 million years.
(b) Calcine—Calcines are relatively soluble and have high surface-
to-mass ratios; hence, full dissolution is expected to occur in less
than one year. Such high release rates are of special concern during
the first few hundred years because of undecayed fission product radio-
activity. In this time period, the canister, not the matrix, would be
the primary barrier.
(c) Spent Fuel—As already noted in Section B-2.2.3, leach rates for
spent fuel elements are highly speculative. In particular, the leach rate
for the TJ02 fuel material can vary over a wide range, depending upon
temperature, groundwater chemistry, and oxygen content. The range of
possible release rates spans several orders of magnitude, from a value
as high as 10 g/crn^-day to a value of about 10 g/cm -day. Consider-
ably more effort is required to reduce the uncertainty band in spent-fuel
15
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leach rates, especially with respect to the probability of the presence
of oxygen-rich groundwater which would cause rapid deterioration of the
VOn ceramic material.
Given the large uncertainty in spent-fuel leach rates, the simplify-
ing assumption has been made that spent fuel could have one of two
possible total release rates: (1) a value equal to that of calcine, and
(2) a value equal to that of high-leach-rate glass.
B-2,5.3 Solubility Characteristics of Leached Waste Forms
In cases where water flow through the repository is low, relatively
high concentrations of certain isotopes in the leachate may be calculated
from glass dissolution rates, A correction has been supplied to allow
for the limited solubility of the actinides in the leachate. In low-flow
situations this solubility may limit the rate at which actinides leave
the repository.
B-2.6 COST CONSIDERATIONS
Data have been developed to indicate comparative costs for the various
reference cases chosen; these data will be needed in order to evaluate
subsequently the cost effectiveness of alternative disposal techniques.
Since only limited amounts of cost data have been published, and
because the scope of this study did not call for making independent
cost estimates, the resulting cost data should be considered only as
rough estimates. Nevertheless, they provide a useful perspective on the
relative costs of various disposal techniques.
By far the largest single contribution to waste-handling costs
is that of interim spent-fuel storage, which, at $80-$150/kg HM*, can
amount to roughly 8-15% of total fuel cycle costs. Other disposal costs
for solidified HLW are estimated to range from $30-$70/kg HM, and for
spent fuel are estimated to range from $40-$75/kg HM, because of higher
shipping, canister, and burial costs.
Differences between calcine and glass solidification costs appear
small in comparison with the other uncertainties in waste disposal
costs.
*HM = "heavy metal" (U and Pu) fed to reactors.
16
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B-3.0 ANALYSIS OF TECHNOLOGY ;. <•• . .
B-3.1 GENERAL
This section of the report deals with the various technological as-
pects of engineering controls to reduce the probability of release of
radioactivity to the environment. These aspects include (1) the form of
waste material, and (2) the design of the geologic disposal facility.
Various technologies are evaluated for encapsulating and containing dif-
ferent types of high-level radioactive wastes. For geologic disposal,
primary emphasis is on bedded salt, since this medium is the only one for
which extensive information is available; a review of available data on
other geologic media is also presented, in order to indicate some of the
difficulties arising from insufficient data. In general, the available
data do not permit detailed evaluation of the effectiveness of engineering
controls for media other than salt.
Before proceeding to the detailed technology assessment, another
general point should be noted concerning the assumed "age" of waste mate-
rial at the time of geologic disposal. In the past, it has generally
been assumed that fuel would be reprocessed relatively quickly (within
6-12 months of discharge) and that wastes would be solidified 1-5 years
after reprocessing. In light of today's indefinite moratorium on re-
processing, however, it now appears that wastes will "age" as spent fuel
for ten years or more until reprocessing is initiated or a decision is
made to "throw away" the spent fuel by burial in a geologic repository.
In either case, a substantial amount of storage capacity for spent fuel
is required.
Having made such an investment in storage capacity, the industry
will probably use the storage facilities for a long time. Moreover,
substantial economies may be effected in the design of reprocessing plants
and/or waste processing and disposal facilities with such extended storage.
These economies result from the reduced heat generation and radioactivity
of the aged material after radioactive decay. As already shown in the
Task A report, heat generation and radioactivity decrease by roughly a
factor of 3.4 between 2-year and 10-year decay times and by an additional
factor of 1.6 between 10-year and 30-year decay times. Krypton-85 and
tritium levels fall by factors of two or more during such storage, as well.
Perhaps even more important than the potential cost savings are the
simplifications that can be made in waste processing if only aged wastes
are processed (those at least ten years old). The problem of heat transfer
out of vitrified matrices is greatly reduced, eliminating the need for elabo-
rate provisions for heat transfer, such as the composite-metal matrix. Con-
cerns over glass devitrification are also reduced because of low glass
temperatures. Finally, the entire handling sequence can be made simpler,
and therefore more reliable, because of reduced radioactivity and lower
heat-generation rates.
17
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In view of these advantages, the reference cases are based on the
assumption that only aged wastes would be processed and buried in a
geologic disposal facility.
B-3.2 ALTEBNATIVE DISPOSAL FORMS FOR HIGH-LEVEL WASTES
This section of the report describes the programs and processes used
to prepare (solidify) and package high-level wastes for permanent dis-
posal in a Federal repository. The following sections cover high-level
liquid wastes (HLLW), cladding hulls, fuel bundle residues, and spent
fuel elements. In this section, repeated reference is made to EKDA 76-
43^', currently the most comprehensive document on the subject. Other
sources are also used, but the exhaustive, thorough nature of ERDA 76-43
makes it an indispensable component of the discussion that follows.
B-3.2.1 High-Level Liquid Wastes
High-level liquid wastes (HLLW) are generated by a spent fuel repro-
cessing plant. These wastes are the highly radioactive solutions that
remain after recovery of uranium and plutonium for reuse in the nuclear
fuel cycle. Reprocessing of spent fuels from the commercial nuclear'
power industry is not being done in the United States at present, but a
substantial amount of the HLLW from past reprocessing of defense materials
is on hand. The Task A report discusses the present and probable future
amounts and character of HLLW. Present regulations require solidifica-
tion of these wastes within five years after reprocessing, and transfer
of the solids to a Federal repository within ten years after reprocessing.
As discussed elsewhere in this report, there may be economic and other
advantages to delaying the reprocessing of spent fuel until at least ten
years have elapsed since the discharge of the fuel from the reactor.
This would substantially reduce the radioactivity (and therefore the heat
removal problem) of the wastes.
Before solidification, the high-level liquid waste from commercial
sources, generated at a rate of about 5000 liters/metric ton of uranium
(MTU) reprocessed, will be concentrated to a liquid volume of about 380
liters per MTU.(2) This concentration would be achieved by evaporation,
the purpose of which is to concentrate the HLLW and recycle water and
nitric acid. The heat generated by the HLLW is a function of reactor
specific power level, burnup, and time after discharge. Fuel irradiated
at a power level of 38.4 MW /MTU to a burnup of 33,000 MWt-days/MTU is
considered typical of the LWR fuel that will be reprocessed in the United
States in the next two or three decades. The rate of decay heat
generated in such a fuel or in the HLLW made therefrom would;be as
follows:
18
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Approximate decay heat rate
T±me_ out of reactor kW/MTUcharged to reactor
160 days 24
1 year 12
10 years 1,1
Extensive research and development has taken place In the United
States and throughout^the world on possible alternatives for treatment
of nuclear high-level wastes. Practical application of some of these
techniques on a commercial or near-commercial scale has been successfully
demonstrated. The results and status of these projects will be
summarized in this section. It should be borne in mind that, since the
wastes arising from commercial reprocessing will be "younger" and more
concentrated than those from the nation's defense waste programs, the
measures discussed in this report should be more than adequate for
disposal of defense wastes from which the sodium salts have been removed.
The possibilities for some form of direct disposal of liquid wastes
have been studied extensively. In the present regulatory framework,
however, it appears unlikely that such techniques would be seriously
considered for commercial application. Handling of liquid wastes
presents safety problems for transportation if the waste disposal area
is off the reprocessing site. Moreover, it leaves the waste in a more
readily dispersible form in the near term, when the hazard is greater.
Although direct disposal of liquids may eventually prove to be acceptable
under some circumstances, especially if transportation is not involved,
it has not been considered in this study.
Although waste solidification appears desirable, this solid waste
may take a number of forms. In early practical demonstrations liquid
waste was converted to a calcine by roasting at high temperatures in
an oxidizing environment. Lately, work has centered on incorporation
of the waste into a glass, which has substantially improved characteristics
of leach rate, thermal conductivity, resistance to dispersal during trans-
portation or handling accidents, and fire resistance. Most major programs
in other countries (France, England, USSR, Germany, India, and Italy)
favor glass for the final form prior to disposal.
Several research projects are studying waste forms even more durable
than glass, e.g., glass-ceramics. Better-than-glass forms may be beyond
the point of diminishing returns on improved product form, however^ '
Presently both calcine and glass are considered to be acceptable forms
of solidification of high-level liquid wastes.' ' Typical waste calcine
characteristics from acid HLLW, such as from commercial reprocessing,
are shown in Table B-2;^' Typical waste-glass properties from similar
HLLW are shown in Table B-3>4'
19
-------
TABLE B-2
TYPICAL WASTE CALCINE CHARACTERISTICS
Appearance
3
Bulk density g/em
Leach rate at 25°C
distilled water, g/cm -day
Thermal conductivity, W/»°C
Nitrate content, % by wt
Water content, % by wt
Thermal and radiation
stability
Friable granules, scale or powder
0.7 to 2.0
ID"1
0.2 to 0.4
0.05 to 10
0,01 to 0.2
Requires post-calcination treatment
to remove nitrate and water (at
9008C)
TABLE B-3
TYPICAL WASTE GLASS CHARACTERISTICS
Appearance
Bulk density, g/cm
Leach rate at 25°C,
distilled water, g/cm -day
Thermal conductivity, W/m°C
Formation temperature, "C
Maximum storage tempera-
ture, °C
Black, shiny with conchoidal fractures
3.2 to 3.4
10~8 to 10"6
0.8 to 1.4
1050 to 1150
<300 (adjusted for 10-year-old wastes)
20
-------
The specific values of the calcine and glass characteristics are
dependent on the process, the operating parameters employed, and the
waste compositions^ Therefore, the characteristics given tend to
vary somewhat from reference to reference, but the above values appear
to be typical. It should be noted that the calcines have relatively
low thermal conductivity and high teachability; they contain residual
water and nitrate, but in commercial calcining processes, these are
typically baked out at about 900°C.
In the conversion of calcine to glass, the addition of glass-forming
additives increases the weight of the resultant product by a factor of
2.5-3.5, The density also increases, resulting in a net product volume
increase of only about 50%. As Tables B-2 and B-3 show, the leach rate
for glass is much lower than for calcine. The thermal conductivity
is increased by a factor of 3-5 (allowing the use of larger-diameter
waste canisters, assuming similar centerline temperature limits for
glass and calcine products). With its lower leach rate and higher
thermal conductivity, glass is therefore generally considered to be a
better waste form.
The solidified waste, either a calcine or a glass, will probably be
contained in a metal canister; metals considered for this application
are; carbon steel, stainless steel, Inconel, and titanium. Choice of
canister type will be discussed in more detail in Section B-3.2.3.
The remainder of this section, B-3.2.1, covers the solidification of
HLLW to calcines and to glass. The calcination and glassification pro-
grams that are in the most advanced stage of development, i.e., most
nearly re^ady for commercial application, will be described first.
Discussion of several processes that are in earlier stages of development
will^follow.
The emphasis is on the conversion of HLLW to solid forms suitable
for final disposal. Other associated operations, such as off-gas treat-
ment and inspection and handling of waste canisters, are covered in sep-
arate sections.
B-3.2.1.1 Solidification Technology Under Development
For the reader not interested in process detail, Tables B-4 and B-5
summarize the status of the principal U.S. and foreign solidification
programs in an advanced stage of development. Each of the programs
is discussed at more length in the subsequent sections. The characteris-
tics of the products of all these solidification processes are shown in
Table B-6.
Processes that are still in an early stage of development are dis-
cussed in Appendix B-II.
21
-------
TABLE B-4
PROCESSING/TECHNOLOGY SUMMARY. DEVELOPED U.S. SOLIDIFICATION PROCESSES
Fluidized
Inert Bed
OPERATING
TEMPERATURE
500-800*
HEAT CONTROL LEVEL OF DEMONSTRATION
In-Bed INEL application at 400-500
Combustion liters/hr (military wastes).
(kerosene, 02) BNWL inert bed tests (non-
radioactive): 38 runs 350 hr,
up to 40 liters/hr (25 liters/
hr nominal).
Spray
Calcination
700-800*
(wall)
E!.Pir.t:ric
Resistance
Furnace
In-Can
Melting
1000-1100
Zone
Furnace
Development since 1959. 13
fully-radioactive engineering
scale runs (WSEP), solidifying
8700 liters in 600 hrs. Current
WFP: development unit 44 runs
875 hrs (non-radioactive); full
scale unit 3 runs at 210 liters/
hr (non-radioactive).
7.5 kg/hr radioactive; 50 kg/
hr non-radioactive; 22 runs
2500 kg
COMMENTS AND POTENTIAL
OR UNRESOLVED PROBLEMS
INEL process difficult to
adapt to commercial wastes.
Tests show inert material
precludes bed melt in event
of bed collapse. Sodium
limit cf 1.2 M without addi-
tives. No data on ruthenium.
No nozzles. Conceptually
more complex than spray cal-
ciner.
1-2% of ruthenium offgassed,
sodium retention up to 2 M.-
Requires vibrator to reduce
wall scale. Early diffi-
culties with spray nozzles
appear to be corrected. No
on-line commercial experi-
ence.
Canisters subjected to high
temperatures; possible mate-
rials limits. Possible diffi-
culties measuring tempera-
tures/levels in canisters
(new canister used for each
melt.) Temperature restraints
may result in crystalline
phases in glass.
*Followed by 900°C bakeout of water and nitrates.
-------
TABLE B-4
(continued)
PROCESSING/TECHNOLOGY SUMMARY. DEVELOPED U.S. SOLIDIFICATION PROCESSES
Continuous
Ceramic
Melter
OPERATING
TEMPERATURE
1100-1200
HEAT CONTROL
Electrodes;
Startup
Using NaOH
LEVEL OF DEMONSTRATION
Non-radioactive: 23 months,
4000 kg, 80 liters/hr. Tilt-
pour demonstrated.
COMMENTS AND POTENTIAL
OR UNRESOLVED PROBLEMS
Direct feed of liquid waste
may be possible; transfer
of molten glass requires
more equipment and higher
risk of maintenance. Not
tested radioactive.
-------
TABLE B-5
PROCESSING/TECHNOLOGY SUMMARY, DEVELOPED FOREIGN SOLIDIFICATION PROCESSES
Operating
Temperature
Heat Control
Level of Demonstration
Comments & Potential
or Unresolved Problems
Si
*>
Foreign
French
Rotary
Kiln
Calciner
French
Rotary
Kiln with
Continuous
Melter
500
1150
(melter)
Zoned 5000 hrs pilot and engineering
external scale (non-radioactive). No
resistance radioactive experience; startup
furnace of commercial unit scheduled
this year.
Induction (see item above for calciner).
heater Tests of PIVER process (similar)
(melter) for 4 yrs (radioactive), 164 runs
non-radioactive pilot tests of new
process.
Potential maintenance problems
due to rotating machinery
(bearings and seals). No radio-
active tests.
(see also item above). Freeze
valve for liquid transfer may
create reliability problem
(plugging or unexpected valve
opening); PIVER had frequent
vessel failures.
German
Spray
Calciner with
Continuous
Melter
British
Rising
Level
Glass
(HARVEST)
600
(calciner)
1150-1200
(melter)
1000-1050
Superheated
steam or
wall heater
for calciner,
induction or
joule heater
for melter.
Multi-zone
furnace
Pilot scale (non-radioactive)
1000 hrs, 1500 kg.
Similar FINGAL process tested in
8 radioactive production runs,
64 non-radioactive runs. 13.8
liters/hr non-radioactive HARVEST
test.
Front-end denitration minimizes
production of non-condensable
gases. Use of superheated steam
complicated. Freeze valve
potential maintenance problem;
spray nozzle used (potential
plugging).
An in-can process, and discussior
of in-can melting in Table B-4
applies. Eliminates calciner.
-------
TABLE B-6
SOLIDIFICATION PRODUCT CHARACTERISTICS
PROCESS
A.
B.
PRODUCT
DEVELOPED SOLIDIFICATION PROGRAMS IN U
Fluidized Bed
Spray Calcination
In-Can Melting
Continuous Ceramic
Melter
Calcine
Calcine
Borosilicate
Glass
Borosilicate
Glass
DEVELOPED SOLIDIFICATION PROCESSES IN
French
Rotary Kiln Calciner
French Rotary Kiln
Calcine
Borosilicate
PARTICLE BULK SURFACE TO
DIAMETER DENSITY MASS RATIO
ym g/cm cmr/g
.S.
100-300 2-2.4 103-10A
,2-5 0.5-1.3 ^105
3.0-3.6 0.05-0.5
3.0-3.6 0.05-0.5
FOREIGN COUNTRIES
2-5 1-1.3 ^105
3.0-3.6 0.05-0.5
LEACHABILITY
g/cm^-day
KICK
KICK
LOW
LOW
HIGH
io-7
THERMAL
CONDUCTIVITY
W/m°C
0.18-0.33
0.18-0.33
0.9-1.3
0.9-1.3
0.18-0.33
0.9-1.3
Continuous Metallic
or Ceramic Melter
Glass
German Spray Calciner Borosilicate
Continuous Melter
British Rising
Level Glass
Glass
Borosilicate
Glass
2.9-3.1
3.0-3.6
0.05-0.5
0.05-0.05
-6
10
LOW
1.2-1.3
1.2-1.3
C. U.S. PROCESSES IN EARLY STAGES OF DEVELOPMENT
Supercalcine
Crystalline
Ceramic
LOW
-------
TABLE__B-6
^continued)
SOLIDIFICATION PRODUCT CHARACTERISTICS
ro
ON
PROCESS
PRODUCT
PARTICLE
DIAMETER
Vim
BULK
DENSITY
g/cm
C. U.S. PROCESSES IN EARLY STAGES OFDEVELOPMENT (continued)
2.4-3.3
Sintering
Metal Matrices
Glass-Ceramic
Coated-Pellet
Ion Exchange
Glass-
Crystalline
Phase
Waste
Particles
in Metal
Crystal Plus
Residual Glass
Phase
Coated Oxides
in Metal
Matrix
Ceramic
2.85
SURFACE TO
MASS RATIO
LEACHABILITY
g/cm^-day
-8
more than
4.5
about 10
Depends on
waste form
in metal
LOW
VERY LOW
LOW
THERMAL
CONDUCTIVITY
W/m°C
about 0.7
5-35
2.2
Depends on
Metal Selected
-------
TABLE B-6
(continued)
SOLIDIFICATION PRODUCT CHARACTERISTICS
PARTICLE BULK SURFACE TO THERMAL
DIAMETER DENSITY MASS RATIO LEACHABILITY CONDUCTIVITY
PROCESS PRODUCT ym g/Cm3 cm2/g, g/cm2-day w/m°C
D. FOREIGN PROCESSES IN EARLY STAGES OF DEVELOPMENT (continued)
Julich Borosilicate Borosilicate 2.9-3.1 3.6 x 10~5 0.9-1.3
Glass
Pamela Phosphate 5 x 10~ 24-35
Glass Granules
in Lead Alloy
ro
^ Thermite Ceramic-Metal
When no number appears, data cire either not available or not applicable. In general, a "high"
leachability means >10 g/cm-day and a "low" leachability means <10 g/cm2-day.
-------
B-3.2.1.1.1 Calcination
B-3.2,1.1.1.1 Fluidized-Bed Calcination
Fluidized-bed calcination has the longest history of operation of
any technology for radioactive waste solidification in the United States.
It was developed at the Idaho National Engineering Laboratory (INEL). The
first engineering-scale facility for the solidification of liquid
radioactive wastes was the Waste Calcining Facility (WCF), which began
calcinating wastes from the Idaho Chemical Processing Plant (ICPP) in
1963, The fluidized-bed calcination process was originally developed
for the acidic aluminum nitrate solutions arising from the processing
of uranium-aluminum alloy fuels and was intended to be only a demonstra-
tion project. The WCF plant, however, became an integral part of the
waste-management program at ICPP and it has been used to solidify
aluminum nitrate, zirconium fluoride, and stainless steel sulfate waste
solutions.^) Fluidized-bed calcination has operated successfully at ICPP
for about 14 years, resulting in calcination of 10.2 million liters
(2.7 million gallons) of waste at production rates up to 500 liters/hr.
Calcining by the fluidized-bed method is accomplished by, pneumati-
cally atomizing the waste solution into a bed of fluidized solid granules
at a temperature of 500°C. Heat is supplied by in-bed combustion of
kerosene with oxygen. See Figure B-l for a schematic representation of
such a unit.
The original facility used a liquid-metal (NaK) heat-transfer system
that worked adequately for the first three processing campaigns during
which the system operated at temperatures greater than 600 C for 35,000
hours* / The plant capacity was then increased by providing a more
thermally-efficient in-bed combustion system. The in-bed couibi-stion
system also has the apparent advantage of reducing the amount of ruthen-
ium that leaves the calciner when processing zirconium wastes containing
fluorides. Possibly unburned hydrocarbons or carbon monoxide reduce the
volatilized ruthenium to a non-volatile lower valence state. See
Figure B-2 for a schematic diagram of the off-gas handling system.
The fluidized-bed calciner converts liquid waste to a mixture of
powdery solids and granules ranging from 0.1-0.3 mm in size. Density
varies from 2.0-2.4 g/cnr and the specific volume is about 40 liters/MTU
for commercial reprocessing si- .". as that originally planned for the Allied-
Gulf Nuclear Services Plant at Barnwell, SC (AGNS), The in-bed combustion
temperature is 500-800°C, and the calcine product can be stabilized (de-
nitrated and dehydrated) at about 900°C for storage, or the calcine can
be fed directly from the calciner to a glass melter. Thermal conductivity
of the product varies from 0.18 W/m°C at 40°C to 0.33 W/m°C at 600°C.
The capacity of a fluidized-bed waste calciner is dependent on the
process conditions chosen and on the particular waste being solidified.
Data obtained at the WCF indicate that a 75-cm diameter calciner could
accommodate the feed from a 5 MTU/day plant. The ICPP fluidized-
28
-------
BAFFLE
HLLW, PLUS ADDITIVES
IF REQUIRED
*• OFF-GAS
CYCLONE
OXYGEN
KEROSENE'
FLUIDIZINGAIR
:i>!
*
&
^HL
J§P
CANISTER FURNACE
FORNITRATEBAKEOUT
«90d°C)
STORAGE CANISTER
FIGURE B-1 FLUiDIZED-BED CALCINATION
Source; Alternatives for Managing Wastes from Reactors and
Post-Fission Operations in the LWR Fuel Cycle.
ERDA 76-43, U.S. Energy Research and Development
Administration, May 1976.
29
-------
UJ
o
FlUIOIZING AIR
PREMEATER
QUENCH TOWER
CTCLONE FINES
SEPARATOR
.STl CALCINE*
ASTE rttO
TOMIIIN8 AIR ^3
FUEL
oxrcEN "5
v/
T>
S
\.
h
I
/T
V
SCRU8BII.S SO
SURGE T
T
^
I SOLIDS STORAGE CYCLONE
1 AIR HEATER
..BLOWER
SOL'C, STORAGE
SINS
FIGURE B-2 GENERAL CALCINATION FLOWSHEET FOR THE NWCF
Source: Anderson, F.H., et al. Design Criteria for the New Waste
Calcining Facility at the Idaho Chemical Processing Plant.
In: Management of Radioactive Wastes from the Nuclear
Fuel Cycle. Proceeding Series STI/PUB/433, International
Atomic Energy Agency, 1977.
-------
bed calciner solidified relatively low-activity wastes, which were of a
substantially different chemical composition than the high-activity
fission-product nitrates expected from commercial reprocessing. Pilot-
scale testing with simulated (non-radioactive) commercial HLLW feed
compositions has been conducted to demonstrate the feasibility of
fluidized-bed calcination with commercial wastes. Some difficulty was
experienced in the INEL commercial pilot plant studies with agglomera-
tion of sodium at higher than 0.4 M concentration' ' (sodium nitrate
exists in a molten, undecomposed state over the temperature range 350-
833°C). It is not clear that higher sodium concentrations would nec-
essarily be present in commercial fuel, however, depending on the design
of the reprocessing plant. (Sodium is, of course, present in neutralized
wastes.) Commercialization tests at INEL showed that addition of
powdered iron inhibited agglomeration at 1.13 M sodium. Calcination
tests of dilute HLLW revealed some problems in scale-up to comtnercial-
size 30. 5- cm (12-in) calciners. The INEL defense waste calcination pro-
gram experienced plugging of off-gas filters with ruthenium compounds;
this did not seem to happen in the commercial program, but the reasons
are not clear. The conclusions of the INEL commercialization studies
were that fluidized bed calcination could be used for commercial
application, but that not all the questions of sodium content, or of
off -gas filter plugging, appeared to be fully answeredM'
The maximum credible accident for a fluidized-bed calcination system
has been postulated to be a collapsed bed. If no corrective action were
taken, a collapsed bed of one-year-out-of -reactor calcine in a 75-cm dia-
meter vessel would result in a molten core surrounded by sintered calcine.
In such a situation, the vessel wall could dissipate the decay heat by
convection and radiation at wall temperatures well below the melting
point of stainless- steel. Nevertheless it appears that the effects
of a collapsed bed may be reduced by processing ten-year-old waste and
by installing redundant air— supply and bed-discharge systems. The
major objectives of the ongoing development program
1. Verification of the calcinability of simulated Allied General
Nuclear Services (AGNS) waste at steady-state conditions in
a large-scale pilot plant unit.
2. Verification of the ability of the off-gas cleanup system to
remove particulate and volatile fission products and noxious
chemicals to acceptable levels.
3. Verification on a suitable scale of the ability to address
major safety questions, e.g., removal of a collapsed
configuration.
4. Design of a practical calcine storage system.
5. Evaluation of the need for further equipment development.
31
-------
6. Prototype equipment and remote-maintenance mockup testing
as necessary to minimize the risk of significant process
or equipment problems during plant operation.
Completion of the development and prototype testing programs is
scheduled for 1978. It is expected, based on non-radioactive testing,
that all of the data for detailed design would be available in 1977.
B-3.2.1.1.1.2 Continuous Inert-Bed Calcination
If the calcine is to be used for glassmaking, large amounts of
silica must be added to it. If, Instead, this silica is added to the
fluidized-bed calclner, better control of a more stable "inert bed" is
achieved, producing a calcine already pre-mixed with silica.
After the INEL studies of commercialization of fluidized-bed calcin-
ation, BNWL continued to pursue commercialization studies of fluidized-
bed calcination, using an inert-bed process variation. BNWL's process
injects an inert material (silica) into the fluidized bed, thereby simpli-
fying calciner operation and allowing higher sodium concentrations in the
feed material (up to 1.2 M Na, compared with a limit of 0.4 M Na In the
INEL tests). A schematic diagram of such a calciner is shown in Figure
B-3. (8)
Agglomeration may be avoided by continuously sweeping "the bed with
fresh bed material. Since the coated material is continuously withdrawn
and the calcine powder is continuously swept from the bed, the concentra-
tion of calcine in the bed may be kept low. This reduces the possibility
of bed collapse and the need for emergency drains or bed coolers (for the
case of loss of fluidizin? air).
The inert-bed concept has been shown to be usable over wide operating
ranges. Simulated non-radioactive HLLW has been calcined at feed
rates up to 40 liters/hr.
BNWL may recommend addition of the glass-forming materials to the
liquid waste before it is introduced to the calciner. In this way a
single feed inlet would suffice. The one experimental run made using
this concept was successful. BNWL is also considering the use of a sand
filter instead of sintered metal filters because there is some concern
that volatilized materials may plate out on the sintered metal and plug
the openings. The sand in a sand filter would be continually changed,
with the discharged sand forming the inert bed feed to the calciner.
Design of a scaled-up fluidiEed-bed calciner has been initiated.
Startup and testing is planned for late 1977. Major development objec-
tives include verification of long-term HLLW operation, pilot-scale
radioactive testing, and full-scale, remote demonstrations. It is hoped
that radioactive runs can be started in 1979. BNWL is quite optimistic
about the commercial feasibility of such a process, but prefers the spray
calciner for near-term commercialization.
32
-------
Offgas
Vapors and Fines
Inerts Addition
/ Bed Section
.Waste
itemizing Air
CFuel
Oxygen
Fluidizing Air
Calcine and Inerts
FIGURE B-3 CONTINUOUS INERT BED FLUIDIZED-BED CALCINER
Source: Management of Radioactive Wastes from the Nuclear
Fuel Cycle. Proceeding Series STI/PUB/433, Inter-
national Atomic Energy Agency, 1977.
33
-------
B-3.2.1»1.1.3 Spray Calcination
Spray calcination does not have as extensive an operating history
as fluidized-bed calcination, but it is conceptually simpler and equally
compatible with a downstream vitrification system. It has been under
development at BNWL for over 15 years. Early spray-calcination programs
uncovered difficulties with spray nozzle performance, development of
wall scale, and carryover of fines into the off-gas. In current demon-
strations these problems appear to have been corrected.
Despite the'lack of commercial-scale operating history, spray cal-
cination has been tested in 13 fully-radioactive engineering-scale runs
during the Waste Solidification Engineering Prototypes (WSEP) program, '
According to BNWL, the WSEP program resulted in a total of 600 operating
hours and solidified 8700 liters (2300 gallons) of waste. The present
BHWL waste fixation program has produced a development spray calciner
(44 non-radioactive runs, 875 hours at 70 liters/hr feed) and a full-
scale spray calciner (3 non-radioactive runs, 210 liters/hr feed).
Actually, the "full-scale" spray calciner is larger than would be re-
quired for a 5 MTU/day plant (estimated at 100 liters/hr feed). The
spray calciner presently appears to have a significant edge over fluidized-
bed calcination as a choice for commercialization.
The full—scale development unit is shown in Figure B-4. An active
pilot plant will be placed in operation in 1978. During late 1979, test-
ing of mockup equipment will be used to verify remote design features.
Development should be complete by the end of 1980.'-^
The HLLW at about 40°C is pumped to the calciner through a pneumatic
spray nozzle. The waste solution spray, about 70 ym diameter droplets,
is evaporated and converted into oxides and reaction gases. The final
product is very fine (2-5 ym), typically contains less than 0.5% by weight
moisture and less than 1% by weight nitrate, and has a density of 0.5-1,3
g/cm . The calciner walls are controlled at a temperature of 600-800°C by
an external multizone resistance furnace for experimental purposes, but a
single-zone heater is expected to suffice for a commercial unit. Deposits
on heat transfer surfaces within the calciner are removed by the periodic
operation of side-mounted vibrators.
The gases released are primarily H.,0, HNO^, air and NO , with small
amounts of CO , H,,, 0,, and N?. The off-gas at about 350°C is passed
through sintered stainless steel filters to remove entrained calcine
particles. Less than 0,1% by weight of calcine passes through the filters
and less than 21 of the ruthenium is estimate-* to escape to the off-gas
system. The balance of the radionuclides, except for volatiles such as
iodine, remain in the calcine. The filters are periodically reverse-
pulsed with air to prevent the accumulation of an impervious calcine
cake. Off-gas is routed to effluent treatment for radionuclide and NO
removal.
34
-------
Furnace
\
Vibrator
Atomizing Air
HLLW Feed In
'
' \
14% ft [77
y. tf...-!
s
s
s
s
V
s
s
s,
s
V
^
s
s
s
s
s
s
s
\
s
s
-
u
1
s *
s >
s -
S s
s >.
S s
s ,
Filter Blowback Nozzles
JLJ
Rfll
"\f\
u.
i
3"1 Vapors Out
s
Sintered SS
Filters
Insulation
Powder Out
FIGURE B-4 SPRAY CALCINER FOR 5 MT/DAY REPROCESSING FACILITY
Source: McElroy, J.L. et al, Waste-Solidification
Technology U.S.A. In': Proceedings of the
International Symposium on the Management of
Waste from the LWR Fuel Cycle.
Denver, Colo., July 1976. CONF 76-0701.
-------
BNWL favors spray calcination over other calcination processes for
near-term commercialization, citing these advantages:
* One moving part (associated with the vibrator);
* Overall simplicity resulting in fast startup and shutdown;
* Calcination of a wide range of waste compositions with sodium
content to over 2 M;
* Negligible inventory of calcine entrained in the unit;
• Demonstrated low release of radionuclides;
* Production of a fine calcine well-suited to glassmaking (but
perhaps not as desirable as a final waste form).
The major development objectives if calcine is to be the final
product would be to .increase bulk density and thermal conductivity of
the calcine product."'
B-3.2.1.1.1.4 Rotary Kiln Calcination
The French program- to solidify HLLW dates from 1959. The early work
was centered mostly on a pot vitrification process (called "FIVER").
Since batch processes have limited capacity, it was decided to develop a
continuous rotary kiln calciner coupled to a continuous melter. After
5000 hours of testing of the rotary kiln calcination process, a full-size
industrial rotary kiln was manufactured in 1971. Testing on this unit
started in 1972 and is still continuing.' '
Figure B-5 is a schematic diagram of a rotary kiln calciner. It
consists of an externally-heated (500°C) rotating cylinder about 3,5m
long by 27 cm diameter, operating at a slight slope. Deacidified HLLW
is dried and almost completely denitrated before it is discharged. A
loose bar within the kiln keeps the calcine free-flowing and prevents
it from sticking to the wall. The calcine product is further heat-treated
in a canister furnace at 900°C to assure total decomposition of the
nitrate.
Air pressure in the kiln is maintained slightly negative relative to
the ambient pressure. The kiln is heated externally by a four-zone
electric resistance furnace. Average processing time for conversion of
HLLW to calcine is about 4 minutes. '
Because the French have always fed the calcine directly to a vitri-
fication unit, no actual data on the calcine are available. It has a
bulk density estimated at 1.0-1.3 g/cm .
36
-------
HLLW,
PLUS ADDITIVES
IF REQUIRED
DRIVE GEAR
30rpm
GRAPHITE SEAL
GRAPHITE SEAL / LOOSE BAR TO
BREAK UP SOLIDS
4-ZONE
RESISTANCE FURNACE
CANISTER FURNACE
FOR NITRATE
BAKEOUT
<90Q°O
STORAGE CANISTER
FIGURE B-B ROTARY Ki LN CALCINATION
Source; Alternatives for Managing Wastes from Reactors
and Post-Fission Operations In the LWR Fuel
Cycle. ERDA 76-43, U.S. Energy Research and
Development Administration, May 1976.
37
-------
This process has been under development in France for more than ten
years in connection with a continuous Vitrification process. During much
of this time an engineering-scale unit has been in use with simulated non-
radioactive waste, A radioactive engineering-scale rotary calciner is
scheduled to be in operation in 1977 as part of a vitrification facility.
Much of the technology developed from this installation would be appli-
cable to a calcine product facility, were it desired to stop at the cal-
cine state. In that case, the major development objectives for a facility
with calcine as a final product would be to increase the bulk density and
the thermal conductivity of the calcine.^ ^
B-3.2.1.1.2 Classification
Borosilicate glass is the preferred form in most glass programs
(U.S., U.K., France, India, and Germany). It has relatively low tempera-
tures of formation, low corrosiveness to container material during for-
mation, and low leach rates., Borosilicate glass typically has the pro-
perties shown in Table B-7.
The manufacture of glass in a radioactive facility is complicated
by the difficulties of remotely operating and maintaining equipment;
special attention must be given to questions of reliability. In this
regard, simple, rugged equipment may have substantial advantages over
more complex (and more sensitive) equipment. In addition, since the
long-term stability of the radioactive glass product cannot be measured
directly, special problems arise in the design and operation of process
control instrumentation in order to insure uniform product quality.
B-3.2.1.1.2.1 In-Can Melting
In the in-can melting process, the calcine produced from HLLW is
dropped into a storage canister along with frit material, and is melted
in the storage canister, using a multizone furnace below the calciner.
Off-gas from the melting mixture is vented to the calciner off-gas system,
Because heat is generated by the waste, the temperature below the melt
level would tend to rise as the canister fills; the lower zones of the
furnace, therefore, are turned off as the melt level rises above them
and cooling is initiated. After the canister is filled, the calcine and
frit are diverted to another canister in a parallel furnace.
A schematic diagram of the in-can melter coupled to a spray calciner
is shown in Figure B—6.
After cooling, the canister is capped and removed from the furnace.
The cap is seal-welded and the canister is checked for leaks, and decon-
taminated prior to being sent to storage.
38
-------
• ' TABLE B-7 * • • •: • • •
CHARACTERISTICS OF BOROSILICATE GLASS
CoTOposition
S102 25-40 wt%
B203 10-15 wt%
Alkali Metal Oxides 5-10 wtZ
ZnO 0-20 wtZ
Waste Oxides 20-35 wtX
3
Density 3.0-3.6 g/c»
Theijnal Conductivity 0.9-1.3 W/(a-°C)
ProcessingTemperature 1000-1400°C
39
-------
HLLW
CALCINER
FURNACE
*-ATOMIZING AIR
OFF-GAS
PROCESS ING AND
FINAL STORAGE
CANISTER
SINTERED STAINLESS STEEL FILTERS
(BLOWN BACK PERIODICALLY)
ZONED CANISTER
FURNACE
FIGURE B-6 IN-CAN MELTING
Source: Alternatives for Managing Wastes from
Reactors and Post-Fission Operations in
the LWR Fuel Cycle, ERDA 76-43, U.S.
Energy Research and Development Adminis-
tration, May 1976.
40
-------
The in-can melting product monolith is described as follows; "Be-
cause the melting is done in a metallic storage canister, processing
temperature restraints prevent complete assimilation of all waste oxides
into the glass matrix. The result is a glass matrix in which small amounts
of several crystalline phases are dispersed. Cerium dioxide is the most
prevalent crystalline phase present. The leachability of the glass is
not significantly increased by the crystalline phases."(2)
On the formerly prevalent assumption that wastes might be relatively
fresh, several techniques have been tried to increase the heat removal
rate within the glass. Installation of internal longitudinal fins in
the canister increases thermal conduction, Increases processing capacity,
and approximately doubles the amount of heat-producing radioisotopes
allowable in the canister.
According to BNWL, the advantages of the in-can melting process are
that itW):
• Minimizes process steps and equipment;
• Does not require transfer of melt;
• Assures that everything entering the melter, with the exception
of some volatile species, is fixed in the storage canister;
• Eliminates the problem of melter deterioration and disposal;
* Is not affected by the addition of reducing agents for phase
separation control; and
* Has been demonstrated both in the United States and abroad.
BNWL has conducted radioactive tests of in-can melting at 7.5 kg
glass/hr.^ ' The on-going waste fixation program at BNWL has produced
2500 kg of melt in 22 engineering-scale runs (non-radioactive) at melt-
ing rates up to 50 kg/hr in 30-cm (12-in) diameter cans.
To insure continuous smooth operation of the in-can melt, the process
requires (1) temperature controls (zoned-furnace and controlled-heating/
cooling requirements complicated by canister sensitization temperature
limits and internal heat generation), and (2) level instrumentation (level
is an important input to the temperature controls). There nonetheless
seems to be little doubt as to the commercial practicability of in-can
melting as an option, and the product is well defined.
Major development objectives in the current in-can melting program
include final specification of canister design; optimization of process
procedures, using a multiaone furnace; and full-scale remote demonstra-
tions. <2)
41
-------
B-3.2.1.1.2.2 Continuous Ceramic Melter
This process (also called "joule heating"), may ultimately prove
superior to in-can melting; it has been commercially utilized (for 30
years) in the glass industry. It is carried out in a ceramic-lined
melter with internal electrodes. In such a unit, molten glass acts as its
own electrical resistance heating element. This type of melter is popular
in the glass industry because of its high capacity and.,long life while
producing glass of superior quality with low off-gas effluents.*2) These
same characteristics are desirable for an HLLW vitrification system. The
process has been developed at BNWL; the French and the Danes have also
explored joule heating melters. The work at BNWL started in 1973 and is
described in ERDA 76-43 P) Following laboratory-scale tests in 1974,
an engineering prototype was built and kept at 100Q-13QQ°C continuously
for 11 months. Intermittent operation during this period produced a
total of 4 MT of glass at rates up to 60 kg/hr. During this time only
minor corrosion of the refractories and electrodes was noted. The unit
was restarted in January 1976 and has accumulated over 23 months of
exposure to molten glass without failure (including the time prior to
restart).
Figure B-7 is a schematic diagram of a joule-heated continuous
ceramic-line melter. Sufficient heat can be transferred to allow drying
of the waste as well as melting, possibly eliminating the need for a sep-
arate calciner. In this case, liquid waste would be fed directly to the
melter.
To start operation of such a unit, the glass must be brought to a
temperature at which it becomes adequately conductive. Sacrificial heat-
ing elements were initially used, but once the glass became molten, the
sacrificial element was^dissolved by the glass. BNWL has had success
with a recently-developed startup/restart technique in which sodium hy-
droxide solution is used as the initial carrier of the electric current.
This eliminates the need for temporary electrodes and resistance elements'
For initial startup, the melter is charged with a 20-cm (8-in) depth of
glass frit. NaOH solution is then added to conduct the electricity until
the glass starts to melt. The NaOH also acts as a flux, which causes the
glass to melt at a lower temperature.
Once startup is complete, calcine and frit may be added as molten
glass is drawn off. With calcined waste and frit fed to the melter,
capacities higher than 20 kg of glass/hr/100 cm of surface area have
been demonstrated.^ ' The BNWL melter has been operated at rates as
high as 85 liters/hr.
A tilt—pour mechanism has been devised to eliminate the need for a
freeze valve (a potential source of failure).
42
-------
HU.W
ADDITIVES
OFF-GAS
COOLING
PLATE
PIVOT
•-»
WIPED FILM
EVAPORATOR
CONTAINER SHELL
STORAGE
CANISTER
BOTTOM DRAIN SCREW JACK
f OR SHUT DOWN FOR ON-OFF
ONLY) DRAIN CONTROL
FIGURE B-7 PROCESS SUMMARY - CONTINUOUS CERAMIC WELTER
Sourcej Alternatives for Managing Wastes from
Reactors and Post-Fission Operations in
the LWR Fuel Cycle, ERDA 76-43, U.S.
Energy Research and Development Adminis-
tration, May 1976.
43
-------
Some of the advantages of the ceramic melter are;
• Long (demonstrated) melter life,
* Flexibility in waste input composition,
• High capacity per hot-cell area,
* Ability to produce molten glass for monoliths, marbles, coated
pellets, etc.
The engineering-scale melter has been shown capable of recovering
from electric power outages (more than 25 min), sudden additions of
liquid solution, periodic under- and over—powering, and the addition of
metal.^°' The process must now be demonstrated with extended remote
operation using radioactive wastes for feed.
A more advanced version of this process involves direct liquid
feeding of the HLLW to the ceramic melter. This has been tested at BNWL,
and a capacity of more than 6 liters/hr of HLLW/1000 car surface area
has been demonstrated.' ' If direct liquid feeding is employed, the waste
is transferred to a mix tank, mixed with frit, and the slurried waste is
then fed by gravity into the melting cavity; there, heat from the glass
pool evaporates the liquid to steam (the safety of this step will have
to be demonstrated) and decomposes the metal nitrates to oxides that dis-
solve in the glass. Even if the molten surface is completely covered
with 20-40 mm of solution, no loss in processing rate is noted and less
than 0.5% of the waste is entrained in the off-gas stream. This approach
is also being pursued at BNWL.
B-3.2.1.1.2.3 French Rotary Kiln - Continuous Metallic or Ceramic Melter
The French rotary kiln calciner described under Section B-3.2.1.1.1.4
has always been closely coupled to a vitrification furnace. Figure B-8
is a schematic diagram of the kiln coupled to an induction heater to con-
vert the calcine into a borosilicate glass.
The Inconel, induction-heated, vitrification furnace is connected
directly to the rotary kiln calciner by an expansion bellows. Glass frit
and calcine from the kiln are continuously added to the furnace (1 m
high x 35 cm diameter), where they are melted.
The level of molten glass in the 'furnace is allowed to rise to a
preset value, and the glass then flows through a freeze valve into a
product container. Two vitrification units, each of 20 kg/hr product
capacity, are connected to each kiln.^ ' The off-gas from the vitrifi-
cation units is routed through the rotary kiln and becomes part of the
kiln off-gas.
44
-------
OFF -GAS *
DENITRATEO HLLW —-
DRIVE GEAR v
3A rnm
x"
B SLOPE »
5
AVWl AAAV1 /WVVI/WW
/GRA
v x
GRAPHITE SEAL / LOOSE BAR
TO BREAK UP SOLIDS
4 ZONE
RESISTANCE FURNACE
GLASS FRIT
INDUCTION HEATED
CONTINUOUS MELTER
(1150°O
INDUCTION HEATED
"FREEZE DRAIN VALVE
STORAGE CANISTER
FICURE B-8 FRENCH ROTARY KILN - CONTINUOUS METALLIC MELTER
Source '• Alternatives for Managing Wastes from Reactors and
Fost-Flsslon Operations in the LWR Fuel Cycle.
ERDA 76-43, U.S. Energy Research and Development
Administration, May 1976.
45
-------
The final disposal canisters are exposed to somewhat lower tempera-
tures than with in-can melting, but there are also the disadvantages of
potential rapid melter deterioration (discovered during the FIVER pro-
gram, when vessel replacement was required every 30 cycles), and the need
for a freeze valve (malfunction or plugging of which shuts down the en-
tire operation). Nonetheless, the French had sufficient success with
this arrangement to incorporate a melter vessel into the new "AVM" facil-
ity at Marcoule, which will be capable of handling the entire HLLW output
from the reprocessing plant.
The major development objective for the near future is to test the
system under high-level radioactive conditions.
B-3.2.1.1.2.4 German Spray Calciner-Continuous Melter
Research and development on the glassification of radioactive wastes
has been under way in Germany for about ten years. The "VERA" process
being developed in Karlsruhe is the most advanced of the German programs.
Figure B-9 is a schematic diagram of the original basic equipment con-
figuration.
Initially the waste solution is denitrated by reaction with formic
acid. (The volatilization of ruthenium is thus reduced to 0.01% of the
total ruthenium content.) The waste is then pumped to the top center of
the spray calciner where it is steam-atomized, then dried and calcined
in flight by super-heated (650°C) steam. The calcine is removed by 3-
micron sintered metal filters, and is blown bacK periodically to fall
into the melter. Addition of SiCU to the HLLW seems to reduce filter
clogging.
In the induction-heated Inconel melter, the calcine and the glass
frit are heated to about 1150°C for about two hours. Molten glass is
drained from the melter via a freeze valve into the final storage canis-
ter (stainless steel).
A later concept combined the denitrator, calciner, and filter into
a single integrated unit, reducing the size of the unit itself and of
the hot cell.^ ' This combination unit has successfully processed at
least 3000 liters of simulated HLLW so far. A flowsheet is shown in
Figure B-10. Complete condensation of the heating steam and recycling
of the condensate decreases the radioactivity of the off-gases by several
orders of magnitude. ^
A still later variation of the VERA process consists of a modified
calciner and ceramic melter integrated into a single unit. The simulated
high-level waste, mixed with glass frit, is sprayed directly into the
wall—heated calciner/melting unit where the droplets are dried and the
product is melted to borosilicate glass at 1200°C. In this version the
heating steam circuit is eliminated. A large portion of calcining heat
is supplied by radiation from the ceramic melter directly below the cal-
ciner.
46
-------
Steam Recycled from Evaporator
Denitrated HLLW
No Calciner Furnace
Heat Supplied by
Superheated Steam
lntroduced@600°C
Off-Gas
Sintered Stainless Steel Filter$
(Blown Back Periodically)
CT"
Induction Heated
Continuous Melter
(1150°O
Induction Heated
Freeze Drain Valve
Storage Canister
FIGURE B-9 ORIGINAL VERA PROCESS
Source: Alternatives for Managing Wastes from Reactors and
Post-Fission Operations in the LWR Fuel Cycle.
ERDA 76-43, U.S. Energy Research and Development
Administration, May 1976.
47
-------
-p-
m
FIGURE B-10 VERA PROCESS SYSTEMS
Source: Management of Radioactive Wastes from the Nuclear Fuel Cycle.
Proceeding Series STI/PUB/433, International Atomic Energy
Agency, 1977.
-------
The pilot-scale melter with a design capacity of 30 liters of waste
solution per hour started operation in September 1976. In the first long-
range test of 100 hours a total of 6600 liters of simulated high-level
waste solution was converted to 1500 kg of glass that was filled into
25 steel cylinders,^5'
Work with the ceramic melter is expected to continue, culminating
in a radioactive prototype installation in 1983 or 1984. (")
Compared with the U.S. spray calciner design, the German VERA unit
has the advantage of elimination of HNO^ and reduction of volatile radio-
activity In the off-gas by injection of formic acid. On the other hand,
the German design requires a more complicated atomizing noz2le design,
which increases potential for clogging from waste dried by the steam
heat before passing through the nozzle opening.
B-3.2.1.1.2.5 British Rising Level Glass Process (FINGAL/HARVEST)
Development of a rising level vitrification process called FINGAL
was begun in the late 1950fs at Harwell, England.^ ^ This process
fed HLLW, together with an aqueous suspension/solution of silica and
sodium tetraborate, Into a stainless steel cylinder at 1050°C In a multi-
zone resistance heated furnace (see Figure B-ll). Evaporation, denitra-
tion, sintering, and glass formation proceeded simultaneously during the
filling cycle. The feed rate was kept constant and the free liquid level
rose at a rate equal to the rate at which glass was produced. The off-
gases from the first, or glassmaking,vessel were passed through the
second and third vessels, which contained filters to trap particulate
material and volatile ruthenium. When the first vessel was filled with
glass, it was removed to storage and the vessels in the other positions
were progressively moved into the furnace. The filter In each cylinder
was Incorporated into the glass melt as vitrification proceeded in the
furnace.
The FINGAL process was shut down in early 1966 after completing 72
runs, of which 8 involved processing radioactive wastes from the Wind-
scale works. In 1972, it was decided to resume development of a solidi-
fication process that would be suitable for full-scale industrial use.
The modified process is called HARVEST.
The HARVEST concept differs from the FINGAL program in batch size,
off-gas system, and geometry of the container vessel. The maximum batch
size for the FINGAL plant is about 80 kg of glass; the HARVEST batch size
Is intended to be 1-3 MT. A HARVEST pilot plant has been constructed at
Harwell. Figure B-12 is a flow diagram of the new pilot plant.
The main advantages of the rising level glass process are the elimin-
ation of a separate calciner and use of the process vessel as the final
storage canister.
49
-------
FISSION PRODUCT WASTE
SILICA /BORAX SLURRT
SPENT FILTER
1050%
250°C
250°C
HNO, COM3ENSATE
FIGURE B-11 SIMPLIFIED FLOWSHEET OF FINGAL PROCESS
Spurce.: Techniques for the Solidification of High-
Level Wastes. Technical Report Series 176,
International Atomic Energy Agency, 1977.
50
-------
To building vent
FIGURE B-12 NEW HARVEST PILOT PLANT FLOW DIAGRAM
Source; Management of Radioactive Wastes from the Nuclear Fuel Cycle. Proceeding Series
STI/OUB/433, International Atomic Energy Agency, 1976.
-------
B-3,2.1.2 Solidification Processes in Early Stages of Development
Several processes under development were considered too preliminary
for consideration at this time. They are:
1. Supercalcine
2. Sintering
3. Metal Matrices
4. Glass-ceramics
5. Coated Pellets
6. Ion Exchange
7. German Thermite Process
8. German (Julich) Borosilicate Glass Process
9. German (Pamela) Phosphate Glass Process
These processes are discussed in Appendix B-II.
B-3.2.2
Current plans for commercial fuel reprocessing include an acid leach
of fuel bundles which have first been sheared into short lengths. The
solid residue from the leach consists of fuel cladding, fuel bundle support
rods, poison rods, end fittings, fuel support grids, springs, and spacers,
This hardware weighs about 325 kg/MTU. As recovered from the acid dissolver,
the hardware residues are a relatively low-density (about 1 kg/liter)
waste. The composition of these residues will vary as fuel manufacturers
change their materials of construction. The current composition is
mainly Zircaloy, with lesser amounts of stainless steel, Inconel, and
other materials. The cladding, after being acid-leached, contains about
0.1% of the actinide content of the original fuel; fission product content
of the cladding is assumed to be higher (0.2%) to allow for the effect of
fission fragment recoil into the cladding. For actinides, ERDA 76-43^ '
gives an estimated range of 0.05-0.5%; the AGNS Final Safety Analysis
Report (FSAR)^1^ and ORNlX13) settle on 0.1% as the most likely value.
NFS experience was that 0.05% of the fuel activity was contained in the
hulls.*14*
Table B— 8 gives expected properties of a fuel bundle waste stream
for an LWR fuel (assumed to be PWR fuel) .
If the cladding is assumed to be compacted to 701 of its theoretical
the final volume would be about 60 liters/MTU. (1-*) Th±s
compares with 60-80 liters/MTU of waste glass produced from the HLLW. * '
Thus, the volume of hulls and fuel bundle residues would, if thoroughly
compacted, approximately equal the volume of high-level glass from the
same fuel.
52
-------
TABLE B-8
FUEL BUNDLE PROPERTIES
Fuel Assembly Parameters
Total assembly weight 666.8 kg
Zr-4 fuel cladding/assembly 110.5 kg
Zr-4 end plugs/assembly 3.4 kg
Zr-4 in control rod and Instrument 17.9 kg
tubes/assembly
Total Zr-4/assembly 131.8 kg
Total effective Zr-4/assembly^ 122.8 kg
Total Inconel 718/assembly 6.2 kg
Total 304SS/assembly 5.3 kg
Total effective 304SS/assembly^ 2.0 kg
Fuel weight/assembly, heavy metal 523.5 kg
oxide
Fuel weight/assembly, heavy metal content 461.4 kg
Weight of total assembly structure^ '/ 0.3107
weight of fuel, kg structure/kg
heavy metal
Overall assembly length 4.1 m
Overall assembly length with rod cluster 4.3 m
control assembly in place
Fuel rod length 3,9 m
Active fuel length 3.7 m
Assembly cross section .21 x .21 m
Specific power 38.4 kW/kgHM
*Weights corrected so that effective weight reflects the average
axial flux. This is the value used in the ORIGEN analysis for
structure activation. Correction based upon Westinghouse data
for the axial flux profile for a 3.7-m active core.
All non-fuel assembly components.
53
-------
The expected radioactivity and heat generation of accumulated cladding
hulls has been discussed in the Task A report. Total volume through the
year 2000 is on the order of 5000 ra3.
The only commercial disposal of cladding hulls and fuel bundle resi-
dues to date has been done by Nuclear Fuels Services, Inc. (NFS). In
this case, the fuel assembly nozzles or end boxes were removed from the
fuel bundle prior to chopping. These end boxes, containing no appreciable
uranium or plutonium, were packed into drums separately. The cladding
wastes, including hulls cut into short sections, fines, and grids were
processed through the leaching cycle, monitored for undissolved fuel,
and dumped into 114-liter (30-gallon) steel scrap drums. The drums of
hulls and the drums of nozzles or end boxes were placed in burial holes
at an on-site facility designated for this purpose, and backfilled with
at least 1.3 m (4 ft) of earth cover over the top drum. NFS observed
clad waste quantities (uncompacted) of 350 liters/MTU. ^^
The AGNS FSAR indicates much the same procedure, except that it wasn.,,.
not planned to separate the end fittings from the hulls before leaching.^ '
After leaching, the dissolver baskets were to have been monitored and then
the hulls dumped into a hull disposal container. Since Zircaloy fines can
be pyrophoric under certain conditions, provision was to have been made to
add sand to the disposal container if the hulls should ignite. The
container was to have a capacity of 1470 liters (0.9 m i.d. x 2.24 m high)
which is sufficient for the waste components from 3 MTU (3 dissolver baskets),
equating to about 490 liters/MTU. The AGNS container shell was to have
10-cm (4-in) thick sidewalls and a 15-cm (6-in) thick bottom of integrally
cast reinforced concrete with anchor bolts cast into the sides for securing
the cover, fabricated from 3.2-cm (1,25-in) thick carbon steel. A rubber-
covered asbestos gasket was to provide the seal between the cover and
container. The exterior surface of the container was to have been covered
with an epoxy coating to aid in decontamination with a high-pressure hose.
The hull container was to have been placed in a 1.52-m (5-ft) diameter,
4-m (13-ft) deep hole in the ground. The hole was to have been backfilled
and compacted in order to leave a minimum earth cover of 1.22 IP (4 ft)
above the top of the cask. The area over the buried casks was to have
been mounded to shed rain water.
Despite these plans for surface burial, it now appears that because
of their TRU contamination (see Task A report), cladding
wastes will probably be handled by geologic disposal in a manner
similar to that of high-level wastes. At 0.1% of fuel plutonium and fission
product content, fresh cladding wastes are contaminated with approximately
3 kg/m actinides, activity levels of 2 x 10 Ci/nH, and heat rates of
approximately 1 kW/m .'l^) It will be shown in Section B-4.7 that costs
for disposal of low-level TRU wastes in deep geologic formations will
exceed $860/m . Cladding disposal costs should be above this figure,
a substantial economic incentive for volume reduction, and a number of
possible treatment steps may be warranted or required.
54
-------
Figure B-13 shows various possibilities for chop-leach fuel bundle
residue treatment. There are many alternative choices among such variables
as mixing agents, melting techniques, and conversion processes. Prelim-
inary sorting of residues for size and material, or surface cleaning for
decontamination and oxide removal further extends the treatment
possibilities and the range of waste -product forms. Except for .segregation
and the dissolving of Zircaloy in aqueous fluorides, all the pretreatment
and treatment technologies are in the small pilot plant or Conceptual
stages requiring f urther" development. '
following discussions of cladding-waste technologies are
adapted from ERDA 76-43.
B-3.2.2.1 Pretreatment
Sorting (material segregation) and surface cleaning are the principal
pretreatment options. Sorting fuel cladding by size or material may be
desirable for reasons of safety, simplified processing, and storage.
Sorting can be done before or after the chop-leach process. The recovery
of end fittings, grids, guide tubes, etc., as non-TRU wastes and improved
efficiency of separation by materials are incentives for fuel bundle dis-
assembly and segregation prior to the shearing operation. Because the
fuel assembly parts that have high TRU levels (hulls) are readily
distinguishable mechanically from other parts, this step should be
reasonably inexpensive.
If an overall evaluation of safety and economics indicates a need
to sort the cladding wastes (as appears likely at probable disposal costs
of thousands of dollars/m ), then three size classifications seem desirable.
The massive group would contain end fittings and grid sections; the inter-
mediate group, the hulls and plenum springs; and the third group, the
remaining material. Undissolved fueJ. would be in the intermediate or
fines group. The fines could be rendered safe for handling by casting
in a matrix or by conversion to a more stable compound through controlled
oxidation or chemical change.
The design of the General Electric (GE) fuel processing plant at
Morris, Illinois included provisions for segregation of fuel bundle
residues. Equipment was installed to withdraw fuel rods from the bundle,
prior to running the fuel rods through a shear. After leaching,the hulls
were rinsed to remove fines. This operation was demonstrated on cold
fuel bundles, but was not used on irradiated materials.
Surface treatment is another pretreatment option with possible
advantages. It can either reduce the radioactivity of the major volume
stream or enhance the characteristics of the final waste form. Zircaloy
fuel cladding contains as a residue about 0.1% of the original fuel.
About 50% of this is leachable and 50% is nonleachable and integral
with the corrosion product oxide on the fuel surface clad.^*•' Usually
55
-------
PACKAGE.
INTERIM STORAGE,
SHIPMENT
PACKAGE,
INTERIM STORAGE,
SHIPMENT
FIGURE B-13 DECISION AREAS IN THE MANAGEMENT OF FUEL BUNDLE RESIDUES
Source; Alternatives for Managing Wastes from Reactors and
Post-Fission Operations in the LWR Fuel Cycle.
ERDA 76-43, U.S. Energy Research and Development
Administration, May 1976.
56
-------
less than 0.1% of the TRU element contamination of the cladding is
found In the metal. Thus, surface decontamination factors (DF) of about
10 are attainable by cleaning.the zirconium dioxide product from the
surface. Whether it would prove economically advantageous to provide
surface cleaning, even at these levels of decontamination, is dependent
upon a number of variables, such as regulatory differentiation between
higher and lower orders of TRU contamination, and the extent to which
the decontamination wastes may be reduced In volume and stabilized. Even
a DF of 1(H does not appear attractive economically if the proposed
10 nCi/g rule Is left unmodified, for a typical hull waste contamination
level of 10° nCi/g would still have to be disposed of In geologic formations
after a contamination reduction of 10 . However, if the decontamination
residue can be made considerably more stable than the surface contamina-
tion (as seems likely), then a technical case could certainly be made
for decontamination. In summary, a conflict may exist between the
direction in which the current form of proposed regulations push hull
disposal and implementation of the best technology for treating those
hulls.
B-3.2.2.2 Treatments
B-3.2.2,2.1 Matrix Formation
Incorporation of the hull waste into a matrix can reduce pyrophoriclty,
leaehability, and requirements for shielding. Even if fines are separated
at the reprocessing plant, a matrix binder will reduce the formation
of fines in subsequent handling operations. Some form of precompaction
or flattening would probably be desirable prior to matrix casting to
increase the hulls~to-matrix ratio. Densification and compaction are
discussed briefly later in this report. The matrix technology is
essentially available.^ '
B-3.2.2.2.1.1 Concrete Matrix
Concrete is an obvious matrix material for Immobilizing cladding
hulls, based on limited experience with concrete matrices for low- and
intermediate-level wastes.' ' There has been concern regarding potential
maintenance problems at a disposal facility for concrete-matrix hulls,
but the concept appears sufficiently simple to allow for dealing with
these concerns.
B-3.2.2.2.1.2 Bitumen Matrix
Bitumen has been used in a radioactive waste disposal system for
handling low- and intermediate-level wastes, but it does not appear
to be suitable for highly radioactive solid metal waste materials because
of radlolytic decomposition of the bitumen."' The potential decomposi-
tion problem, together with flammabllity, has discouraged serious con-
sideration of the use of a bituminous matrix for high-level wastes.
57
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B-3V2.2.2.1.3 Glass/Matrix
It is conceivable that HLW glasses could be used as a matrix
material for hulls. In so doing, two waste forms could be combined,
resulting in volume and weight savings. Interactions between glass and
zirconium metal alloys would have to be investigated to determine the
effects of the zirconium on the glass properties.
B-3.2.2.2.1.4 Sand Matrix
AGNS incorporated provision for adding sand to the hulls before
burial, but intended to do so only if the hulls ignited due to
pyrophoricity. NFS, before abandoning plans to restart their reprocessing
plant, had been considering sand as a matrix material for treatment of
cladding wastes before and after storage to prevent fines ignition. Sand
has the advantage of being relatively free-flowing, and the maintenance
problems are probably less than for other matrices. Metal recovery
from a sand matrix, if desired, is also relatively simple.
B-3.2.2.2.2 Densification
B-3.2.2.2.2.1 Mechanical Compaction
•The void fraction of untreated hulls, end adapters, grids, etc.,
is 80-90%.^ ERDA 76-43 further estimates that compaction and flattening
can at least triple the weight of hulls that can be stored in a given
volume. As mentioned earlier in this section, there is an economic
incentive for densification to optimize the handling, transport, and
storage of the fuel bundle residues.
Unirradiated Zircaloy tubing has been compacted, and briquettes
(27.9 cm diameter x 12.7 cm high), weighing 3.6 kg have been produced at
about 72% of the theoretical density of Zircaloy.^ ' Irradiated hulls
may be embrittled and may fall apart when compacted, however.^' Any
fines so produced might increase pyrophoricity.
' A commercially-available compacting unit could handle the hulls
from a 5 MT/day spent fuel reprocessing plant and achieve 70% of the
theoretical density. ^
B-3.2.2.2.2.2 Melting
Densification by melting can reduce the fuel hull waste volume by
a factor of about 6.'^) This represents a final volume about 40% less
than that achieved by compacting to 70% of theoretical density. Zircaloy,
stainless steel, and Inconel are currently produced using well-established
melting technologies, though not in a radioactive environment. The
58
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melting of Irradiated fuel hulls has yet to be demonstrated In quantity.
The cost of melting would probably be higher than that of mechanical
compaction.
B-3.2.2.2.2,3 Cold-Crucible Melting
The chemical activity of zirconium at its melting point of approxi-
mately 1850°C requires that it be melted in vacuum or an inert atmosphere.
Conventional crucible materials react vigorously with molten zirconium.
Consequently, cold-crucible processes were developed for producing
high-purity zirconium.
The Inductoslag process, which was developed by the U.S. Bureau
of Mines in Albany, Oregon, seems the most promising cold-crucible
melting method for densification of fuel bundle residue. This process
consists of induction-melting in a segmented, water-cooled, copper
crucible. The system demands a high-purity slag, such as calcium fluoride,
in an inert-gas atmosphere. The process has been demonstrated for the
melting of simulated hulls. It produces a clean melt with a minimum of
splatter. Disadvantages of the process are a relatively low production
rate and the problems associated with slag handling and recycle.
Most densification studies with the Inductoslag process were con-
cerned with the melting of decontaminated (descaled) fuel hulls poten-
tially suitable for non-TRU waste storage or for reuse in the case of
a segregated Zircaloy product. Melting of uncleaned, undescaled fuel
hardware waste for the purpose of volume reduction alone may be desirable.
Highly oxidized Zircaloy-4 was melted by the Inductoslag process In order
to determine the ability of the process to densify uncleaned hulls.
Melting characteristics were nearly identical with those of clean hulls.
B-3.2.2.2.2,4 Hot-Crucible Melting
Hot-crucible methods include ceramic crucible melting (for stainless
steel or Inconel) and graphite crucible melting (for Zircaloy cladding
or fittings). The major disadvantage of hot-crucible melting is the
limited life of the crucible, necessitating remote replacement of crucibles
and disposal of spent crucibles.
B-3,2.2.2.3 Conversion from Metal Form
B-3.2.2.2.3.1 Dissolution
Concentrated sulfuric and nitric acid mixtures have been used to
dissolve hulls on a laboratory scale. Zircaloy dissolves rapidly, but
dissolution of Inconel or stainless steel is slow. The dissolved TRUs
could be separated and added to the main process stream in a reprocessing
operation.
59
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Similarly, fluoride solutions also attack Zircaloy rapidly, but are
slow in dissolving stainless steel or Inconel. Little work has been done
on the treatment of dissolved Zircaloy or on the separation of the dissolved
TRUs.
B-3.2.2.2.3.2 Volatilization
In this process the chopped and leached fuel hulls are treated
with a chlorinating agent (gaseous HC1 or molten ZnCl^) that disintegrates
the hulls and hardware and converts them to metal chlorides. Since the
zirconium chloride produced "by this reaction is volatile, it can be
separated from most of the other constituents of the primary wastes.
The zirconium chloride could be reacted with steam to form zirconium
oxide, which could then be incorporated in glass.
B-3.2.2.2.3.3 Oxidation with Hydrogen Fluoride-Oxygen Mixture
Fines of zirconium and stainless steel can be oxidized by hot air
or oxygen. Larger pieces of Zircaloy or stainless steel can be burned
in oxygen mixed with hydrogen fluoride gas. (The behavior of Inconel
in hydrogen fluoride-oxygen mixtures at elevated temperatures is not
known.) The resultant oxides would be suitable for incorporation into
a glass.
B-3.2.2.2.4 Further Conversion of Nonmetallic Waste Forms
B-3.2.2.2.4.1 Conversion to Glass
Zirconium and iron-chrorae-nickel alloy oxides are inert and thermo-
dynamically stable and may be suitable forms for long-term storage.
They do, however, have a large surface area that might be attacked by
possible leaching solutions. Conversion of the oxides to a monolithic
glass more resistant to leaching may therefore be desirable, although
it increases waste volume.
B-3.2.2.2.4.2 Conversion to Adsorbents for High-Level Wastes (HLW)
Processes for making HLW adsorbents from zirconium have not "been
extensively investigated, although they are currently being studied
at Sandia Laboratories. The ability to minimize leaching by adsorption
and fixation of HLW ions on hydrous zirconium oxide is presently poorly
defined and requires substantial development before it can be evaluated.
The promise of volume reduction for the total waste is a primary motiva-
ting force for its evaluation. The volume of HLW calcine will be increased
by adding the hull waste, however.
60
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B-3.2.2.3 Evaluation of Alternatives
The alternatives described for the treatment of fuel hulls differ
considerably in their state of development, but in each case the general
character of the treatment product can be established. The operability
of the processes is not so well established, nor are comparative costs.
The available information on process description, status of technology
and product description is summarized in Figure B-14 taken from EKDA
76-43. ' Since demonstration experience and cost data are lacking,
selections at this time -tend to over-emphasize the product form. Never-
theless, the following can be noted from the Figure B-14 summary:
* Melt densification provides maximum volumetric efficiency,
even with an allowance for some secondary waste resulting
from slag recovery or decontamination.
• Zirconium in ingot form is highly resistant to corrosion,
fire and leaching.
• Metal reuse may or may not prove to be economically feasible,
but ingots melted from descaled hulls can be readily
refabricated.
In summary, it appears that sorting to remove non-contaminated
wastes, and decontamination are probably attractive cladding treatments.
Sorting appears to have a strong economic incentive: this is not the
case for decontamination under the proposed 10 nCi/g rule, but the
principle of reducing the levels of TRU in the hulls and concentrating
the residue for ease of handling is technologically sound. Revising
the proposed 10 nCi/g rule might provide an economic reflection of this
technical principle.
¥olume reduction by compaction will probably be an economic advan-
tage, but it does not reduce potential risks from waste disposal.
Further stabilization by incorporation into concrete or glass matrices
may be worthwhile, depending on the degree of integrity that could be
achieved at reasonable cost. These stabilization technologies are not
sufficiently developed at present, however, to permit optimization of
waste form for these cladding wastes.
Although there are technological options for decontamination and/or
increasing the leach resistance of cladding hulls, use of these
options may not be cost-effective, in light of the relatively low
activity of these wastes.
61
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PROCESS
STATUS Of
TECHNOLOGY
ON
TREATMENT ALTERNATIVE
CONTAINED ONLY, NO TREATMENT
MATRIX
MECHANICAL COMPACTION
MECHANICAL COMPACTION AFFER
DECONTAMINATION
DENSIFICATIQN BY MELTING
OENS1FICAT10N BY MILTING WITH
DECONTAMINATION
OENSIFICATION BY MELTING WITH
DECONTAMINATION SEGREGATION
DISSOLUTION IN SULFURIC ACID
DISSOLUTION IN AQUEOUS FLUORIDES
FUEL HULL VOLATILIZATION
BURNING FUEl HULLS TO OXIDES
CONVERSION OF SIDE WASTt STREAMS
PRODUCT DESCRIPTION
/
/////////////////////////
/*/$/ ^ / & / & / $/$/&/ * / <£ / $/£/ COMMENTS
ES
WS
YES
YES
B
B
[
YES
YES
YES
YES
YES
YES
YES
YES
YES
YES
C
B,C
8
B
C
B
B
B
C
B
LOW ] YES
LOW
7
'
7
7
?
?
LOW
7
7
YES
NO
NO
YES
7
NO
NO
YES
NO
NO
NO
P4.ANT
.,
-
—
LAB
LAB
LAB
LAB
PLANT
LAB
LAB
LAB
YES
YES
YES
NO
NO
NO
NO
YES
YES
YES
7
?
1/1
1+/1
10
10 +
1(6
1/6 +
1/6 +
III
III
III
HI
III
7
YES
?
7
YES
YES
YES
YES
YES
YES
YES
YES
7 YES
YES
?
YES
YES
YES
YES
7
?
?
YES
NO
YES
NO
YES
YES
YES
YES
YES
YES
YfS
NO
NO
NO
NO | CONSTITUTES PRESENT PRACTICE
NO
NO
i
NO NO
j
NO YES
WHILE NOT DEMONSTRATED, ADDITION OF CONCRETE OR
LOOSE SAND APPEARS TO BE WELL WITHIN CURRENT
TECHNOLOGY
CONCEPTUAL ONLY - MECHANICAL PROPERTIES OF
IRRADIATED FUEL HULLS LARGELY UNKNOWN
AS ABOVE
MELTING DEVELOPED THROUGH PROOF OF
PRINCIPLE STAGE
7
YES
NO
NO
YES
NO
NO
YES 1 DECONTAMINATION DEVELOPED THROUGH PROOF
YES
NO
NO
NO
NO
NO
OF PRINCIPLE STAGE
SEGREGATION OF FUEL HULL RESIDUE METALS HAS NOT
BEEN DEMONSTRATED
CAN HYDROLYZE TO 0X1 DE AND INCLUDE IN GLASS
CAN HYDROLYZE TO OXIDE AND INCLUDE IN GLASS
CONVERStON TO HI.W ABSORBENT
CAN BE INCLUDED IN GLASS
PROOF Of PRINCIPLE, TECHNIQUES FOR HANDLING SIDE
WASFE STREAMS NOT OEMONSfRATED
FIGURE B-14 TREATMENT OF FUEL HULLS - PROCESS-PRODUCT CHARACTERISTICS AND STATUS
Source: Alternatives for Managing Wastes from Reactors and Post-
Fission Operations in the LWR Fuel Cycle. ERDA 76-43, U.S.
Energy Research and Development Administration, May 1976.
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B-3.2.3 Storage Canisters
B-3.2.3.1 Canister Sizes
Canister sizes vary over a wide range of both diameter and length,
depending on the heat-generation rate and the throughput of the process.
ERDA 76-43 gives the following probable criteria for HLW containers:
* Maximum diameter, 40 - 60 cm
• Maximum length, 3 - 4.5 m
• Maximum linear heat release rate, 2.9 kW/m
3
* Maximum radiation, 1 x 10 retn/hr neutron dose rate and
1 x 10 rem/hr gamma dose rate, measured 1 meter from any
point on the surface of the canister
* Surface contamination not defined but must be low enough
to meet allowable contamination levels at the Federal repository*- '
The above limits are based on factors that become restrictive after
the solidified HLW leaves the reprocessing plant, i.e., during transporta-
tion and at the Federal repository. The limits on linear heat release
rate and radiation, for example, can be exceeded when the canisters are
initially filled.
In general, the diameter is controlled by the heat loading of the
solidified product, the thermal conductivity, the heat removal capacity
and the maximum acceptable surface and/or centerline temperature. Opinions
vary on the limit that should be imposed on the maximum temperature.
If devitrification of the glass would lead to a more leachable product,
the temperature should be kept low enough to prevent crystallization.
On the other hand, if certain additives are used to create insoluble
crystalline phases, the above temperature limitation no longer applies.
The longer the fission products are cooled, either by storage of
spent fuel elements before processing, or by storage of liquid wastes
after reprocessing but before solidification, the larger the diameter
of the canister that can be used. A ten-year cooling of fission products
is assumed as the reference case (see Section B-3,3). Cooling for ten
years will reduce the decay heat rate by a factor of about ten relative
to the one-year value. Therefore, the discussion of heat dissipation
from waste canisters is somewhat less important than would be the case
with relatively fresh wastes. Nonetheless, heat dissipation efficiency
does allow optimization of canister volume regardless of the waste
specific heat.
-------
Four possible schemes have been proposed for increasing the allowable
heat generation in a canister. They are:
a) Use of annular Inner containers,
b) Use of cylindrical canisters with internal cooling tubes,
c) Use of cylindrical canisters with internal fins, and
d) Incorporation of metals with the waste to give improved
thermal conductivity.
these will be discussed below.
B-3.2.3.1.1 Annular Containers
Annular containers have been proposed for the British Rising Level
Glass Process (HARVEST) because a high surface area for heat removal
is needed as well as a large batch size. The design presently favored
has a 1.22 m (48 in) o.d. and a 76.2 cm (30 in) i.d. with a 2.5 cm (1 in)
wall thickness on both inside and outside shells. The canisters would
be 2.74 m (9 ft) high and filled to 1.83 - 2.13 m (6-7 ft) with glass,
and would contain approximately 1 m^ of glass. The containers are
designed for a glass with a maximum heat release of 140 W/liter and 140
kW total heat output per container.
An annular container for product storage is also proposed for the
rotary kiln-ceramic melter unit for La Hague if the heat release is
above 100 W/liter.
B-3.2.3.1.2 Containers with Internal Cooling Tubes
Internal cooling tubes have been proposed for cases in which the
canister is filled by pouring glass directly from a melter. Each con-
tainer would be fitted with seven tubes ranging in diameter from one-
twelfth to one-fourth the diameter of the container. In this way, it is
possible to increase the heat removal from a given outer container by a
factor of 10 with some sacrifice in the filled volume.
B-3.2.3.1.3 Canisters with Internal Fins
Another method of increasing the heat removal from a cylindrical
container is by fitting a number of fins inside the canister. In this
way the heat removal rate can be doubled. The latest design leaves a
13 mm (0.5 in) gap between the fin and the wall of the canister. In
this way, the wall temperature of the canister is more uniform and hot
spots are avoided where the fins would otherwise touch the wall. The
fin assembly is fabricated as a simple drop-in unit.
64
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B-3,2.3.1.4 Canisters With High-Level Waste Solids Incorporated In a
Metal Matrix -..--. . • , , , ;
The incorporation of solidified wastes in metal matrices is discussed
in Appendix B-II of this report.
B-3.2,3,2 Canister Materials
The selection of canister material is determined by tfie properties
of the contained waste, by the solidification process, and by the desired
degree of durability in the storage facility. We assume corrosion to be
much more of a threat to canister Integrity than stress caused by rock
or salt movement in the short term. If the canister does not have to
withstand a high temperature for more than a short time, as in the case
of a ceramic-melter process in which the molten material is poured into
the canister and cooling begins almost Immediately, the properties of the
glass may govern. For this situation where a relatively noncorrosive
borosilicate-glass is produced, a carbon steel canister may be adequate.
For cases where a high temperature must be sustained for a significant
time, stainless steel may be required, even with a noncorrosive glass,
because of its better structural ability to withstand the heat. This is
the situation with an in-can melting process for example. Stainless
steel may also be required for storing calcine where a relatively high-
temperature bakeout period is employed. Because phosphate glass is
corrosive to stainless steel, Inconel has been proposed for such service.
Resistance of the canister to corrosion in the geologic disposal
medium may or may not be a critical factor, depending upon the degree of
primary containment desired. To be conservative, it should be assumed
that the canister will be in a hot, moist, salt-saturated environment.
Pertinent corrosion data are presented in Appendix B-V,
Carbon steel and stainless steel have very poor corrosion resistance
in hot brine. Although Inconel has outstanding resistance to corrosion
by hot seawater at high velocity, it is quite vulnerable to pitting and
serious localized attack in quiescent seawater. '
Titanium may be the best choice for highly leachable waste forms.
Titanium and its alloys are outstanding in their resistance to seawater
under all conditions of temperature and velocity.'") Titanium corrosion
rates in hot salt water are on the order of 0.002-0.02 mm/yr (0.1-1.0 mils/
yr), the lower rate being for alloys, which corresponds to a lifetime
of 1000-10,000 years for a 25 mm (1 in) canister.
It appears that carbon steel would be the best choice for canister
material where high external corrosion rates may be tolerated. Titanium
is the best material for long-lived canisters. Stainless steel might be
the material of choice for glass forms because of the higher process
temperatures, provided a rapid corrosion rate in the geologic medium Is
acceptable.
65
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B-3.2.3,3 Post-Fill Canister Handling . • ,. ...,..•
The major post-fill operations required are seal-welding the canister
and decontamination. Other operations shown in Figure B-15 can be employed
for routine and non-routine quality control. These operations include
calorimetry, radiation profiloraetry, and nondestructive measurement of
wall thickness, and leak checking. Annealing of vitreous waste forms
may also be included as a post-fill operation.
Although it is not presently envisioned as a requirement, it may be
worthwhile to "overpack" the primary HLW canister with a secondary canister,
Possible reasons for such overpacking include:
* Simplified decontamination. The secondary canister will
not have been exposed to the contaminated environment of
the processing cell.
* Improved quality of external containment. The secondary
canister will not have been exposed to the high temperature
(500 - 1050°C) of processing and will also have a lower
differential expansion stress.
The disadvantages of overpacking include the extra costs involved
and the requirement for reducing the heat release rate in the primary
canister.
B-3.2.4 Off-Gas Treatment
Appendix B-III discusses off-gas control technology in detail. The
following information is based on that Appendix.
Gaseous effluents from nuclear fuel reprocessing plants include
chopper and dissolver off-gas and gases generated at the waste solidifi-
cation facility. The chopper and dissolver operate at about 100°C. At
this temperature, krypton, iodine, carbon dioxide, carbon monoxide,
methane, nitrogen oxides, and some hydrogen and water vapor will all
be evolved. At the higher temperatures of the solidification facility,
there may be release of ruthenium and the remaining iodine. Nitrogen
oxides are mainly a problem as interference with off-gas chemical
processes, but all of the other items carry radioisotopes of concern:
Kr-85, 1-129, C-14, H-3 (tritium) and Ru-106.
Regulations (40 CFR 190) effectively require removal of about 80%
of krypton and 99.6% of the iodine from the reprocessing plant off-gas.
Regulations for C-14 control are being considered, but the level of
removal that will be required is not yet known. No regulations
exist for controlling tritium (H-3) or ruthenium, other than the site-
boundary concentration limits of 10 CFR 20.
66
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FILLED CANISTER
FROM
SOLIDIFICATION
CELL
WELD ANNEAL
(OPTIONAL)
L
CANISTER
WELD
FINAL
DECONTAMINATION
fe £J
MAIN PROCESS ROUTE
t
I
i
i
i
i
i
I NON-ROUTINE ROUTE
CALORIMETRY f
RADIATION THICKNESS
PROFILOMETRY CHECK
TT
OVER PACK
(OPTIONAL*
ANNEAL
(OPTIONALI
TEST |
TO TRANSFER
CANAL
AND STORAGE
WELD
REPAIR
FIGURE B-15 TYPICAL POST-FILL TREATMENT OF SOLIDIFIED HLLW CANISTERS
Source! Alternatives for Managing Wastes from Reactors and
Post-Fission Operations In the LWR Fuel Cycle.
IRDA 76-43, U.S. Energy Research and Development
Administration, May 1976.
67
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Krypton and iodine removal to meet the levels specified by 40,CFR 190
appears achievable with existing technology. Removal of carbon (mainly
as C0?) and ruthenium does not appear to be difficult, but the technology
is not at the same level of development as for krypton and iodine removal.
Tritium control is very difficult because it appears as HT gas and as
HTO vapor and liquid, tending to seek isotopic equilibrium with H»0
streams.
Because removal of krypton and iodine from reprocessing plant off-
gas streams is required, storage for them will have to be provided.
The need for carbon and tritium removal and storage is still under
discussion. Ruthenium removal and temporary storage may be required to
meet site-boundary concentration limits.
For convenience of discussion, it will be assumed that the off-gas
streams from the dissolver and from the HLLW solidification are combined.
Therefore, a single removal facility (off-gas treatment plant) will be
considered.
B-3.2.4.1 Krypton-85
Appendix B—III.l discusses krypton control technology. It is clear
from that discussion that krypton removal technology is available and
will probably be incorporated into future reprocessing plants to meet
the requirements of 40 CFR 190. Because Kr-85 has a short half-life
(10.7 years), the most probable storage method is in pressurized bottles
at the reprocessing site. Steel bottles in sheltered storage are capable
of lasting at least 10 half-lives (enough time for the radioactivity to
decay by a factor of more than 1000),
B-3.2.4.2 Iodine-129
Appendix B-III.2 discusses iodine control technology and concludes
that the technology is available for removal of most of the iodine from
reprocessing plant off-gas streams. If all the iodine in the fuel is
released in the dissolver, then 99.6% removal is required to meet the
requirement of 40 CFR 190, and technology is available to achieve this.
Some iodine may, however, carry over into process streams and deposit
throughout plant processing equipment, with an undetermined effect on
the ability to meet the intent of 40 CFR 190.
The immobilization of 1-129 will be necessary because of its
extremely long half-life (1.6 x 10 years). Future storage standards
may require a solid matrix with a large loading capacity, low leachability,
nonflammability and simple handling characteristics.
68
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Investigators at Oak Ridge National Laboratory have studied the
use of cement for immobilizing iodine, and have found that 9% by weight
iodine in the form of barium iodate can be incorporated into the matrix.
This technology is designed especially to immobilize iodine collected
by the lodox process.
Use of silver and lead forms of zeolites for recovery and storage
of 1-129 has the advantage that collection and fixation of iodine occur
in the same process. Additional immobilization of the loaded zeolite
(haterials could be accomplished by the use of cements, glasses, or other
bulk solidification methods.
B-3.2.4.3 Carbon-14
The need for removal of earbon-14 from the off-gas of a reprocessing
plant (Appendix B-III.3) is still under study. Since most of the C-14
appears as CO™, however, the removal technology itself would not be
particularly difficult (C09 removal from air is a common industrial
process).
The most obvious long-term disposal option for C-14 is to convert
the wastes to calcium carbonate (CaCO_) and package it in a container
for burial in a deep geologic disposal site.
B-3.2.4.4 Tritium
Appendix B-III.4 discusses various tritium control technologies in
detail, concluding that tritium removal from reprocessing streams is
not technically achievable at present, although several methods may
prove to be practical over the long term. It is not clear that tritium
presents sufficient danger to warrant control. If, however, the tritium
should be collected, then it may either be stored in tanks as tritiated
water until sufficient decay has occurred to allow release, or it may
be placed in cement for intermediate-term solid disposal.
B-3.2.4.5 Ruthenium-106
Appendix B-III.5 briefly discusses ruthenium control, pointing out
that the release point will be at the waste solidification facility;
the release mechanisms are poorly understood. Because ruthenium
volatility is strongly dependent on temperature, it may be that ruthenium
release at the solidification facility is more a process engineering problem
(plateout and plugging) than a release problem. If it is determined that
ruthenium constitutes a release problem, several technologies are possible
for collection of the ruthenium, for example, adsorption on silica gel.
69
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With a half-life of only one year, and substantial doubt as to
whether much ruthenium-106 would be released from solidification processes,
the ruthenium control problem seems not to be serious. Additional work
needs to be done in this area, however.
B-3.2.5 Spent Fuej^AssemblyJD±sposaI
Historically, it has been assumed that spent fuel assemblies, after
some short-terra storage, would be reprocessed. The valuable products,
unburned uranium and the bred plutonium, would then be recycled into a
subsequent batch of fuel, and the waste from this reprocessing operation
would be solidified as described in previous sections of this report.
If, however, a national policy decision is made to forego indefinitely
the reprocessing of spent nuclear fuel, it will be necessary to dispose
of the spent fuel assemblies without any prior reprocessing; this is
the "throwaway fuel cycle."
To date, no criteria have been established for spent fuel disposal.
The present practice of interim storage in pools at the reactor site
or reprocessing plants will undoubtedly continue for some time. Possibly
some "stand alone" spent fuel storage pools will be built to serve for
the extended interim storage period. In this throwaway fuel scenario,
however, a permanent disposal plan for spent fuel assemblies must be
developed eventually. Although some estimates have been made of the
costs of such permanent disposal, the literature seems essentially devoid
of any account of engineering development of the required technology.
Following are some general thoughts pertaining to preliminary concepts
for disposal of spent fuel assemblies.
Three basic steps, with several sub-branches, must be considered in
a total fuel assembly disposal cycle. Briefly they are:
1. Shpr t - to In.t ermediat ^e^Tem Stor age On-S it e: High-
density rack installations in existing pools or in a
"new" independent pool to be constructed.
2. Intermediate-Term Storage Off-Site: Commitment to a
storage facility that might be constructed by a commercial
organization, by utilities (either individually or in a
consortium) or by the Government.
3. Fuel Asjsembly^^ispc-sal: Shipment to a Federal facility
for suitable preparation and geologic disposal, perhaps
after a period of near-surface (retrievable) storage.
The cost and timing of each of these steps will be determined not
only by the particular circumstances of the utility and its own plans
for storage, but also by regulations yet to be established. Technical
criteria are only now being considered for subsequent decisions that will
70
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lead to facilities to be constructed by industry1 and by the Federal
Government. Each of these steps is discussed below, together with some
of the considerations that bear on the timing and costs of the particular
step.
B-3.2.5.1 Short-Term (Local) Storage
This first step in the disposal chain, as planned by utilities,
consisted of a storage pool for spent fuel from the reactor. As ori-
ginally conceived, storage would last only about six months, at which
time fuel would be transferred from the reactor pool to a reprocessing
plant. This plan of operation (including a "one-core spare" concept)
set the pool size at about 1.3-1.7 reactor cores for a given plant.
When it became apparent that reprocessing would not be carried out
as originally planned, most utilities elected to increase local storage
capacity by substituting high-density fuel storage racks for the con-
servatively designed low-density concept. This approach represents a
relatively short-term postponement of the problem, providing relief for
up to ten years, depending upon the original pool configuration and type
of reactor.
B-3.2.5.2 Interim Storage
Interim storage represents an undefined time period, beginning when
the expanded short-term storage capability approaches its limit and
continuing to the point where fuel assembly disposal (or at least long-
term storage) under government control can be implemented. The final
disposal or long-term storage has not yet been identified by the NRC,
and until such definition the timing and actual handling procedures
cannot be established. For many utilities it is clear that some interim
storage requirement will be needed before the last disposal step. Several
plans have been suggested for these alternatives.
From a technical standpoint, interim storage may be implemented
either as an independent facility, or, if the timing of future nuclear
installations at a particular site is suitable, as an expansion of the
pool capacity of an additional unit planned or under construction. In
this way, provision can be made for longer-term storage of fuel from
the new unit, and for fuel from other units on that utility's system
as well. Technically, either solution appears to be feasible, although
an extra transportation leg is necessary with the independent facility
(unless the storage site is located at the ultimate disposal site).
The fuel storage approach is assumed to use the "conventional"
pool-type design that has been used at reactors and at reprocessing plants.
Atlantic Richfield Company has proposed a totally different approach,
in which after a storage period of up to 10 years, fuel assemblies are
canned in nitrogen- or helium-filled containers and this canned assembly
71
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is then put into individual, relatively shallow, lined holes in the
ground, with a sealed concrete cap for shielding. An array of these
caissons or wells is spread over a relatively large area on about 7.6 ra
(25 ft) centers. The advantage of such a storage plan is that It does
not require much maintenance or surveillance. It utilizes a totally
passive heat removal system, with heat flowing from the caisson through
the ground to the surface of the earth, using the atmosphere as an
ultimate heat sink. Conceptually, this kind of an approach is halfway
between interim storage and ultimate disposal; for longer-term storage
it may have substantial merit. According to Atlantic Richfield, if the
storage array is sufficiently large, the approach may be even less
costly than conventional independent pool systems for interim storage.
B-3.2.5.3 Fuel Assemblies Disposal
Extensive investigation of permanent disposal for fuel assemblies
is only now starting. A number of studies are being planned or are
beginning; criteria and specifications may be developed within the
next year or so.
The decision might be made to go to permanent deep disposal after
a period of near-surface (retrievable) burial. The sipplest of the
fuel-disposal options involves canning the fuel assembly with an inert
gas, such as nitrogen or argon. (After ten years of above-surface
cooling, it is unlikely that it would be necessary to pack"the spent
fuel in a more efficient heat-transfer medium such as metal or a salt.)
The canned assembly would then be placed in a mined-hole array like
that proposed for high-level waste disposal canisters, or in deeper
vertical holes drilled from the surface. The disposal area could be
either salt or rock.
It is clear that the technology of disposal in salt is much further
advanced than that of disposal in crystalline rock. Table B-9 summarizes
the several alternative paths or decisions that may be implemented.
The above discussion assumes that the entire fuel assembly would
be inserted in a canister, a gas added, and the canister sealed
by welding. In the case of PWR spent fuel, a single assembly would 'be
inserted in each canister. Since BW1 fuel contains less fissionable
material per assembly, two or three assemblies might be placed in a
single canister. The canister cross section might be round, rectangular,
or square. Round is simplest and would probably be least expensive from
the standpoint of the canister alone. The subject of canister material
has been explored in Section B-3.2.3.
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TABLE B-9
FUEL ASSEMBLY DISPOSAL SUMMARY
Process.. Step
Short Term
Time Frame
Up to about 10
years after dis-
charge .
Implementation Alternatives
Storage in own reactor pool
(pool capacity increased
to maximum extent).
Storage in other reactor
pool which has spare space.
Intermediate
Term Storage
About 5-20 yrs.
First independent
facilities possi-
ble about 1982
with earliest
disposal not
before 1990,
An independent storage
facility built by utility,
government or third party.
Independent dry storage
facility (ARHCO proposal
of wells in dry ground).
Fuel Assembly
Disposal
l&D 1978-1982.
Pilot operation
1982-1986.
Disposal 1990
earliest.
Process Steps Involve:
1. Shipping
2. Canister Alternatives
a. Carbon or stainless
steel
b. Titanium
c. Double canning
3. Packing Alternatives
a. Gas (N_, He, Ar)
b. Inorganic (S, NaCl)
c. Metals (Pb, Al, Sn, Zn)
4. Disposal Alternatives
a. Preliminary retriev-
able near-surface
storage
b. Deep Mine vs. Hole
c. Deep Salt vs. Rock
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B-3.3/ SELECTION OF'REFERENCE TECHNOLOGIES
It was shown in Section B-3.2 that many alternatives are being
actively pursued for the treatment of high-level radioactive wastes.
Some programs are ready for commercial application (spray, rotary-kiln,
or fluidized-bed calcination; in-ean or continuous melters). Alternative
processes or products are being developed (supercalcine, sintering, metal
matrix, glass-ceramic, coated?*pellet, thermite, ion exchange).
The evidence from a technology assessment of waste solidification
processes is that a selection could be made today from a number of pro-
cesses that, assuming competent design, construction, and operation,
would function satisfactorily for solidification of commercial reprocessing
wastes. Existing defense wastes could also be accommodated, but unless
the sodium nitrate content is removed extremely large volumes of solid
would result.
Given the wide range of possibilities, the varied state of techno-
logical development, and the differing degrees of effectiveness for the
alternatives, it is apparent that some weeding out of the less promising
alternatives is essential in order to focus the evaluation effort on the
most suitable techniques.
The approach used for this selection process has been to choose a
set of reference cases that span a reasonable range of possibilities in
terms of (a) waste form, (b) processing (solidification) approach,
(c) packaging (containment) method, and (d) isolation effectiveness. In
selecting reference cases, an attempt has been made to span a range of
possibilities wide enough to permit extension to other combinations not
specifically included in the reference cases.
In selecting reference technologies, the following criteria were
used:
1. The reference case waste characteristics should span the
reasonable range of possibilities available in the near
term (within ten years).
2. The reference case technologies should be available for
full-scale deployment for treatment of commercial reprocess-
ing wastes.
3. Reference cases are based on U.S. technology, since
there are several processes available in the United
States that will meet commercial solidification needs.
Three solid waste forms have been selected for reference cases—
calcine, glass, and spent fuel, all in suitable canisters. Disposal of
waste as liquid has not been considered as a reference case because
present regulations forbid leaving wastes as liquid over the long term.
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(This should not be interpreted as precluding the possibility of direct
disposal of liquid or slurry waste forms at some time in the future.)
B-3.3.1 Calcination
Early practical demonstration of waste solidification at the Idaho
Chemical Processing Plant (ICPP) used the fluidized-bed calciner, which
operated reasonably well in this capacity and has the most extensive
'operating history of any process in the United States. As detailed in
Section B-3.2.1, however, some difficulties have been experienced in
adapting this process to commercial reprocessing wastes. Based on an
extensive development program that has included commercial-sized faci-
lities, Battelle Northwest Laboratory (BNWL) favors the simpler spray
calciner for commercial application, an opinion that appears sound.
Therefore, the reference calcination process chosen is spray calcination.
Because the spray calciner product is a finer powder than the pro-
duct from any other processes, conservative evaluations of the leaching
behavior will result.
B-3.3.2 Classification
The reference case for glassification is taken to bfe in-can melting
rather than joule-heated melting, because of its more extensive demon-?
stration history with radioactive wastes and the simplicity of the
process operation, which does not require transfer of wastes from a
separate melter vessel.
Borosilicate glass is assumed as the reference glass because it is
universally the most accepted type.
B-3.3.3 Spent Fuel Disposal
Section B-3.2 defines a number of options for disposal of spent fuel
directly as waste. Although disposal options for spent fuel could in-
clude an intermediate processing step to reduce the spent fuel to a
modified form (e.g., glass), such options have not yet been seriously
studied, as the concept of "throwaway" fuel is very recent. The simplest
option for spent fuel disposal, and the most attractive in terms of pos-
sible intermediate retrievable storage ,is packaging in canisters with an
inert gas blanket. Although design optimization might result in use of
metal packing (for efficient heat removal) or double-canning, the simplest
case is taken as the reference—i.e., single-canning with a nitrogen
blanket.
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B-3.3,4 Canisters
Each of the above-defined reference waste treatment processes (spray
calcination, in-can melting, direct spent fuel disposal using nitrogen
blanket in a single can) requires a protective canister.
It was pointed out in Section B-3.2.3 that carbon steel is the least
expensive choice for canister material, but that it also will not last
long in a wet, warm, salt-saturated environment. Stainless steel is not
expected to be any more corrosion-resistant than carbon steel and, in
fact, might be worse. Inconel has exhibited pitting characteristics,
which put its survlvability in question. Titanium is the only practical
metal with reasonable probability of long-term (up to 1000 years) life
in wet, hot salt. It was also pointed out that the need for long canister
life is less in the case of glass than in the case of calcine.
In the reference cases, calcine product is assumed to be 'placed either
in carbon steel cans or in titanium cans. Because calcine is the most
leachable of the solidified HLW products, the carbon steel can provide a
worst-case bound on waste-disposal scenarios.
The reference canister for glass forms is taken to be stainless steel,
on the basis that in—can melting temperatures may create problems with
carbon steel cans. Use of a titanium can would be of little advantage
when used with the glass form, which has a very low leach rate.
Two reference cases are assumed for spent fuel canisters—again,
carbon steel and titanium. Because of the possibility that spent fuel
might be highly leachable, this pair of cases provides a test of the
sensitivity of high-leachability forms to canister corrosion rates.
B-3.3.5 Off-Gas Residue Jisposal '
Volatile wastes released to the reprocessing plant off-gas system
(either from the chopper/dissolver or from the solidification step),
may need to be collected and stored prior to disposal. Section B-3.2,4
and Appendix B-III show that the volatiles of concern are krypton-85,
iodine-129, carbon-14, tritium, and perhaps ruthenium-106.
Krypton removal by either cryogenic distillation or fluorocarbon
absorption is feasible, and there are no significant barriers to appli-
cation at commercial reprocessing plants. In either case, the resultant
product will be krypton gas (mixed with quantities of other gases,
depending on the technique selected). Although several "exotic" krypton
stabilization techniques are being investigated (zeolite or clathrate
entrapment), it is expected that the storage method finally selected
will be containment in carbon steel cylinders, which are expected to
have a lifetime far in excess of that needed to ensure decay of the
krypton to very low levels.
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Iodine removal is feasible and available for commercial application;
the Mercurex process followed by silver zeolite polishers was selected
for application at Allied General Nuclear Services (AGNS). The silver
zeolite polisher concept is conceptually simple and may achieve almost
any degree of decontamination desired,by addition of more polishers.
The Mercurex process competes with the lodox process for preliminary
Iodine removal (see Appendix B-III.2 for details); since the Mercurex
process generates the larger volume of waste, it will be conservatively
Assumed that the Mercurex process is used for the treatment ahead of
the silver zeolite beds. The waste form assumed for the Mercurex process
is mercuric iodate. The simplest long-term disposal procedure would then
be simply packing the mercuric iodate and spent silver zeolite in steel
canisters.
Carbon-14 will mostly be in the form of carbon dioxide, which may be
removed by a number of processes that yield calcium carbonate as their
common product. This could then be placed in steel canisters for ultimate
disposal.
Tritium control does not appear to be feasible or necessary at
present, and the reference case is, therefore, taken to be release.
The half-life of ruthenium-106 is so short that no reference case Is
presented.
B-3.3.6 Cladding Hulls and Fuel Bundle^ Residues
The reference case for disposal of fuel cladding hulls and fuel
bundle residues is based on proposed regulatory framework implicitly
requiring geologic disposal of wastes containing greater than 10 nCi/g
TRU contamination. Such a regulation provides strong economic Incentive
(discussed in Section B-3.4) for volume reduction to reduce geologic
disposal costs. The proposed 10 nCi/g TRU rule, therefore, encourages
sorting of fuel bundle residues in order to reduce the volume of conta-
minated wastes that must be dealt with. This process step is particularly
attractive if, as appears to be the case, the contaminated and non-
contaminated wastes are relatively simple to separate—in this case, by
mechanical means.
On the other hand, the 10 nCi/g rule would probably discourage the
addition of a decontamination step to the cladding treatment, since the
reprocessor will achieve no cost savings for reducing the amount of TRU
contamination present in the cladding, as long as it is contaminated to
higher than 10 nCi/g. (This may be something of an oversimplification,
since decontamination may allow reductions in handling costs at subsequent
processing steps.) Although decontamination appears logical (since
it may concentrate most of the contamination for greater ease of handl-
ing and immobilization), the reference case assumes no decontamination,
reflecting what appears to be the most likely actual process.
77
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Because volume reduction appears to carry with it economic rewards
through reduced costs for geologic disposal, a simple mechanical volume
reduction step is assumed (compaction). Compaction is estimated to
achieve three-fold volume reductions. Because no practical melting
technology is at hand, only mechanical compaction is assumed.
In Section B-3.4 it will be pointed out that there may be a need
for further definition of TRU wastes into low- and high-level categories
in order to delineate more precisely the actual level of risk, instead of
treating identically all wastes contaminated to greater than 10 nCi/g.
What may be called for is a risk assessment that considers lower con-
tamination levels and higher stability of form. Were such a program
implemented, then further treatments of cladding waste (matrix formation
in glass, sand or concrete, or conversion to non-inetals) could be con-
sidered as part of this optimization process. As an example of this
type of trade-off, it is not clear whether, under a sliding-scale TRU
regulation, decontamination of the hulls followed by solidification of
the decontamination residue and packaging of the hulls for low-level
storage would be better than solidification and disposal of the
undecontaminated hulls directly.
The reference case used is disposal of chopped hulls, sorted to re-
move end fittings, grids, and guide tubes, compacted to about one-third
of the original volume, and packed into carbon steel containers for ul-
timate disposal. This reference case, based upon proven technology, is
conservative insofar as it produces a waste form that could be improved
upon by more advanced treatments.
B-3.3.7 Summary of Reference Cases
Following are the assumed reference cases for HLW disposal:
• Calcine: spray calcination, calcine contained in carbon steel
or titanium cans for deep disposal.
* Glass: in-can melt, glass contained In stainless steel cans
for deep disposal.
* Spent fuel: nitrogen-blanketed in carbon steel or
titanium cans for deep disposal.
• Cladding residue: mechanically sorted from end pieces,
compacted to one-third the original volume, placed in
carbon steel cans for deep disposal.
In addition, for associated wastes, reference cases are:
* Krypton gas storage at the reprocessing plant in carbon
steel gas cylinders.
78
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* Iodine recovery by Mercurex process followed by silver
zeolite beds, the mercuric iodate and spent zeolite
packed into carbon steel canisters for deep disposal.
* Carbon-14 recovery as calcium carbonate contained in
carbon steel canisters for disposal,
• Tritium release during reprocessing, or from spent
fuel as the fuel is leached.
B-3.4 DISPOSAL OF OTHER TRU-CONTAMINATED WASTE
Task A described the nature of low- to intermediate-level transuranic-
contaminated (TRU) wastes that originate in the nuclear fuel cycle and
estimated the possible quantities of TRU waste arising from various fuel
cycle scenarios. The subject of TRU waste treatment covers a broad
range because of the wide variety of waste characteristics. Wastes may
include general process trash (rags, plastic, paper, tools, glass and
metal), failed or expended equipment (machines, valves and glove boxes),
and liquids (slurries, sludge, resins, oil, and grease). In this section,
the technologies for treatment of these wastes will be discussed.
As was pointed out in Task A, considerable uncertainty exists con-
cerning the physical characteristics of TRU waste arising from various
stages of the fuel cycle. A degree of uncertainty still exists about the
ultimate disposition of TRU waste, although the regulatory thrust seems
to be in the direction of requiring geologic disposal of waste contaminated
to levels above 10~^Ci/g. If this proposed "10 nanocurie rule" is adopted,
there will be considerable economic incentive to reduce TRU waste volume.
The fact that ERDA's New Mexico salt project is primarily designed to
accommodate low-level TRU waste, albeit retrievable, '•••") Is Indicative of
the trend toward geologic disposal of TRU wastes. The current Office of
Waste Isolation (OWI) commercial radioactive waste disposal program also
includes provisions for TRU waste disposal,' '
As an illustration of the economic incentive for TRU waste volume
reduction, cost figures to be given in Section B-4 for a large low-heat-
generation geologic waste disposal facility will show that for every cubic
meter of TRU waste volume reduction, there is a potential cost saving
of at least $860. It is thus not surprising that a comprehensive program
is under way to develop volume reduction techniques for handling TRU
wasted18'
TRU waste, although it takes many forms, may be classified into
three categories for purposes of discussing treatment methods: combus-
tible, non-combustible, and liquids/sludges, which would be evaporated to
a sludge or cake and fixed in a solid such as concrete. Figure B-16 shows
an overview of possible TRU waste material flows through the various waste
treatment stages.^ '
79
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Gaseous
Metal
(Equip., Debris,
Cladding Hulls)
Combustible
(Paper, Resins,
etc.)
Liquid
(Solvents, Acids
Filter
• Smelt
• Compact
• Pkgw/o
Treatment
Incinerate
• Evaporation
• Incinerate
• Clariflocculation
• Volume Reduot
• Pkgw/o
Treatment
• Metal Recovery
• Fix
• Encapsulate
Fix Ash
Fix Residues
CO
o
Metal
(Equip., Facilities,
Debris)
Combustible
etc.)
Liquid
(Solvents Acids.
Effluents)
• Smelt
* Compact
Treatment
• Evaporation
• Clariflocculation
• Volume Reduce
• Pkgw/o
Treatment
• Metal Recovery
• Fix
1 nterin
i Storage
ERDA and Commercial
Inttrim Stored Wactt
Metal
Debris
Solidified
Combustible
• Smelt
• Compact
• Pkg w/o
Treatment
Pkg w/o Treatment
Incinerate
• Metal Recovery
• Fix
• Encapsulate
• Fix Ash
• Encapsulate
Drums
— Packaqe _ Interim Storage
1
FIGURE B-16 MANAGEMENT METHODS FOR TRANSURANIC-CONT AMI MATED WASTES
Source: Wolfe, R.A. The Research and Development Program for Transuranic-Contaminated
Waste Within the U.S. ERDA. In: Radioactive Wastes from the Nuclear Fuel'Cycle
AIChE Symposium Series 154:72, 1976.
-------
TRU waste is mostly "low-level TRU waste", which means that the gamma
radiation is sufficiently low to allow handling of packaged wastes without
shielding (less than 10 mrem/hr surface radiation). These wastes are
characterized'!'/ as having an average plutonium content before compaction
6f 9 g/m-% corresponding to a plutonium processing loss of about 1%.
Also generated will be "intermediate-level TRU waste", which is character-
ized as having sufficient external radiation (10-1000 mrem/hr) to require
some shielding and special handling, but a lower average plutonium content
(1 g/m^, corresponding to a 0.025% plutonium loss). Because this inter-
mediate-level TRU waste appears in lower quantities than low-level TRU
waste (see Task A), it may be possible to eliminate shielding requirements
for intermediate-level TRU waste by combining the wastes, and thereby
achieve self-shielding sufficient to reduce surface radiation levels.
A 'first step toward reduction of TRU waste volume is minimizing the
waste generation from fuel cycle processing steps. Waste minimization may
be accomplished both by the design of a facility and by administrative
controls, such as procedures that minimize the production of analytical
or clean-up waste. Facility design can contribute to reduction of waste
quantities by increasing chemical concentrations in process streams,
recycling waste streams, and optimizing ventilation flow rate controls to
increase filter life in dirty conditions. Administrative controls include
such techniques as recovering spills with solution chemistry rather than
rags and towels, using corrosion-resistant tools, and decontaminating the
waste itself (e.g., tool washdown and re—use). Assay and sorting may
achieve further volume reduction by releasing wastes with less than 10
nCi/g. At upwards of $860/m~ disposal costs, a sorting/decontamination
operation could perhaps be economically justified.
After steps have been taken to 1) reduce primary waste quantities;
2) decontaminate waste to return some of it to unrestricted use; and 3)
sort the contaminated waste into combustibles, non-combustibles, and
liquids, steps may be taken to reduce waste volume, and to stabilize and
package the waste.
B-3.4.1 Combustibles
The simplest treatment for combustibles is compaction, which is also
inexpensive. In Task A it was shown that a four- to five-fold volume
reduction may be achieved. Compaction of combustibles is probably not the
best solution, however, because it leaves two potential problems with the
waste: fire hazard, which is a remote possibility, especially if the
compacted wastes are packaged in metal drums or other fire-resistant
packages; and radiolysis of waste hydrocarbons, releasing hydrogen.
Incineration appears to be the most promising method of treating TRU
wastes. Six potentially attractive incineration methods—controiled-air,
pyrolysis, molten salt, fluidized-bed, cyclone, and acid digestion—have
baen identified and discussed in the literature.^18,20,21,22) xhe state
of development of these six techniques is shown in Figure B-1?A8) A
brief description of each method follows.
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Incineration System
Fluidized Bed (Rocky Flats)
Cyclone (Mound)
Acid Digestion (HEDL)
Molten Salt (Ai)
Pyrolysis (PNL)
Controlled Air (LASL)
Target Dates
FY 75 | FY 70 | FY 77
Capacity
~ 20 Ib/hr
~ 180 Ib/hr
< 50 Ib/hr
< 10 !b/hr
< 100 ib/hr
< 50 ib/hr
~ 400 ib/hr
V Testing with Non-Contaminated Waste
IT Testing with Plutonium-Contaminatid Waste
FIGURE B-17 DEVELOPMENT OF INCINERATION METHODS FOR
TRANSURANIC-CONTAMINATED WASTES
Source: Wolfe, R.A. The Research and Development Program for
Transuranic-Contaminated Waste within the U.S. ERDA.
AIChE Symposium Series 154:^2, 1976.
82
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B-3.4.1.1 Controlled-Air Incineration^18'22^
Los Alamos Scientific Laboratory (LASL) is building a,45 kg/hr
(100 Ib/hr) developmental incinerator using an adaptation of standard
municipal incinerators, with a combustion and off-gas burning chamber to
achieve complete combustion. Off-gases will be scrubbed and liquid waste
will be sent to a disposal facility. -Although a volume reduction factor
of 20 is conservatively predicted,^3) LASL has informally estimated a
net volume reduction factor of 30 from primary waste after inclusion of
secondary wastes arising from the incineration system.
B-3.4.1.2 Pyrolysis(18>22)
The pyrolysis-burning concept (beating in an oxygen-deficient atmos-
phere, to gasify part of the waste material, which is then combined with
air and burned in a secondary combustion chamber) is being developed at
BNWL at Mchland, Washington, and is an adaptation of systems used to
treat industrial waste. The process achieves a significant volume
reduction by incineration. The process has been demonstrated with non-
radioactive simulated waste at 15 kg/hr (33 Ib/hr).
f1<3 on 221
B-3.4.1.3 Molten Salt Combustion^ '''
The molten salt combustion process is being developed by Atomics
International Division of North American Rockwell Corporation at Canoga
Park, California. Waste volume is reduced by utilizing molten sodium
carbonate or lithium carbonate as a medium to burn organic materials, trap
particulates, and react chemically with any acidic gases produced during
combustion. The radioactive materials are contained in the molten salt.
The melt-ash mixture is processed to separate ash for disposal, to recover
the salt for recycle, and potentially to recover plutoniutn. Combustion
has been demonstrated at 0.5 kg/hr (1 Ib/hr) on a bench scale with wastes
contaminated with plutonium and beta-gamma activity, and on a pilot scale
with uncontaminated waste.
B-3.4.1.4 Fluidized-Bed Combustion '
A fluidized-bed combustion system is currently being installed in a
plutonium recovery facility at the Rocky Flats nuclear defense plant,
operated for ERDA by the Rockwell International Company. A prototype unit
was tested with noncontaminated waste for about a year and became opera-
tional about May 1975 for the recovery of plutonium-contaminated scrap
materials. This prototype system utilizes a fluidized combustion bed,
a fluidized catalytic afterburner, and a porous metal filter. The
system has a continuous capacity of about 10 kg/hr (22 Ib/hr). A
demonstration plant rated at 82 kg/hr (180 Ib/hr) is being built.
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B-3.4,1.5 Cyclone
The cyclone incinerator concept has been tested at the ERBA Mound
Laboratory operated by the Monsanto Research Corporation in Miamisburg,
Ohio. This system utilizes the concept of igniting waste within a metal
drum container while injecting air in a vortex pattern to effect complete
combustion. The gaseous effluents are passed, through a self-cleaning
fiber-bed filter system. The water used for scrubbing the filter will be
neutralized and recycled within a closed-loop system. The demonstration
unit being tested has a capacity to incinerate up to 23 kg/hr (50 Ib/hr)
of plutonium-contaminated combustible waste.
B-3.4.1.6 Acid Digestion(18'21'22j
This process is being developed at Hanford Engineering Development
Laboratory (HEDL) to reduce the volume of combustible waste by digesting
it in hot (230-270°C) concentrated sulfuric acid containing nitric acid
oxidant to form noncombustible residues having less than 40% of their
original volume. The residue is also in a form that could provide for
recovery of the plutonium. HEDL has demonstrated this process at 0.5 kg/hr
(1.1 lb/hr)with nonradioactive waste. A radioactive waste demonstration
unit of the same size is being constructed.
B-3.4.1.7 Summaries of Incineration Development
Cooley and Clark catalogue incineration processes, including vortex,
agitated-hearth, and moving-belt approaches.(22) Table B-10 lists
incinerators that have been or are being used for combustion of radioactive
solid wastes; Table B-ll lists processes under development. Perkins has
provided a further review.'") -j^g most promising techniques appear to
be the LASL controlled-air and the Rocky Flats fluidized-bed processes.
B-3,4.2 Non-Combustibles(18>22^
After they are decontaminated and sorted to reduce waste quantity to
a minimum, non-combustibles may be compacted or smelted to reduce volume.
Compaction may consist of mechanical disassembly and/or compression with
a heavy compactor. Depending on radiation levels, this may be accomplished
by contact maintenance (Rocky Flats uses a special room with maintenance
personnel in pressurized suits) or by remote maintenance for higher-gamma
radiation.
Smelting is under development and has the advantages of immobilization
of waste (decreasing the surface-to-volume ratio) and volume reduction.
84
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TABLE B-10
oo
Ln
INCINERATORS USED IN THE UNITED
STATES FOR THE
SOLID WASTES.
Incinerator installation
Knolls Atomic Power Laboratory
Argonne National Laboratory
Bcttis Atomic Power Laboratory
Shippingport Atomic Power Station
United Nuclear Corporation
Metals and Controls
Pratt and Whitney Aircraft Company
Douglas United Nuclear
Yankee Rowe Atomic Electric Company
National Lead Company of Ohio
Union Carbide K-25 Plant
Union Carbide -Paducah
General Electric Company,
Atomic Power Equipment Department
Goodyear Atomic Corporation
Gulf General Atomics
Nuclear Fuel Services, Inc.
Union Carbide Y-12 Plant
Babcbck & Wilcox
Los Alamos Scientific Laboratory,
plutonium-recovery incinerator
Mound Laboratory,
plutonium-recovery incinerator
U.S. Army Nuclear Defense Laboratory
Location
New York
Illinois
Pennsylvania
Pennsylvania
Connecticut
Maine
Connecticut
Washington
Maine
Ohio
Tennessee
Kentucky
California
Ohio
California
Tennessee
Tennessee
Virginia
New Mexico
Ohio
Maryland
Year
built
1949
1951
1953
-
-I960
-I960
-I960
1967
-1968
1954
1972
-
I960
1971
1963
-
1955
1972
1952
-1972
1963
Operational
status
Not in operation
Not in operation
Not in operation
Not in operation
Not in operation
Not in operation
Not in operation
Not in operation
In operation
In operation
In operation.
Not in operation
Not in operation
In operation
In operation
In operation
In operation
'In operation
In operation
In operation
Not in operation
COMBUSTION OF RADIOACTIVE
Operating features
Single chamber; pure OT;
550°C
Single chamber; 450 to
600 C;vibrating grate;
gas fired
Single chamber; cyclone
air feed
Cyclone air feed
-
-
-
Natural draft incinerator
for open-pit burning
Cyclone air feed
Dual chamber; 980°C;
gas fired
Dual chamber; 930 to
1 100°C; gas fired
-
Triple chamber; 760 to
820°C;gasfired
Dual chamber; 815 to
1000°C; gas fired
Dual chamber; 900 to
1 200°C; gas fired
Dual chamber; gas fired
Single chamber; 870°C;
gas fired
Single chamber; 1090°C
Single basket; 800°C;
electrically heated
Single basket; 800°C;
electrically heated
Dual chamber; cyclone
air feed
Capacity
(kg/hr)
9 to 18
68
-
18
95
90
270
-
18
1000
_
-
45
68
20
270
\20
80
<1
23
23
-------
TABLE B-10
(continued)
oo
Incinerator installation
Kerr-McGee Nuclear Corporation
Los Alamos Scientific Laboratory
(development)
Mound Laboratory (development)
Rocky Flats Plant
(development-agitated hearth)
General Electric Nuclear Fuel Plant
Westinghouse Nuclear Fuel Division
United Nuclear Corporation
Rocky Flats Plant
(development-rotary kiln)
Rocky Flats Plant
Babcock & Wilcox
Nuclear Materials Division
Atlantic Richfield Hanford Company
plutonium-recovery incinerator
Location
Ohio
New Mexico
Ohio
Colorado
North Carolina
South Carolina
Rhode Island
Colorado
Colorado
Pennsylvania
Washington
Year
built
1972
(1976)
1975
(1980)
1972
1974
1967
(-1980)
1959
1969
1961
Operational
status
In operation
Under construction
In operation
Under construction
In operation
In operation
Not in operation
Under construction
In operation
Not in operation
In operation
Operating features
Dual chamber; gas fired
Dual chamber; 500 to
800°C; controlled air
Single chamber; cyclone
air feed; 1100°C
Agitated hearth; 800 to
1000°C; fired with
dicsel fuel
Vortex burner; 815 to
980°C; gas fired
Dual Chamber; 650 to
1 200°C; excess air
burner
Single chamber; gas fired
Rotary kiln; 600 to
800 C; fired with
diesel fuel
Dual chamber; 1200 to
1400°C
Batch burner; dual
chamber; gas fired
Dual chamber; 700 to
800 C; moving grate
Capacity
(kg/hr)
70
45
35
70
450
_
45
40
16
9
2
Source: Cooley, C.R. and D.E. Clark. Treatment Technologies for Non-HIgh-Level Wastes
(USA). In: Proceedings of the International Symposium on the Management of
Wastes from the LWR Fuel Cycle. CONF 76-0701, U.S. Energy Research and
Development Administration, 1976
-------
oo
vj
TABLE B-ll
PROCESSES UNDER DEVELOPMENT IN THE UNITED STATES FOR THE COMBUSTION OF
RADIOACTIVE SOLID WASTES
"
Development status
Radioactive unit
Process
Controlled-air incineration
Cyclone incineration
(excess air)
Fluidized-bed incineration
Rotary-kiln incineration
Agitated-hearth incineration
Controlled pyrolysis
Molten-salt combustion
Acid digestion
Noncombustible
product
Highly refractory oxide
Highly refractory oxide
Refractory oxide
Refractory oxide
Refractory oxide '
Refractory oxide
Salt-ash or an oxide if
salt is leached
Sulfates and oxides
Site of
development
Los Alamos
Scientific Laboratory
Mound .Laboratory
Rocky Flats
Rocky Flats
Rocky Flats
Pacific Northwest
Laboratories
Atomics International
Hanlord Engineering
Development Laboratory
Tested
capacity
(kg/hr)
45
35
9
2
4
15
50
5
Design
capacity
(kg/hr)
45
35
82
40
70
None
50
5
Scheduled
startup
1977
1978
1977
1980
1980
b
1978
1977
Commercial
availability
1978
1980
1980
1982
1982
ft
1978
1980
"All processes lead to volume-reduction factors in the range of 20 to 40, assuging the same mixed waste for each process and initial processing only (Le., no
final conversion of residue nor secondary waste generated by the process itself).
*Not currently funded;estimated 3 years needed to develop process for commercial application.
Source; Cooley, C.R. and D.E. Clark. Treatment Technologies for Non-High-Level Wastes
(USA). In: Proceedings of the International Symposium on the Management of
Wastes from the LWR Fuel Cycle. CONF 76-0701, U.S. Energy Research and
Development Administration, 1976
-------
B-3,4.3 Wet Wastes(Liquid and Sludges)(18>22)
There are a number of options for treatment of wet wastes: evapora-
tion, drying or calcination for high-solids content; filtration or ion
exchange for low-solids content; combustion for combustible liquids.
These techniques are widely used for both radioactive and non-
radioactive industrial processes, and the choice of technique will depend
on design compatibility and economics. The resultant residue will be a
solid, either salt cake, resin or calcine. Reverse osmosis or flocculation
may be used as a concentrating mechanism for low-solid liquids, but a
second drying or filtering step must be used to solidify the resultant
concentrates.
B-3.4.4 Final Packaging
TRU waste is now placed .in crates or drums. The crates are fiber-
glass-coated plywood with polyvinylchloride (PVC) and fiberboard liners
built to meet Department of Transportation Specification DOT-7a. The
plastic-lined, 208-liter (55-gal) drums are built to meet Department of
Transportation Specification DOT-17c.(25) These packages are suitable for
retrievable storage. For ultimate TRU waste storage, these retrievable-
storage drums or crates may be used, or the wastes may be transferred to
a container that is easier to handle in a geologic disposal facility.
Although TRU waste is presently defined simply as waste containing
greater than 10 nCi/g transuranic contamination, in fact there will be a
wide range of contamination levels, up to perhaps several million nCi/g
on some ventilation filters. The options ultimately selected for fixation
of these TRU wastes should reflect this range of contamination, with higher
contamination levels subjected to high levels of fixation. It may, for
example, be desirable to fix higher-level wastes in concrete, whereas
drums may be satisfactory for lower levels. In general, the work necessary
to define these options has not yet been performed. There is considerable
experience at ORNL, however, in using concrete and grouts to fix inter-
mediate-level wastes.'26; jn addition, Atlantic Richfield Hanford
Company (ARHCO) and BNWL have investigated placing alpha wastes in glass
on a laboratory scale, successfully converting certain types of waste
to immobile silicate glasses.' ''
B-3.5 ALTERNATIVE GEOLOGIC DISPOSAL TECHNIQUES
Disposal of radioactive wastes in deep, stable geologic formations
has long been considered the preferred method for isolation of these
wastes from contact with man's environment. A number of possible
geologic media have been considered for such disposal; these include
salt beds, salt domes, crystalline rock forms such as granite or basalt,
shales, limestones, certain types of clay beds, and others. To
date salt deposits have received the most attention, especially in the
88
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United States, because of their demonstrated stability over very long
time periods, their homogeneity, and their capability of plastic flow
(self-healing) in the "presence of stress. The self-healing characteris-
tics of salt effectively eliminate the possibility of extensive cracking,
thereby preventing the opening of pathways to the environment.
An alternative to salt is a stable crystalline rock, such as basalt
or granite. Again, there are abundant examples of suitable depth and
age with demonstrated seismic stability. Crystalline rock does not
have the self-healing characteristics of salt, but possesses other
advantages, e.g., resistance to water intrusion, that make it a desirable
medium for geologic disposal of radioactive waste.
In addition to salt and crystalline rock, other geologic media
have been considered, e.g., shales and clay deposits. In general,
these have both desirable and undesirable characteristics. For
example, the laminar structure of horizontally-bedded shale reduces
water permeability in the normal (vertical) direction, but the presence
of water could lead to high stresses aud possible disintegration of
the shale at high temperatures. Certain types of clay-till deposits
also have the advantage of low water permeability, but the disadvantage
of indeterminate long-term stability characteristics.
In light of the foregoing, the following have been selected
as reference cases for the purposes of this study:
(a) Disposal in bedded salt
(b) Disposal in granite
(c) Disposal in salt domes
A uniform depth of 460 m (1500 ft) has been assumed for all
three cases.
The design of disposal facilities for radioactive wastes generally
follows conventional practices with two important exceptions: (1) special
provisions are required for safe transport and cooling of radioactive
material; and (2) waste canister spacing and arrangement must be in
accordance with thermal design criteria for the specific medium. These
criteria generally take the form of bulk temperature limitations and/or
temperature gradient limitations to ensure reasonable integrity of the
repository. Because of their Importance, these thermal criteria will
be discussed in considerably more detail in Section 8-3.5.3
Until recently, salt was the leading contender for geologic disposal,
but more attention is now being focused on other media, such as crystalline
rock. Despite growing Interest in these other media, nearly all of the
work performed to date on engineering design of disposal facilities has
been for salt deposits; the information available for media other than
salt is extremely limited. Therefore, the major focus of this section
of the report will be on salt deposits, principally bedded salt.
-------
B-3.5.1 Design Characteristics for Disposal in Bedded Salt
The reference design described below assumes non-retrievable
underground mine storage in a bedded salt deposit. The state of the
art for the technology required to develop such a facility is assessed
and areas where further study is required in order to optimize waste
handling and facility design are noted. Fully retrievable storage and
storage in other media, such as granite or shale, are also discussed
briefly.
Since it is not clear at present whether reprocessing of spent
fuel will be resumed, two reference waste burial facility designs will
be considered in this section. The first design is based on storage
of the wastes from the reprocessing of plutoniutn-recycle fuel. The
second design is based on storage of spent fuel and other wastes from
a throwaway fuel cycle.
The two facility design concepts are similar. Each will have a
surface facility; a series of shafts to transport personnel, equipment,
and wastes to the mine level; and segregated storage locations in the
mine for storage of HLW or spent fuel, cladding wastes, intermediate-
level TRU wastes, and low-level TRU wastes. Major differences are in
the equipment dimensions, mine corridor dimensions, waste canister spacing,
and the overall number of waste units to be buried. Because of the basic
similarity for the two reference facilities, the following discussion
will concentrate on the design of a facility to bury wastes from a
reprocessing fuel cycle. The major design differences for the throwaway
fuel cycle burial facility will be noted at the end of this discussion.
The reference design for this study assumes burial in a mine located
460 m (1500 ft) below the surface. The facility's general arrange-
ment is similar to the waste isolation facility developed under contract
to the Office of Waste Isolation (OWI).'!'' Various components of the
facility will be discussed in more detail below.
The reference facility is sized to accept ten-year-old wastes from
reprocessed equilibrium mixed-oxide fuel or ten-year-old spent fuel from
a throwaway fuel cycle. The ten-year delay between fuel discharge and
waste burial is arbitrary. NRC regulations require that all HLW from
a reprocessing stream be solidified within five years and sent to a
Federal repository within ten years. With older fuel, there are handling
advantages, such as reduced shielding and heat removal requirements, and
mine loading can be increased somewhat. Further study is needed to
establish an optimum age for waste burial, but this age is likely
to be at least ten years.
Limiting planar heat input criteria for HLW have been established.
A limit of about 150 kW/acre has generally been assumed, but the design
presented herein assumes a somewhat lower value (126 kW/acre). The
facility is sized to receive all the HLW, the cladding wastes, and
90
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the other transuranic wastes expected to be generated by the LWR fuel
cycle through the end of the century.^ * ^ This corresponds to all
the wastes generated by 140 GW'of installed nuclear capacity over a
30-year lifetime or the equivalent of approximately 107,000 MT of heavy
metal originally charged to the reactors. Thus, for the 700 GW
additions to total LWR capacity assumed in this study (see Task A report,
Section 5.6), five such reference facilities must eventually be built.
It should be noted that the 140 GW facility size is purely arbitrary.
Should the number of suitable sites available for waste burial be limited,
a greater amount of waste can be placed at an Individual site.
B-3.5.1.1 Waste Packaging
For purposes of this study, it has been assumed that HLW and
cladding/intermediate TRU are packaged in canisters with the same dimensions
in order to permit standardization of shipping casks, waste receiving and
handling equipment, and in-mine waste handling and placement equipment.'-^ '
Canister size for a given facility will depend on whether HLW or spent
fuel is being buried. For HLW, a cylindrical canister with a 30.5-ctn
(12-in) i.d. and a length of 3.05 m (10 ft) has been suggested. Canister
designs have not been developed to date for a throwaway fuel cycle, but
would probably be approximately the same diameter and about 4,57 m (15 ft)
in length, i.e., long enough to contain a fuel assembly packaged intact.
It should be noted that cladding and, in general, the TRU wastes
produce only small quantities of heat.(l?»19) Furthermore, TRU will
require less shielding than HLW. It is, therefore, believed that
the use of shielded bulk shipping containers will be considered for TRU.
Such containers could consist of stainless steel—lined, pre-fabricated
concrete of sufficient thickness to provide shielding that will permit
direct personnel contact and handling with conventional equipment, such
as fork lifts.
Packaging of low-level TRU wastes in standard 208-liter (55-gallon)
drums with a unit package consisting of 12 such drums has been suggested.
Bulk packaging in 1,4-m3 (50-ft3), 2.8-m3 (100-ft3), or 5.6-m3 (200-ft3)
containers should also be considered to optimize handling and costs.
B-3.5.1.2 Waste Receipt at Facility
The waste burial facility must be capable of receiving shipments
of all forms of waste by either truck or rail transport. Shipments of
low-level TRU will probably be made by truck, with shipment of up to 8
units of 12 standard 208-liter (55~gallon) drums each, or an equivalent
freight volume of bulk shipping containers. Canistered waste shipments
of HLW, spent fuel, cladding hulls, and intermediate-level TRU will
probably be made by rail in large shielded casks capable of transporting
approximately 10 to 12 canisters each. Receipts will be assumed to
occur over a ten-year period.
91
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A cask is a massive shielded, sealed container, 'with heat removal
capability, designed to transport radioactive material safely through
non-restricted areas. The cask must be designed to have a very low
probability of radioactivity release, even in the event of a serious
shipping accident. Such rail casks (for HLW canisters) have not yet
been developed, although the technology is readily available from
experience with spent fuel casks. Limiting cask design parameters are
expected to be in the areas of shielding, heat removal systems, weight,
and size. Cask designs must be reviewed and licensed by the U.S. Depart-
ment of Transportation prior to use. Because of weight limitations on
highway transport, smaller casks designed for truck shipment will be
able t-o carry only one or two canisters per shipment.
The amounts of the various materials to be received at the reference
facility are shown in Table B-12. Since a total of approximately
350,000 canisters must be received at the HLW facility, and as many as
450,000 canisters at a throwaway cycle facility, the larger volume-per-
shipment rail shipping mode is indicated. If bulk shipment of intermediate-
level TRU and/or cladding hulls is shown to be feasible, the shipments
may be made by truck rather than rail because of the greater flexibility
in handling and routing.
B-3.5.1.3 Surface Facility
The surface facility must include all the buildings and services
required to support the receipt, handling, and in-mine placement of
the various types of waste. Essential components of the surface facility
are discussed briefly in Appendix B-IV. The equipment, safety systems,
and building structures must be designed to appropriate codes and
specifications in order to ensure the health and safety of the operators
and the general public during both normal operations and abnormal
occurrences. NRC has yet to publish specific design criteria for waste
burial facility construction and operation.
B-3.5.1.4 Site Access
A rail spur and access highway must be extended to the repository
site. Adequate switching facilities and rail siding must be provided
on-site in order to receive and handle efficiently as many as 10 to
12 rail casks per day for the receipt of HLW spent fuel, cladding hulls,
and intermediate-level TRU canisters.
Electrical, sewer, and water services must also be extended to the
site.
92
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TABLE B-12
v.O
Type of Waste
HLW
Cladding hulls
Intermediate-le
Low-level TRU
VOLUMES OF WASTE TO
Unit (A)
Packaging
Description
canister
canister
;1 TRU canister
55-gal. drum
BE STORED IN REFERENCE
(*)
HLW Facility v '
No. of Waste
Units
35,500
58,500
255,000
655,000
FACILITY
Volume
(m3)
10,750
9,240
46,800
140,000
Throwaway
Facility
No. of
Waste Units
319,500
0
130,000
330,000
*Based on information contained in ERDA~76^43, Appendix C.
HLW canister is 30.5 cm diameter x 3.05 m long, total weight is about 900 kg. Spent-
fuel canister would be about the same diameter and 4.6 m long; total weight would be
about 1200 kg.
-------
B-3.5.1.5 Below-Ground Layout
B-3.S.1.5.1 Thermal Considerations
In order to minimize area commitments and waste disposal costs,
it is desirable that a maximum amount of HLW be placed in a given
mined area, subject only to as yet imperfectly-understood constaints
on temperature distributions in the repository. Certain trade-offs
must be evaluated, using good engineering judgment. The major areas
to be considered in setting limits on repository temperature distri-
butions are as follows:(2,28) These should be viewed only as possible
considerations relating to repository thermal loading.
* Maximum allowable temperature at the centerline of the waste
container
For glass, the centerline temperature limit is considered to
be on the order of 600 C, somewhat below the devitrification
level for glasses and substantially below processing temperatures
in order to preclude additional off-gassing.
* Temperature limitationsimmediately adjacent to the storage
canister
/o OQN «
It has been suggested . ' that temperatures of 250 C should
not be exceeded in more than about 1% of, the salt within Jne
boundaries of the entire canister array. This temperature
limit is set by concern over too high a surface temperature near
the canisters. The corrosion problems associated with it would
also minimize the amount of salt in a highly plastic state in
the immediate canister region, i.e., extending about 20 cm
into the salt from the canister surface.
* Temperature limitations throughout the mined area
Because of the salt plasticity, temperatures must be limited
to values that will not cause stability problems during
mining operations. Somewhat higher bulk temperatures are
desirable after mine closure, however, in order to insure
fairly rapid filling of voids. Taking account of both .
considerations, it has been estimated that about 200°C
should be used as a temperature limit for approximately 25%
of the salt volume. In the analysis of salt repository
layouts this limit has been equivalent to constraining the
peak temperature at the midpoint between canister positions.
The 200°C temperature limit is actually set by a plastic
flow analysis in which the calculated gross thermal expansion
of the salt field area would be offset by "collapsing"
of the void volume in the salt cavity itself. The specifics
of this balance will differ among designs and the calculations
94
-------
themselves are imprecise. This particular thermal 'limit
value may well be modified as salt mechanics become better
understood.
* Temperature rise in fresh water aquifers
Fresh water aquifers may be adjacent to or above the repository.
It is desirable that the temperature of these aquifers not rise
significantly; temperature rises of about 10°C at'a 30-m depth
and 30°C at a 90-m depth would be "acceptable."(28) Jt appears
that these values are achievable.
* Stability of the geologic formation when_subjected to
stresses caused by thermal expansion
Overall net surface expansion should not be so great as to create
geologic stresses that would cause a breach of repository
integrity. This concept is very important, because it directly
affects the planar heat density and, hence, mine size.
More effort is needed to determine the importance of such
a limit and to set appropriate values,
• Ground surface temperatures
It has been suggested that the average surface temperature of
the ground not exceed 0.6°C (1°F) above the ambient.(28) This
is easily achieved.
* Temperatures beyond the property boundary
Beyond the buffer zone, subterranean temperatures should be
limited to a 0.6^ rise.(28) fhe value appears to be quite
restrictive, requiring a fairly large exclusion radius;
furthermore, the rationale for choosing this low value
does not seem to have been developed in the literature.
When all of these factors are considered together, they result in a
planar heat input limit (for a salt repository) of about 150 kW/acre for
ten-year old HLW. ^°) -par HLW of a younger age, the value would be some-
what greater at the time of deposit, while for an older waste, the value
would be reduced appropriately. If the wastes are buried earlier, the
net effect would be to increase repository size. This increase relates
primarily to the "gross salt temperature limit," e.g., that 25% of the
total volume not exceed 200°C. The local temperature effects, e.g.,
the canister centerline temperature, or the 1% salt criterion, can be
met simply by different packaging approaches, such as use of a smaller-
diameter canister or a less dense heat source.
The value of 150 kW/acre was developed several years ago for a salt
bed approximately 305 m (1000 ft) below the surface. More recent studies
95
-------
have generally continued to assume the same or somewhat lower planar heat
densities. Because an independent repository design was beyond the scope
of the present study, the latest OWI design^- '' has been applied to the
460-m (1500-ft) repository depth. This design has a planar heat density
of about 126 kW/acre. If bulk salt mine temperature is limiting, the
planar heat density limit is relatively insensitive to repository depth
because of the very low rate of heat transfer to the surface. Heat
storage dwarfs heat transfer from the repository for several hundred
years. Hence, changes in depth, which affect the rate of heat transfer,
will have only a small effect on peak salt mine temperatures.
If uplift caused by thermal expansion should turn out to be the
limiting thermal criterion, planar heat density limits will be some-
what more sensitive to depth, since the amount of thermal expansion
depends upon the amount of total stored heat within the geologic forma-
tion. For a depth of 460 m, a maximum stored heat condition is reached
between 500 and 1000 years. Maximum uplift forces caused by thermal
expansion will be reached at the same time.
Based on calculations, the ratio of total stored heat for spent fuel
;vs HLW is in the range of 2-2.5 during the first 1000 years. The heat
production rate for spent fuel is higher than for HLW over extended
periods because of the larger quantities of long-lived transuranic
isotopes present in spent fuel. If it is assumed that the total stored
heat and the resulting physical impact on the geologic formation are
"limiting, then the planar heat density for a spent fuel burial field
aust be reduced accordingly. In order to estimate a reference mine
.layout and waste disposal facility costs (see Section 4.0), it has been
assumed for illustration that the planar heat density for a spent fuel
burial facility will be approximately 40% of that for a HLW facility.
Therefore, 2.5 times the total mine area needed for HLW would be
necessary to store an equivalent amount of spent fuel wastes. Investiga-
tions have not revealed any specific criteria for limits on thermal
expansion; it is possible that greater thermal expansion than has been
assumed in the OWI reference design may prove acceptable.
B-3.5.1.5.2 Mine Level General Arrangement
The mine level will consist of segregated storage areas for HLW,
spent fuel, cladding hulls, and TRU, a distribution system for ven-
tilation intake air, a network of ventilation exhaust tunnels, and mine
level support facilities. (A detailed description of the mine facility
will be found in Appendix B-IV; a conceptual sketch is shown in Figure B-18.)
The mine design will be in accordance with Federal health and safety
standards, "Metal and Non-Metallic "Underground Mines," 30 CFR 57, and other
appropriate mining codes and practices. The design must meet requirements
for ventilation, mine opening and mine cavity stability, the effects of
waste heat input and salt heat transfer coefficients on facility stability,
the efficient use of mining and hauling equipment during mine construction,
and the efficient placement of waste during mine loading.
96
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.X
LOW-UVEL BUILDING
"lOW-LIViL AND M/M SHAFT *
'
FIGURE B-18 ARTISTS CONCEPT - FEDERAL REPOSITORY
Source: Zerby, C.D. The National Waste Terminal Storage Program. Proceedings of Symposium
on Waste Management, Tucson, Arizona, October 1976, CORF 761020.
-------
Once a maximum permissible planar heat density has been determined,
based on heat transfer calculations, the waste canister disposal array
and canister-to-canister spacing can be established. There must be a
compromise between the most uniform heat transfer array (uniform spacing
of canisters throughout the unit burial volume) on the one hand, and mine
stability, mining economics, and ease of waste placement on the other.
These latter considerations require that the waste be placed in long
narrow tunnels. Heat transfer calculations must again be made for the
center-to—center spacing chosen in order to verify that the limiting
salt temperature criteria (Section 3.5.1.5.1) are still met. The total
heat loading per acre that is possible in the tunnel array may be slightly
less than that for a uniform canister array because of localized heating
effects.
For this study,HLW canisters are assumed to contain the wastes from
3 MTHM charged as fresh fuel to the nuclear reactors. For spent fuel,
nine canisters will be necessary to bury this equivalent amount of waste,
based on the assumptions that each canister contains one spent fuel
assembly. For a planar heat density that is 40% of that for HLW, approxi-
mately 3.6 times as many spent fuel canisters will need to be buried per
acre. Loading spent fuel canisters with two or more fuel assemblies per
canister may be possible, thus reducing canister handling operations
and final burial costs.
Adjacent burial tunnels run parallel to each other, with rows of
burial tunnels connected by a network of branch and main tunnels leading
back to the canister waste receiving station/transfer hot cell at the
base of the canister hoist. Mine capacity is increased by adding addi-
tional rows of burial tunnels. Tunnel and corridor dimensions are set
by the size of the waste canister placement equipment, with clearances
reduced to a minimum in order to minimize excavation costs. Corridors,
burial tunnels, and the network of ventilation exhaust shafts are exca-
vated using conventional mining equipment. The mined salt is removed
from the corridors and burial tunnels using conventional mine materials
hauling equipment that deposits the salt at the main shaft for transport
to the surface.
A reference bedded-salt mine general arrangement for storage/burial
of HLW, cladding hulls, and TRU has been developed. ' Based on a
review of the heat generation rate of ten-year-old HLW and on the
canister handling and placement operations, it appears that the canister
spacing and mine general arrangement given in Table B-13 is typical of the
layout that will eventually be used in commercial waste burial facilities.
This arrangement is assumed for HLW in this study as well.
The reference facility for spent fuel disposal will be similar in
design, with dimensions also summarized in Table B-13. Major differences
are that the tunnel and corridor heights must be raised approximately 1.5 m
(5 ft) to accommodate equipment for handling the larger waste canisters;
individual canister holes will hold four spent fuel canisters vs one
HLW canister (the alternative is four times as many holes per tunnel);
98
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TABLE B-13
Capacity (canisters)
Total MTHM charged (to reactor)
Specific heat Input
Canister density
Approximate canister dimensions
Canister hole dimensions
Number of canisters per hole
Corridor dimensions
Burial tunnel dimensions
Canister center-to-center spacing
Number of canisters per burial
tunnel
Number of burial tunnels
Burial tunnel center-to-center
spacing
Salt pillar thickness between
tunnels
Salt pillar thickness between
burial tunnels end to end
HLW/SPENT FUEL RETRIEVABLE STORAGE AREA DATA
HLW
35,500
1.07 x 105MTHM
150 kW/acre (Max)
126 kW/acre (most likely)
32 canisters/acre
0.32 m o.d. x 3.0 m long
0.46 m i.d. x 6.1 m long
1
9.1m wide x 5.5 m high
171 m long x 5.5 m wide x 5.5 m high
5.3 m
32
1112
23.8 m
18.3 m
18.3 m
Spent Fuel
319,500
1.07 x 105MTHM
60 kW/acre
51 kW/acre
115 canisters/acre
0.32 m o.d. x 4.6 m long
0.46 m i.d. x 23.2 m long
4
9.1 m wide x 7.0 m high
171 m long x 5.5 m wide x 7.0 m high
5.3 m
115
2786
23.8 m
18.3 m
18.3 m
-------
the spent fuel storage area will have 2.5 times as many burial tunnels
to accommodate the equivalent amount of waste; the cladding portion of the
TRU burial area will be eliminated; and the volume of the TRU burial
area will be reduced in size, because this type of waste will be
drastically reduced in volume.
Dimensions for the TRU areas are shown in Table B-14. While not
optimal, they are sufficient for the present.
B-3.5.1.6 Total Reference Facility Area Estimates
Table B-15 shows the area breakdown for the two reference facilities
cited in this study. The total controlled area around the surface faci-
lity is assumed to be a distance of one mile radially outward from the
outer edge of all waste burial fields. *• ' This distance was arbitrarily
set to ensure that the thermal effects at any point outside the controlled
area, either on the surface or underground adjacent to the buried waste,
do not result in a temperature rise of more than 0.6°C. Based on a pre-
liminary heat transfer analysis, this would require a distance of approxi-
mately one mile from the edge of the burial field.
B-3.5.1.7 Summary of Major Differences Between the HLW and Spent Fuel
Burial Facility Reference Designs
The general equipment design, surface facility structures, and mine
layout concepts are similar. Major differences are in dimensions and in
the number of waste units that must be handled. These differences are
summarized below:
(1) Planar heat density. Spent fuel will contain more actinides
(see Task A report), leading to a greater integrated heat
source over the stored life of the waste. Thus, the initial
mine loading must be reduced; a reduction factor of 2,5 (i.e.,
from 126 kW/acre for HLW to 51 kW/acre for spent fuel) would
allow for the added heat production over the first 1000 years.
(2) Canister sizings. HLW canisters will be 3 m (10 ft) long, and
spent fuel canisters will be approximately 4.6 m (15 ft) long.
This extra length will change the size of shipping casks, sur-
face handling facilities, mine facilities, and mine handling
equipment.
(3) Excavation requirements. More than three times as much material
must be mined for a spent fuel disposal facility as for a HLW
facility.
(4) Number of canisters. As many as nine spent fuel canisters
must be buried to account for the equivalent, amount of waste
in a HLW canister. Burial of four spent fuel canisters per
100
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TABLE B-14
REFERENCE FACILITY DIMENSIONS FOR TRU WASTE
Cladding Hull/Intermediate-level TRU Burial Field Data
Waste container
Canister dimensions
Capacity - cladding hull canisters
Capacity - TRU canisters
Burial room dimensions
Number of canisters per room
Canister spacing
Number of canisters per hole
Spacing between burial rooms
Number of burial rooms
Low-Level TRU Burial Field Data
Waste container
Capacity, TRU drums
Burial room dimensions
Type of storage
Number of drums per room
Number of rooms
HLW Facility*
Spent Fuel Facility
canister
0.3 m i.d. x 4.6 m long
0
canister
0.3 m i.d. x 3.0 m long
58,500
255,000 less than 130,000+
171 m long x 11.6 m wide x 5.5 m high 171 m long x 11.6 m wide x 7.0 m high
790 790
6 rows, 1.2 m center to center 6 rows, 1.2 m center to center
1 1
23.8 m center to center 23.8 m center to center
398 165
HLW Facility
Spent Fuel Facility
55 gal drum 55 gal drum
655,000 less than 330,000+
171 m long x 11.6 m wide x 5.5 m high 171 m long x 11.6 m wide x 5.5 m high
Warehouse Warehouse
7,000 7,000
94 47
Source: Parsons, Brinckerhoff, Quade & Douglas, Inc., Waste Isolation Facility Description,Y/OWI/SUB-76/16506,
Preliminary estimate.
Office of Waste Isolation, 1976.
May not represent optimized design.
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TABLE B-15
TOTAL FACILITY AREA REQUIREMENTS
HLW Facility Spent Fuel Facility
(Acres) (Acres)
Total controlled area 10,600 12,000
Total area containing waste* 2,000 3,487
HLW/spent fuel 1,112 2,786
Cladding hulls/Intermediate-level 398 165
TRU
Low-level TRU 94 47
Mine shaft exclusion area,
corridors service area, etc. 400 500
*Includes the area of tunnels plus area between tunnels.
102
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hole can reduce the number of holes to 2.5 times the number
required for HLW.
(5) Numberofcanisters perunit burial tunnel. As many as 3.6
times as many spent fuel canisters (9/2.5) as HLW canisters
will be buried in each unit burial tunnel. This can be accom-
plished either by 1) boring 3.6 times as many canister holes
(probably resulting in the need for wider tunnels); 2) multiple
canister loadings in the same number of canister holes (drilled
deeper) as used for HLW; or 3) some combination of each.
(6) Impactof TRU wastes. For the throwaway cycle, much less TRU
waste is generated, reducing the mine volume needed. The
reduction is offset to some extent by the greater number of
spent fuel canisters than HLW canisters.
B-3.5.1.8 Assessment of the State of the Art for Burial in Salt
Heat transfer characteristics and the behavior of salt at elevated
temperatures can be readily predicted. Within the thermal criteria
chosen (no more than 11 of the salt over 250°C, and no more than 25%
over 200 C), general mine stability remains high and local heating around
the canister permits enough plastic salt flow to seal the canister in
the salt formation. Much practical experience has been gained in the
Project Salt Vault Studies.^29^
Mine construction and handling equipment needs appear to be well
within current technology and should prove to be straightforward engineer-
ing jobs. Salt mining can be accomplished using conventional mining
techniques. Canister-handling equipment must perform relatively simple,
straightforward operations, and although all handling must be done by
remote control, equipment design should be well within the current
capabilities of the nuclear industry.
B-3.5.1.9 Facility Design Refinements
The reference facilities in this study represent typical arrange-
ments, but it is not suggested that they represent an optimized design
in every respect. Several areas where further studies may lead to a
less costly facility are suggested below:
* Optimize age of waste to permit balancing higher mine-loading
density against cost of longer-term surface storage.
* Consider loading low-level TRU in HLW tunnels before backfilling
the tunnels, eliminating the need for a low-level TRU storage
field.
103
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* Consider the use of bulk shipping containers for low-level TRU,
as alternatives to drums.
* Evaluate the use of self-shielded intermediate-level TRU bulk
shipping containers instead of canisters. Bulk containers could
be handled directly, using conventional freight-handling equip-
ment, warehouse-style, in the mine. The number of containers
to be handled can be reduced by a factor of 10,
* Evaluate the possibility of loading more than one spent fuel
assembly In each canister to decrease the number of holes that
must be bored in the burial tunnels.
B-3,5.1,10 Retrievable Storage
The reference facilities discussed to this point have assumed per-
manent, non-retrievable waste storage. Under certain circumstances,
however, it may be desirable to store the wastes in an arrangement that
is suitable for permanent disposal, while retaining the option to retrieve
the wastes within a predetermined period of time. Retrievability might
be useful for the pilot waste burial facility, in order to facilitate
periodic canister inspection and testing. Retrievability may also be
useful for spent fuel burial, although this might better be done in
near-surface facilities. Although the present policy on reprocessing
of spent fuel is unsettled , it seems unlikely to many people that spent
fuel, with its intrinsic value as an energy resource, will actually be
"thrown away", but rather that spent fuel would be stored in a facility
from which it might be recovered and reprocessed at a later date.
The general arrangement for a retrievable deep storage facility would
probably be similar to that for non-retrievable deep storage in many
respects. The surface facility and the mine access shafts perform the
same functions and will be identical. The general mine layout will also
be similar, with canisters of waste buried in holes in the floor of long
narrow burial tunnels. Mine stability and integrity over the full length
of the storage period must be verified prior to setting the canister spac-
ing and tunnel dimensions for the particular geologic medium, but spacing
and dimensions are expected to be similar for both retrievable and non-
retrievable storage.
There may be design differences in the following areas:
* Mine Ventilation - The system for a retrievable facility may
have to be designed for a higher capacity, because a larger
unfilled mine volume must be ventilated over a much longer
period of time. The ventilation system may also be used to
remove decay heat from the mine over the retrievable period
as a measure to further enhance mine stability.
104
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Mined Materials Removal System - Much greater quantities of
mined salt must be removed from the mine and stored on the sur-
face for a retrievable facility. For non-retrievable storage,
once a working mine volume has been created» salt from a tunnel
under construction is used as backfill for a tunnel already
loaded with waste. In this way, the total volume of salt trans-
ferred to the surface is minimized. For retrievable facilities,
all the excavated material must be transferred to the surface,
stored for the desired period of waste retrlevabllity, and then
returned to the mine as backfill material during the facility
decommissioning period. For the reference facilities assumed
for this study, approximately 37 x 10° vr must be removed from
a retrievable spent fuel burial facility while as little as
2.5 x 10 m must be removed from an equivalent permanent spent
fuel burial facility.
Canister Burial Hole Arrangement - For non—retrievable storage,
a canister will simply be placed in a hole bored in the "floor of
a waste burial tunnel and the hole will be backfilled up to floor
level with crushed salt to provide radiation shielding to the
operators. To facilitate retrlevability, however, the canister
hole design must be changed. The canister hole will probably have
to be lined to prevent plastic flow in the salt from sealing the
canister in the salt, and have to be sealed with a concrete plug
to provide shielding.- Carbon steel 25 cm (1 in) thick is sug-
gested as a liner material in order to maintain enough integrity
over the desired period to prevent the surrounding salt from
collapsing around the canister.^ ' A minimum of five to six
feet of salt thickness will probably be required to provide
adequate shielding (less than 10 mR/hr) at the surface of the
tunnel floor for 10-year old wastes.
Waste Placement Equipment - Additional equipment, such as a
shielded canister hole isolation valve and a canister hole plug
installation machine may have to be developed to permit waste
placement in retrivable storage.
Canister Material - Retrievability may require that the canister
material be upgraded to one with long-term corrosion resistance
(perhaps titanium).
Facility Decommissioning - For non-retrievable storage, tunnels
are backfilled with mined materials from other burial tunnels
under construction. Thus, ultimate facility decommissioning
consists simply of surface facility decommissioning, with con-
taminated materials from the surface facility placed in open
spaces, such as mine corridors, followed by backfilling of the
mine and, finally, of the mine shaft, as determined by regula-
tory policy at the time of decommissioning. Decommissioning of
a retrievable facility will be a more costly operation.
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B-3.5.2 Disposal In Other Media ,
B-3.5.2.1 Summary Discussion of Thermal Criteria
To date, mbst of the detailed design work on geologic disposal
facilities has been limited to salt deposits, but alternative geologic
media are of interest for the reasons discussed at the beginning of
Section 3.5. Crystalline rocks, such as basalt or granite,.-appear to
b^ of greatest interest, although consideration is also being given to
shales and other sedimentary rocks.
Because of lack of detailed engineering data, it has not been possible
to perform a thorough assessment of disposal technology in media other
than salt. This section will, therefore, discuss only briefly certain aspects
of facility design that might have a bearing on. the integrity of a reposi-
tory in such media as granite or basalt. Some limited engineering data
are available for crystalline rock repositories, but there seem to be
essentially no data on other rock types, such as shale, and therefore
they will not be considered further.
The importance of thermal criteria has been pointed out in the re-
view of salt repository designs and, clearly, they will also have a sig-
nificant impact on facility design in all geologic media. For this
reason, scoping analyses of thermal characteristics have been performed
for granite as well as for salt (Section 3.5.3). In reviewing thermal •'
design criteria, one of the first questions that arises is whetfher there
are analogies between the "defined" criteria for salt and the potential
criteria for granite. A comparison of the two media is given below,
* •Temperature limitations immediately adjacent to the storage
canister. The limit- could possibly be higher for granite than
for salt, since corrosion should be much less of a problem in
granite- and because plastic, flow does not occur in granite as it
does in salt. On the other hand, the lower thermal conductivity
of granite leads to steeper temperature gradients than in salt.
Such gradients could cause localized cracking and/or spalling in
the vicinity of the canister. For non-retrievable storage, a
limited fracture zone should not present serious problems; this
would not be the case for retrievable storage, however. In any
case, the extent of such fractures could be controlled by reduc-
tions in canister size and/or heat release per canister.
* •Temperature limitations throughout the mined area. The problems
are of the same nature in either medium; the primary concern is
stability of the repository. Granite is a less satisfactory
medium in this respect, because of the absence of significant
plastic flow and the associated intolerance of significant tem-
perature gradients. To set thermal criteria, a detailed analysis
of the rock mechanics is necessary. Some work on granite reposi-
tories has been reported in Canada including at least a partial
analysis of rock,mechanics in the mined area, indicating that
106
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structural integrity can probably be achieved at bulk temperature
limits somewhat less than those developed for salt.' '
* Stability of the geologic formation, when subjected to stress •
caused by thermal expansion. As already noted in the discussion
of salt repositories, criteria on maximum permissible uplift are
highly tentative and probably subject to evolutionary changes as
more work is performed on rock mechanics. The Canadian studies
do not appear to have considered the detailed mechanics of the
thermal expansion problem, which is very complex in any case,
and especially for granite where there is little plasticity and
for which the fracture mechanics are imperfectly understood.
Conservative scoping calculations of thermal uplift (to be
discussed below) lead to modest values of 1.2 m (4 ft) at 150 kW/
acre. These scoping calculations cannot allow for the nature and
extent of cracking that might result from any given pattern of
uplift.
The remaining four considerations—a) canister centerline temperatures,
b) temperature rise in fresh water aquifers, c) ground surface tempera-
tures, and d) temperatures beyond the property boundary, are relatively
independent of type of geologic medium.
B-3.5.2.2 Canadian Studies on Granite Repositories
Atomic Energy of Canada Limited (AECL) has been considering rock
repositories for some time, and has funded research and engineering
programs to calculate rock temperature distributions as well as the re-
sulting stresses and failure mechanics associated with these temperature
distributions. A recent study by Acres Consulting Services Limited
(Acres) covered analysis of temperature distribution and failure modes
in a granite repository designed for retrievable spent fuel storage over
a time period of 30 years.(30) This study utilized a planar heat density
of about 130 kW/acre keyed to a five~year-old waste. Allowing for differ-
ences in fuel characteristics and adjusting the 130 kW/acre figure for a
ten-year age results in an equivalent planar heat density of 80-90 kW/acre
or about 60% of the nominal level for HLW in salt (150 kW/acre). The
granite depth in the Acres study was at the 1000-m level (more than
3000 ft) compared with the reference depth in this report of 460 m
(1500 ft). At a 1000-m depth, the resulting allowable planar heat
density is less because of greater stored energy in the repository prior
to reaching equilibrium heat-transfer conditions. (At greater depths,
longer times are required to reach the peak temperature condition.)
For the stress analysis, a two-dimensional finite-element computer
program was used for steady-state and transient calculations. In carry-
ing out these calculatiohs, certain assumptions had to be made regarding
the (a) rock joints and their strength; (b) mechanism and criteria for
failure; (c) non-linear rock behavior under long-term mechanical loads;
(d) stabilization pressures as related to long-term stability; (e) in-
107
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fluence of pre-existing rock stress. These assumptions were necessary
because of Inadequate physical data and/or model limitations. To com-
pensate for uncertainties, the investigators used conservative input
parameters.
Temperature and stress contours were developed and resultant fail-
ures were presented in a series of sectional views. This paper con-
cluded that although minor failure occurred at the bottom and top corners
of the repository tunnels, these failures were not likely to propagate
far into the rock, and the basic integrity of the whole structure would
probably be maintained. On the other hand, anchors might be required to
prevent collapses of the top wall in certain areas if it were important
to minimize local build-up of rubble. The degree of failure during the
first 30 years would be to a very large extent Influenced by the amount
of ventilation and cooling that occurred during this period,which is
characterized by high thermal gradients.
The foregoing remarks applied to the mine structure above the tunnel
floor. For the floor area containing the waste canisters, it seems prob-
able that cracking would occur both adjacent to the canister and above
it. Such cracking could make subsequent retrieval or removal of the
canister difficult. It was therefore tentatively concluded that for a
retrievable facility these holes would have to be lined.
The conclusions of the study were summarized by the authors:
(1) Failure of the jointed rock mass around a ventilated room is
restricted to the portions of the room below the springline.*
The requirements for supporting these regions are minimal.
(2) The floor area needs further engineering to ensure retriev-
ability of the waste canisters. Detailed analyses of the
stresses and the displacements in this region should be performed
using a discontinuum model and field data (heater experiments,
geotechnical investigations).
Although much more analysis and information are needed to design a
granite repository, the Acres study suggests that such a repository is
feasible.
B-3.5.2.3 Possible Thermal Criteria for Granite
In any given medium, temperatures in the immediate vicinity of the
canister are controlled mainly by canister size and heat output. If
localized cracking is deemed to be a problem for non-retrievable storage,
it can be controlled by using smaller-diameter canisters and/or lower
heat release per linear foot of canister. For the same planar heat
The line of intersection of roof and walls,
108
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density, lower canister heat putput would necessitate closer spacing,
with more holes and therefore higher costs. Presently, such parameters
cannot be estimated with any precision because of the lack of detailed
data on localized cracking from high thermal gradients in granite. For
the purpose of this study, there seem to be no inherent feasibility prob-
lems with further subdivision of wastes and/or lining of waste holes if
this should prove necessary.
As for temperature limits in the mine itself, it would appear from
the Acres study that a modest reduction from the planar heat density
assumed for salt in this report would possibly be sufficient to assure
satisfactory integrity of the mine area proper. The planar heat density
also affects gross thermal expansion and uplift of the repository area,
however. The thermal analysis presented below shows that temperature
distribution outside the repository does not differ greatly between
salt beds and granite. These temperature distributions are important in
determining the thermal expansion characteristics of the repository.
It is possible that planar heat density may be set by limitations
on thermal expansion rather than on mine temperature. Vertical thermal
expansion, i.e., the rise in the surface ground elevation compared
with its original level, may be especially important if it results in
connected fractures of the granite. Table B-16 provides perspective on
these effects, based on a very simple expansion model. This table is
based on a planar heat density of 150 kW/acre. The indicated expansions
are gross expansions and any subsidence (collapse of the mine void are%)
would have to be subtracted from the gross expansion figures.
The reference salt bed repository indicates a gross expansion of
2.8 m. This value conservatively assumes that the three-dimensional
volumetric expansion characteristics of the salt and the overburden all
act in the vertical direction and therefore the vertical expansion
coefficient is effectively three times the linear expansion rate. This
is perhaps not an unreasonable assumption for salt because of its high
plasticity. It is probably an overly conservative assumption for a
crystalline rock where horizontal expansion may not be converted to
vertical expansion unless there are suitable slip planes or connected
fractures within the gross rock structure. Nevertheless, the table
indicates that the granite repository would hav?. an expansion approxi-
mately one-third that of the bedded salt.
In comparing the calculated thermal uplift data with other published
results, there is reasonable agreement when corrections are made for diff-
erences in depth and planar heat densities. For example, the Culler
Report indicates a thermal expansion of about 1.5-1.8 m (5-6 ft) and a
subsidence of about 1.2 m (4 ft), with these phenomena occurring about
200-300 years from deposition.(31) The calculated value is higher and
occurs somewhat later in time because of the greater depth of the refer-
ence repository, but it is consistent with the Culler Report data. This
good agreement lends confidence to the preliminary calculations for granite
repositories.
109
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TABLE B-16
CONSERVATIVE ESTIMATES OF GROSS VERTICAL EXPANSION
("THERMAL UPLIFT") FOR DIFFERENT REPOSITORY MEDIA
Time of Maximum ^
Stored Heat Expansion
Salt bed 1000 years 2.8 meters
Granite 1000 1.2
Salt Dome 500 7.6
Assumes that all of volume expansion is in vertical
direction (3 times linear expansion coefficient), i.e.
an ideally plastic material. Linear coefficient assumed
to be 40 x 10~6/0C for salt and 7.9 x 10~6/°C for rock.
110
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An arbitrary limit on planar heat density one-half that of the
corresponding values for salt, should be a sufficiently conservative
assumption to allow for uncertainties in characterizing granite reposi-
tories. Uplift would be calculated to be less than 0.6 m (2 ft) and peak
mine temperatures would be significantly less than those calculated in
the Acres study. It is, therefore, an appropriate assumption for the
purposes of this study, although it should be considered only as a "best-
guess" at this time, recognizing current state-of-the-art limitations
on design of repositories in granite.
B-3.5.3 Thermal Analysis
B-3.5.3.1 Background
Thermal design criteria for a HLW repository have a major impact on
the size and geometry of such facilities. Suggested thermal design(28)
criteria for a "bedded salt" repository at a depth of approximately 305 m
(1000 ft) have been published and are summarized in the discussion of
the reference salt repository design (Section 3.5.1).
It is beyond the scope of this study to perform a detailed thermal
design of a repository; nevertheless, some independent calculations have
been made in order to demonstrate the effects of variables such as time,
medium, waste age, planar heat density, and waste type. Broad conclusions
may then be drawn regarding the effects of different waste management
plans and standards on repository design.
Although no criteria have been established for the design of a rock
repository facility, an understanding of the temperatures involved can
give some insight as to what densities of storage might be attainable.
Three different types of thermal analyses have been performed, each
of which fills a specific need. These are:
1. A three-dimensional (3-D) analysis run in order to determine
the short-term temperature behavior near the waste containers.
This analysis is necessary to determine if the "l%-salt-at-250°C"
criterion is met. It is at this time that maximum thermal
gradients exist.
2. A one-dimensional (1-D) analysis run in order to determine
the long-term temperature distribution gradients in the vertical
direction, both above and below the repository area. It is from
this temperature distribution that the maximum integral energy
deposit can be calculated and from which the maximum possible
thermal expansion can be estimated. This temperature distribu-
tion is a function of the type of rock as well as of the depth
of the repository. If uplift from thermal expansion should
become a limiting factor, planar heat density would be set by
this temperature distribution.
Ill
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3. A two-dimensional (2-D) analysis run in order to determine
the added heat transfer that will occur from the repository
area by virtue of radial heat flow and to establish if this
effect has much of a bearing on the temperature at the
vertical centerline of the facility. This analysis is also
useful in determining the temperature distribution beyond the
facility itself in order to set buffer zone distances.
The initial planar heat density (kW/acre) is treated as a parameter
in the calculations as is the age of the wastes at time of deposit.
Calculations have been performed for three alternative repository
concepts: a bedded salt reference case, a salt dome case (similar to
bedded salt except that the salt zone is much thicker in the vertical
direction), and a granite repository. All three cases assume the HLW
deposit to be at the 460-m (1500-ft) level.
B-3.5.3.2 Summary of Results
The results of the analyses are presented in a series of graphs
that display the temperature distributions for various points in time
and space, and for the assumed variation in input parameters (planar heat
density and age of wastes). The analytical model and calculational re-
sults are described in detail in the remainder of this section, but a
few general comments are in order:
1. Considering the repository as a whole, the temperature
distributions developed for a given planar heat density
do not differ greatly for bedded salt, salt domes, or
granite media. This may seem surprising, since salt
has generally been thought to be superior because of
its higher thermal conductivity; the lower conductivity
of rock is compensated for, however, by a significantly
higher volumetric heat capacity.
2. The local peak temperature in granite for a given initial
planar heat density (e.g., 150 kW/acre) and canister
energy (e.g., 3.95 kW/canister) would be substantially
greater than in salt. This may necessitate the burial
of less waste per canister, but with a closer canister-
to-canister spacing for granite.
3. The temperature distributions as developed for spent
fuel disposal are not dissimilar to those of HLW. This
is because later, when the higher long-term energy pro-
duction from spent fuel is present, the heat transfer
to the atmospheric heat sink has already started to dis-
sipate the stored energy. This is the case for a 460 m
(1500 ft) repository level; it may not be so for much
deeper repositories.
112
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4. A one-dlmenslonal analysis on the vertical centerline of
the field is a very good representation of the temperature
distribution and two-dimensional heat transfer becomes
important only over very long time frames, i.e., longer than
10,000 years.
5. The peak temperature rise in the mine is roughly proportional
to the original planar heat density. For example, with ten-
year-old waste at 150 kW/acre the maximum temperature rise in
salt is approximately 260°C, but in the same repository at a
density of 60 kW/acre (40%), the temperature rise is 90°C (35%)
6. The calculated temperature distributions in bedded salt are in
good agreement with those from other studies.* * •'
3-3.5.3.3 Modeling of the Repository
For these calculations, in accordance with the reference design,
the waste repository was assumed to be 460 m (1500 ft) below the
ground surface. The canister spacing was 23.8 m (78 ft) apart in the X
and 4.3 m (14 ft) in the Y direction in a rectangular array. Figure B-19
is an isometric sketch of the layout used in the 3-D computer simulation.
Since a rectangular coordinate system is used in the calculational model,
the cylindrical waste container was simulated by a container having a
square cross section of the same size as the cylindrical container,
therefore, the surface area of the container is increased by about 13%.
For the 1-D analysis the heat source was assumed to be homogeneously
distributed in a 30.5-tn (100-ft) thick layer of salt or granite. For
the 2-D analysis the waste field was assumed to be a "pancake" 30.5 m
(100 ft) thick and having a diameter of 1190 m (3900 ft).
The values of the thermal properties for salt, shale, granite, ,„_ ,_
and sedimentary deposits used in the analysis are given in Table B-17. * *
The thermal conductivity of salt and granite, and the specific heat of
granite were assumed to be a function of temperature as shown in
Figure B-20. All other properties were assumed to be constant with
temperature.
The computer code HEATING-3 was used to perform the 3-D, 2-D, and
1-D analyses. (34,35) fo reduce computer running time, the explicit solu-
tion method was utilized. ^6) The time-dependent heat generation rates,
as used for the analysis, are given in the Task A report for the throwaway
and the mixed-oxide-recycle fuel cases respectively.
B-3.5.3.3.1 Results of the 1-D Analysis
The objective of these computer runs was to analyze the long-term
temperature behavior of the geologic formation above and below the mine
containing the waste canisters. Three different geologic formations were
considered, as given in Table B-18.
113
-------
Z-Coordinate
1,500ft
Below Ground
Surface
Region Considered
in 3-D Analyst!
Y-Coordinate
FIGURE B-19 GEOMETRICAL ARRAY USED FOR THERMAL ANALYSIS
114
-------
TABLE B-17
THEBML PROPERTIES OF GEOWGIC
Salt
Granite
Shale
Surface Deposit
Density
g/cm_
2.2
2.7
2.4
2.4
Specific
Heat
Gal
0,22
0.21 r 0.26
0.20
0,20
Thermal
Conductivity
Cal
cm-hr-°C
18-52*
22-31*
14.
23.7
Volumetric
Heat
Capacity
Cal
cm3_°G
0.48
0.58-0.71*
0.48
0.48
* Temperature-dependent
115
-------
60
o
o
(3
> 30
>
"s
•o
so
1
I
100 200
Temperature {°C)
300
FIGURE i-20 ASSUMED VALUES FOR THERMAL CONDUCTIVITY VS. TEMPERATURE
116
-------
TABLE B-18
GEOLOGIC FORMATIONS AT DEPTHS IN MODEL
Vertical I II
Positions Bedded Salt Granite
Meters
0- 150
150- 410
410- 500
500-1520
1520-2430
Feet
0- 500 "Surface" "Surface"
500-1350 Shale Granite
1350-1650 Salt Granite
1650-5000 Shale Granite
5000-8000 Shale Granite
III
Salt Dome
"Surface"
Salt
Salt
Salt
Granite
Note; In all cases, heat source at 460 m (1500 ft)
117
-------
Parameters In this study were the burial time, the type of waste
(throwaway vs recycled mixed oxide) and the planar heat generation rate
at burial time. Cases that were analyzed and their parameters are
summarized in Table B-19. The case numbers, such as I-A, III-B, or
II-C, are keyed to the indicated geologic formation (Table B-18) and
the corresponding heat source (Table B-19).
The results of Cases I-B, II-B, and III-B are given in Figures B-21,
B-22, and B-23, showing the temperature rise vs vertical position at 10,
100, 1000, and 10,000 years after burial at a power density of 150 kW/acre
in all three geologic, formations. The temperature profiles are very simi-
lar for all three cases. The highest temperatures are reached between
100 and 300 years, as shown in Figure B-24. The highest temperature was
predicted to occur in bedded salt (270°C). The maximum temperature
In the granite and in the salt dome was predicted to be about 10 C lower.
The slightly lower peak temperature in the granite vs the bedded salt is
due to the larger volumetric heat capacity in the granite, which more
than offsets the smaller thermal conductivity of granite. The tempera-
ture profile above and below the waste containers stays symmetric until
the ground surface temperature increases slightly and heat starts flow-
ing from the surface to the atmosphere. A heat transfer coefficient of
0.4 cal/(hr - cm2 - °C) or 0.8 Btu/(hr - ft2 - °F) on the surface was
assumed for the analysis. The surface temperature increased by less
than 0.2°C (0.3°F), and the maximum was reached between 1000 and 3000
years after burial.
The effect of burial time on maximum salt mine temperature is given
in Figure B-2S, i.e., Cases I-B, I-C, and I-D. It shows the maximum
temperature vs time in a 9Q-m (300-ft) salt layer after burial of 10-,
20-, and 50-year-old waste (recycled mixed oxide) with a power density
of 150 kW/acre at the ten-year point. The maximum predicted temperatures
are about 250? 190° and 120°C for the three cases, reached in 200-300
years.
The effect of burial time in granite is shown in Figure B-26 for
Cases II-B and II-C. The maximum mine temperature is about 60°C lower
for burial of 20-year-old waste compared with burial of 10-year-old
waste. The same figure also shows the effect of lower initial power
density on temperature rise. Cases II-E and II-F are similar to Cases
II-B and II-C, but power density has been reduced for these two cases
from 150 kW/acre to 60 kW/acre at the ten-year point. For the lower
power density cases, the maximum temperature is reached at about the
same time as for the high power density cases. The temperature rise is
approximately proportional to the initial power density. For Case II-B,
the temperature rise vs vertical position at various times is shown in
Figure B-22, while for Case II-E (60 kw/acre) it is shown in Figure B-27.
The last sensitivity run shows the effect of the type of waste on
the temperature rise in a salt deposit, with an initial power density
of 150 kW/acre at the 10-year point. For throwaway-cycle waste, the
predicted maximum temperature is about 60°C higher than for the recycled
118
-------
TABLE B-19
PARAMETERS AFFECTING THE TIME-DEPENDENT
HEAT GENERATION RATE
Case A B
Waste
Time
Years
Heat (
Rate i
Point
Throwaway Recycled
Cycle Mixed Oxide
of Burial, 10 years 10 years
after Discharge
generation 15Q fcw 15Qkw
acre acre
CD E F
Recycled Recycled Recycled Recycled
Mixed Oxide Mixed Oxide Mixed Oxide Mixed Oxide
20 years 50 years 10 years 20 years
150 kW 150 kW 60 kW 60 kW
acre acre acre . acre
-------
s."
D.
S
H
A
L
E
SALT
T>
X.
s.
e
Q
S
H
A
L
E
Surface
0
150
410
iOO
2,430
100y
Note;
Based on 10-year-old reprocessed
waste with planar heat density
of 150 kW/acre at burial time.
Waste at 460m {1500 ft),
Sedimentary deposits.
100 200
Temperature (°C)
300
FIGURE B-21 VERTICAL TEMPERATURE DISTRIBUTION IN
TYPICAL BEDDED SALT FORMATION
120
-------
ro
*-*
ro
w
4-J
Vi
"8
a
•o
IB
a
s."
D.
G
R
A
N
I
T
E
Based on 10-year-old reprocessed
waste with planar heat density
of 150 kW/acre at burial time.
Waste at 460m (1500 ft).
Sedimentary deposits
1,520
2,430
100 200
Temperature (°C)
300
FIGURE B-22 VERTICAL TEMPERATURE DISTRIBUTION IN TYPICAL
GRANITE FORMATION <150 kW/ACRE)
121
-------
2
35
>*"
o
a
•a
m
f
01
a
S.'
0,
S
A
L
T
G
R
A
N
I
T
E
Based on 10-year-old reprocessed
waste with planar heat density
of 150 kW/acre at burial time.
Wastes at 460 m (1500 ft>.
Sedimentary deposits.
1,520
2,430
0
Temperature (°CJ
FIGURE B-23 VERTICAL TEMPERATURE DISTRIBUTION IN TYPICAL SALT DOME
122
-------
300
Granite Case II-
200
fi
3
I
100
10
Bedded Salt Case I-B
Based on 10-year
processed waste
I i i t i i i i
150kW/acree
time. Wastes at
460m (1500 ft).
I I I 1
j i
100
1,000
Time (years)
FIGURE B-24 MAXIMUM MINE TEMPERATURE VS. TIME IN THREE
GEOLOGIC FORMATIONS (REPROCESSED WASTE)
10,000
123
-------
300
200
o
q__
ff
1
S3
100
10-
10
/
/
/20-
,'50-Year-Old Waste
Note: Horizontal spacing same ^
as previous figure (150
kW/acre at ten-year decay
time)
i ii i i i i i I i i i i r i i L
100 1,000
Time (years)
10,000
FIGURE B-25 MAXIMUM MINE TEMPERATURE VS. TIME FOR 10-.
20-, AND 50-YEAR-OLD WASTE IN BEDDED SALT
124
-------
300
200
o
o.
I
+*«
03
I
E
a;
100
Casell-B
Horizontal Spacing Keyed to
150 kW/acre at 10-Year Point
(Reprocessed Wastes)
1 I I 1 I
Horizontal Spacing Keyed to
60 kW/acre at 10-Year Point
(Reprocessed Waste)
I ill! MIL
1 i Mill
10
100
1,000
10,000
Time (years)
FIGURE B-26 MAXIMUM MINE TEMPERATURE VS. TIME FOR
10- AND 20-YEAR-OLD WASTE IN GRANITE
125
-------
Surface
S.*
D.
o
I
TJ
C
Q.
Q>
O
G
R
A
N
I
T
E
0
150
410
1,520
2,430
100y
10,000y
Note;
Based on 10-year-old reprocessed
waste with planar heat density
of 60 kW/acre at burial time.
Wastes at 460 m < 1500 ft).
I
J_
100 200
Temperature (°C)
300
FIGURE B-27 VERTICAL TEMPERATURE DISTRIBUTION IN TYPICAL
GRANITE FORMATION (60 kW/ACRE)
126
-------
mixed oxide waste. As shown in Figure B-28, the maximum value is reached
at about 170 years for the recycled mixed-oxide waste vs 330 years for
the throwaway waste. The higher maximum temperature is due to the fact
that the decay heat is sustained for a longer time for the throwaway
waste than for the recycled mixed oxide waste.
B-3.5.3.3.2 Results of 3-D Analysis
The objective of the 3-D analysis was to analyze the local tempera-
ture distribution and gradients in the salt or granite around the waste
containers as a function of time. Because the steepest temperature gradient
will occur shortly after burial, since heat flux is highest at that
time, and because of the high cost of long-time 3-D calculations, the
cases were terminated at the two—year point after burial. Three cases
with varying power densities and host materials were analyzed. Heat-
generation rates of 3.95 kW/canister and 1.58 kW/canister were used, cor-
responding- to the power densities of 150 kW/acre and 60 kW/acre, the values
used in the one-dimensional analysis. The given heat production rates
correspond to ten-year-old waste of recycled mixed oxide at the time of
burial. The repository material was either salt or granite. The canister
was assumed to be buried in the rectangular array shown in Figure B-19.
The vertical limit for this 3-D analysis was a 460-tn (150-ft) layer above
and below the waste container, sufficiently large to cause an insigni-
ficant error in the temperature calculation during the time period covered.
The results of the first case, salt with an assumed power density
3.95 kW/canister at burial time, are shown in Figures B-29 to B-31.
Figures B-29 and B-30 show the temperature gradient in the salt along the
three axes after burial. After two years, the temperature gradient close
to the canister is very similar in the X and Y direction and about a
factor of two smaller in the Z (vertical) direction. The hot spot after
one year is 110°C. Figure B-31 shows the isotherms in the salt
at canister mid-plane (Z=0). The isotherms are circular close to the
canister, as expected. Farther away, however, the isotherms are parallel
to the Y axis.
Figures B-32, B-33, and B-34 show the results of the analysis
representing the waste canister buried in granite. The heat production
rate was identical to the first case. Less than 3 m (10 ft) from the
canister, granite temperatures are significantly higher, and the gradient
is about a factor of three larger than in salt. The gradients in the X
and in the Y directions are almost identical and a factor of two smaller
than in the Z direction. The highest granite temperature at the canister
surface is 169°C.
The much higher close-in temperatures for granite than for salt
result from substantially lower conductivity (see Table B-17). Farther
away, where heat flux is lower and heat storage is more important, the
better heat storage capacity of granite (see Table B-17) compensates for
127
-------
300
200 -
o
o
D
Q.
CJ
S Throwaway Cycle
Plutonium Recycle
// (ReprocessedWastes)
\ (spent fuel)
Note: Horizontal spacing keyed to
150 kW/acre at 10-year point.
100 -
100
1,000
10,000
Time (yearsS
FIGURE B-28 MAXIMUM MINE TEMPERATURE VS. TIME FOR 10-YEAR-OLD
WASTE IN BEDDED SALT FORMATION
-------
150
100
NJ
O
o__
-------
150
o
£
3
a
V
50
Note:
Spacing keyed to 150
acre measured at the ten-
year point. Reprocessed
wastes with canister heat
output of 3.95 kW.
10
20
30
40
Distance {m)
FIGURE B-30 VERTICAL TEMPERATURE DISTRIBUTION IN
SALT FOR VARIOUS TIMES AFTER BURIAL
(3-D CALCULATION)
130
-------
U)
t
c
CO
.148°C
70°C
58°C
52°C
6
Distance (m)
49°C
12
FIGURE B-31 ISOTHERMS IN SALT AROUND THE WASTE CANISTER
MIDPLANE (Z = 0) AT THE 2-YEAR POINT AFTER BURIAL
(0=3.95 kW/CANISTER)
-------
200
150
o
100
50
10
Spacing keyed to 150 kW/acre measured
at ten-year point. Reprocessed wastes
with canister heat output of 3.95 kW.
10
Distance (m)
FIGURE B-32 HORIZONTAL TEMPERATURE DISTRIBUTION IN GRANITE FOR
VARIOUS TIMES AFTER BURIAL (3-D CALCULATIONS)
!150 kW/ACRE)
-------
150
Note: Spacing keyed to 150 kW/acre measured
at ten-year point. Reprocessed wastes
with canister heat output of 3.95 kW.
100
£
3
+rf
ro
i
I
50
10
10
20
Distance (m)
30
40
FIGURE B-33 VERTICAL TEMPERATURE DISTRIBUTION IN GRANITE FOR
VARIOUS TIMES AFTER BURIAL (3-D CALCULATION)
{150 kW/ACRE)
133
-------
OJ
-p-
O 0
120°C
82°C
181°CV
59°C
48°C
45°C
12
Distance (m)
FIGURE B-34 ISOTHERMS IN GRANITE AROUND THE WASTE CANISTER AT CANISTER
MIDPLANE (Z=0) AT THE 2-YEAR POINT AFTER BURIAL
(Q = 3.95 kW/CANISTER)
-------
the lower conductivity, and there is very little difference in the
temperature profiles (1-D case).
The effect of a lower heat generation rate in granite (1.58 kW/
canister vs 3,95 kW/canister) Is given in Figures B-35, B-36, and B-37.
The maximum gradient in all three main directions is about a factor of
three smaller than for the higher-power case.
No general criteria have been established or suggested for a granite
repository. The repository temperature distributions can be tailored
to any criteria by varying canister heat output and/or planar heat
densities.
B-3.5.3,3.3 Results of 2-D Analyses
Planar heat densities may be limited by temperature constraints or
by thermal expansion and its resultant thermal stresses. Therefore, 2-D
calculations have been used to estimate the amount of uplift that might
occur in the absence of any constraints on thermal expansion. The 2-D
calculations are also necessary in order to examine "edge-effects" at
the extremities of the disposal area. (See Figures B-38 and B-39).
Based on calculations for a 460-m (1500-ft) repository, maximum stored-
heat conditions are reached in 500-1000 years. Maximum uplift is assumed
to occur at the time of maximum stored energy.
A summary of the key results of these uplift calculations is pre-
sented in Table B-16. For the purpose of this study, it was assumed
that all the volumetric expansion, taken as three times the linear
expansion, acts in the vertical direction. This is perhaps a reasonable
assumption for the relatively plastic salt dome, but for the bedded
salt and granite repositories, where little of the expansion is in salt,
this assumption probably overstates the amount of uplift. Another
element of conservatism is introduced for the salt repositories by virtue
of the fact that void collapse (subsidence) has been ignored.
The results of the 2-D calculations, based on the same volumetric
expansion coefficients, are shown in Figure B-40. The calculated
uplift is quite uniform across the central portion of the storage area,
but drops off rapidly at the boundary. Even with this rapid
drop-off, however, the calculated maximum ground slope near the boundary
is rather small, only 1.5°, 0.5°, and 0.2° for the salt dome, bedded
salt, and granite repositories, respectively. The calculated uplift
decreases to zero about 2000 feet outside the repository boundary.
Recognizing that these uplift values are conservative, the edge
effects do not appear to represent a major problem. If necessary, planar
heat densities can be lowered at the edges of the repository to smooth
out edge effects. Nevertheless, there are still uncertainties, especially
for granite; these relate to the effect of uplift and/or thermal stresses
on the integrity of the rock above the repository. Specifically, it
135
-------
100
o>
o__
£
1
50
10
Note:
Spacing keyed to 60 kW/acre at
ten-year point. Reprocessed waste
with heat output of 1.58 kW/canister.
10
Distance (m) X •»
FIGURE B-35 HORIZONTAL TEMPERATURE DISTRIBUTION IN GRANITE FOR
VARIOUS TIMES AFTER BURIAL (3-D CALCULATION)
(60KW/ACRE)
-------
150
u 100
o
I
0)
Q.
0)
50
10
Note: Spacing keyed to 60 kW/acre at
ten-year point. Reprocessed waste
with heat output of 1.58 kW/canister
.25
I I
10
20
30
Distance (m)
40
50
FIGURE B-36
VERTICAL TEMPERATURE DISTRIBUTION IN GRANITE
FOR VARIOUS TIMES AFTER BURIAL (3-D CALCULATION)
(60 kW/ACRE)
-------
UJ
00
T
-- - 1
*8
C
(D
«-<
tf»
5 o
- \54°C
"\ A
\ oU ty\
• %
• 1 1
41°C
33°C
29°C
o\
96°C
Distance (m)
X
28°C
12
FIGURE B-37 ISOTHERMS IN GRANITE AROUND WASTE CANISTER AT CANISTER
MIDPLANE (Z = 0) AT THE 2-YEAR POINT AFTER BURIAL
(Q = 1.58 kW/CANISTER)
-------
200
O
o
a
E
100
Note: Based on 10-year-old reprocessed
waste with planar heat density
of 150 kW/acre at burial time.
Wastes at 460m (1500 ft).
i I I I I I I
2-D Analysis
I i i I I 1 1
10
100
1,000
10,000
Time (years)
FIGURE B-38 COMPARISON OF PREDICTED MAXIMUM MINE TEMPERATURE VS. TIME
FOR A ONE- AND TWO-DIMENSIONAL ANALYSIS IN SALT DOME
139
-------
300
200 -
o
£
f
a
E
100 -
Based on 10 year old reprocessed
waste with planar heat density of
150 kW/acre at burial time.
Wastes at 460m (1,500 ft).
Radius (km)
FIGURE B-39 TEMPERATURE RISE VS. RADIUS AT CANISTER MID-PLANE (Z = 0)
FOR A1100-ACRE STORAGE FACILITY IN A SALT DOME (RESULT OF 2-D
(RESULT OF 2-D ANALYSIS)
-------
0>
Based on 10-year-old reprocessed waste with planar
heat density of 150 kW/acre at burial time.
Wastes 460 m (1500 ft). The total volumetric expansion
is assumed to act in vertical direction.
Salt Dome
(At 500 Years)
Case III-B
Storage Facility
Bedded Salt
(At 1000 Years)
Case I-B
Maximum
Slope
Granite
(At 1000 Years)
Case I I-B
Radius (km)
FIGURE B-40 MAXIMUM VERTICAL UPLIFT VS. RADIUS FOR A
1100-ACRE STORAGE FACILITY IN THREE
GEOLOGIC FORMATIONS (RESULTS OF 2-D
ANALYSIS)
141
-------
would be undesirable if thermal stresses caused connected faulting to
develop between the repository and near-surface rock; further work
is needed on rock mechanics to determine whether this is a significant
problem. This potential problem can be accommodated by limiting planar
heat densities sufficiently, and by smoothing out edge effects.
It should be emphasized that the preceding discussion addressed
thermal uplift problems for HLW burial. For spent fuel disposal, the
maximum stored energy at 500 years is about twice that of high-level
waste. This difference is best noted in Figure B-28. Thus, for the
same limits on total uplift, planar heat densities would have to be
roughly a factor of two less for spent fuel than for HLW.
B-3.6 EFFECTIVENESS OF ENGINEERING CONTROLS
B-3.6.1 Reference Technologies in Bedded Salt
B-3.6.1.1 Barriers to Transport from Repository
Since even in intact salt there is perhaps 0.1-0.5% water and since
the water tends to migrate toward a heat source, the high-level waste
canisters are assumed to be exposed to saline waters that can corrode
the canister from the time of emplacement. Subsequently, in the case of
HLW glass, slow leaching and removal of the glass itself would occur.
With simple calcines, dissolution is almost immediate for practical purposes
after the canister is destroyed. In spent fuel disposal, after the
canister, the fuel cladding, and finally, the uranium oxide of the fuel
itself will leach away. In each case, the rate of removal of actinides
from the repository may be limited by the low solubility of the actinides
in the water flow through the repository.
The immediately following sections discuss canister destruction,
leach rates, and actinlde solubility limits.
B-3.6.1.2 Waste Canister Integrity
The importance of the integrity of a canister to the transport of
high-level wastes from the repository will depend upon the contents of
the canister. For example, for a canister containing HLW glass, the
canister integrity may be of minor importance because, even with a rapid
destruction rate of the canister, the rate of radioactivity release will
be limited by the low leach rate of the glass.
Conversely, with a more leachable calcine, canister integrity could
be Important, especially over a period of several hundred years.
For the glass reference case, the canister material will probably
be determined by the requirements of the glass loading and handling
operations. As reviewed in Section B-3.2.1 of this report, the "in-can"
. 142
-------
melting process requires a stainless steel canister or other high-strength
alloy material (e.g., Inconel).
For calcine or spent fuel in salt, it is not at all clear that
stainless steel canisters would suffice. The integrity of the canister
is needed for a long period, but the high temperature and the water
present in salt would tend to subject the stainless or high alloy steel
to stress chloride cracking. This potential for rapid destruction suggests
that, for calcine or spent fuel, some other canister material, e.g,
titanium, be utilized. For titanium in the presence of salt and other
chloride-containing compounds, even at fairly high temperatures, data
indicate a corrosion rate of about 0.025 mm (1 mil) per year (see
Appendix B-V). Much better corrosion rates could be possible for
specifically-tailored titanium alloys but these may not be cost-effective.
It seeras clear that, with a relatively modest titanium outer canister
(e.g., 13-mm (0.5-in) titanium), the integrity of the canister could be
about 500 years. A somewhat thicker canister, perhaps 25 mm (1 in), could
be hypothesized to last about 1000 years, even if the "poor environment"
were to develop immediately after the high-level waste repository is sealed.
This kind of time frame, 500 to 1000 years, is critical for the
easlly-leaehable calcine; during this period the initially controlling
hazard, fission products, decays to a level that is unimportant compared
with that of the long-lived transuranic isotopes. The two fission products
of major concern are cesium-137 and strontium-90, which have half-lives
of 30.2 and 28.9 years, respectively.
B-3.6.1.3 Leachability of Matrix
B-3.6.1.3.1 General Considerations
The leachability of high-level waste glasses will be significantly
influenced by several conditions, many of which are unknown and to a
degree, unpredictable. Considering that a wide range is possible, the
estimate developed should be a high one, where there is little likelihood
of a higher removal rate.
Several of the parameters that bear on glass leachability are:
1. The type of glass and its loading of waste solids.
2. The chemical characteristics of the leaching water.
3. The temperature and pressure of the leach system.
4, The physical size characteristics (surface-to-mass ratio)
of the glass.
5. The aging characteristics of the glass in the presence of
radiation.
143
-------
The above factors were considered, the literature reviewed, and
technical discussions held, in order to understand the current status
of technology and to develop best estimates that are related to the
above effects. These points will be discussed in more detail below and
the reference values for the in-situ glass waste storage system will be
suggested.
Besides the HLW glasses, two other alternatives are considered.
The first is that of a calcine of the high-level waste (see Section
B-3.2,1.1,1), where the leach rates are so high that calcine failure can
be considered to take place over a few days* time, once the canister
integrity is breached. The other form of high-level waste to be considered
is that of spent fuel assemblies (see Section B-3.2.5), which appear
capable of behavior anywhere between that of a glass or a calcine. The
bases for the development of the reference leach rate limits will be
reviewed below.
B-3.6.1.3.2 Glass Leach Rates
In choosing an appropriate leach rate for glass, a fundamental
question is how well the various laboratory tests used to evaluate this
parameter reflect the behavior of such glass in a long-term geologic
environment. Furthermore, the procedures used for the tests themselves
can, and do, sometimes give substantially different results for a number
of reasons. One reason is related to whether the leaching process is
one in which water diffuses into the surface of the glass and removes
specific elements, or whether layers of the glass itself dissolve and
with it the elements contained in those layers, a process sometimes
described as a "corrosion effect". In the short-term laboratory
tests, the leach phenomenon appears to be a combination of these two
mechanisms. Such a combination explains a faster leach rate for a few
days of leaching and a much lower leach rate that appears to be approach-
ing an equilibrium removal condition after longer periods, e.g., years.
On the other hand, even a few years is a short period of time compared
with the geologic time involved in the dissolution of glass after
burial.
The data are further clouded by differences in the character of
the leach water, the temperature, and the physical (mechanical) flow C\T\
at the surface of the glass. Except for a long-term field test in Canada
very little data exist to evaluate an actual storage situation, and no
results have yet been published for hot salt solution.
Laboratory measurements of leach rates are generally made by
measuring a particular radioisotope in water after it has been in con-
tact with a glass for a given period of time. On the basis of the
original concentration of that particular isotope in the glass, a cor-
responding average glass leach rate (corrosion) can be calculated. If
this removal is a "corrosion" effect, or removal of a layer from the
surface of the glass, it would not matter which isotope was used for
144
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the measurement, as any isotope would represent the same fraction of
glass removed. On the other hand, if the release were a diffusion re-
moval phenomenon, one element might be leached preferentially to another
and, if calculated as above, would give a higher glass leach rate. Further
more, a short leach time could be surface diffusion-controlled while,
as explained above, a long leach period could be more corrosion-related.
Most tests, however, are usually limited to periods of much less than
one year. Several different sources of leach data have been reviewed.
Table B-20 summarizes the range of rates found, depending upon several
of these conditions.
Both the Mendel data and the Ross data in Table B-20 indicate a
leach rate of 10~~" g/cm^-day at the one-day point and about 10 g/cm^-
day at the one-year point,^' These values are consistent with the 10~
to 10 range given in an international summary paper^ , all apparently
based on room-temperature leach testing utilizing an International Atomic
Energy Agency (IAEA) type of experiment. Ross also indicates an early
leach rate of approximately 10 g/cm -day but in a different (Soxhlet)
type test apparatus. The Soxhlet type of equipment basically distills
water, condenses it, and flushes the hot water over the surface of a
sample. This kind of test is substantially different from the IAEA
procedure in several details:
1. It is at an elevated temperature, e.g., approximately 100 C.
2. The water in contact with the sample has been distilled and
condensed so there is no solid content in the water that
can inhibit leachability.
3. Because of the flushing, there is more agitation of the
liquid at the sample surface, which \ould accelerate
leaching.
The Merritt data in Table B-20 are of interest because they summarize
laboratory and field data on the same material.(37) fwo laboratory tests
were run on the basic glass used in the experiment, a nepheline syenite
glass; the second one, Glass .2 in Table B-20, is applicable to the field
test.
When this glass was put into the ground in a closely-packed array,
in an area where ground water flowed rapidly, the initial leach rate was
about 4 x 10 g/cm -day. It decreased rapidly until at about seven
years the value leveled out at 5 x 10~ g/cm -day. The factors that
specifically relate to this large reduction cannot be identified.
Furthermore, there are apparently no other field tests from which a
generalized conclusion could be evolved. These Canadian tests were at
conditions substantially different from those expected in a salt reposi-
tory.
Recent experiments at Pennsylvania State University , conducted
at extreme conditions (300°C, 300 atmospheres in the presence of water)
show that glass may be destroyed in a matter of weeks; the effect of
this information in limiting the thermal loading of repositories has not
yet been addressed.
145
-------
TABLE B-20
GLASS LEACH RATE CONSTANTS - Gm Glass/cm2-day
Source:
Mendel
Ross
Based on Cs and on Sr
(B)
-5
-6
Based on Soxhlet Test: 10
(Time not indicated)
Based on Cs 10
Merritt(C)
First Test (Glass 1) 10"
Second Test (Glass 2) 10
Field Test (Glass 2) 4x10
-5
-8
Leach Period - Days
1
ID'6
20 115 300
ID'7
700 2900
4x10
"7
2xlO
~7
1.5xlO
~7
10
10
_7
-6
10
-8
7x10
-8
4xlO
~10
(A)
High-Level Waste Glass - Mendel, E., Nuclear Technology, Vol. 32, January 1977.
Devitrification on these same glasses increased leach rate by a factor of
10 - 20.
(B)
Properties and Characteristics of High-Level Waste Glass, Ross, W.A.,
BNWL - SA-6146, Washington, D.C, Soxhlet test is batch "flowing" water
at approximately 99°C. Devitrification increased leach by up to six
times as measured by Soxhlet test.
(C)
^ High-Level Waste Glass: Field Leach Test, Merritt, W.F., Nuclear Technology,
'Vol. 32, January 1977.
146
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Thus, the leach data for glasses are conflicting, imprecise, and
scanty. The effects of temperature, flow, pH, and salinity have not even
begun to be studied in an organized way for the spectrum of candidate
glasses. Laboratory tests to date have not been .lengthy enough to
achieve a measure of the long-term (steady-state) corrosion rate. Theo-
retical studies(40) indicate that these rates should continue to' decrease
with time for considerable periods, a conclusion that is not inconsistent
with the results of the one long-term field test reported in Table B-20.
It appears unreasonable to do more than assume a conservatively
"high value for leach rate constant, allowing for the uncertainties in
the above parameters. (The assumption has been made that temperature
affects the leach rate as it does any diffusion-controlled phenomenon —
i.e., that the rate varies with the square root of the absolute temperature,
Under this assumption, even at 200°C the rate would be only 26% faster than
at 25°C. The relatively minor effect of temperature seems supported by
the Soxhlet data in Table B-20.) A value of 10~6 g/cm -day has been
chosen because it is at least a factor of 7 to 100 times larger than any
data reported after one year. By not assuming it to decrease with time
as do the laboratory values, a further conservatism is introduced into
this value, which should more than compensate for any uncertainty regarding
devitrification (see below). At the same time, it must be admitted that
the data indicate that it is entirely possible that leach rate constants
as low at 1CT° g/cm^-day may be achieved in practice. Accordingly
10~" g/cm -day has been adopted as a reasonably conservative long-term
value, and 10 g/cm2-day as a possibly achievable long-term value.
B-3.6.1.3.2.1 Surface-to-Mass Ratio
The key determinant in establishing a particle size for the high-
level waste glass deals with the glass "casting" operation and its
subsequent cool-down rate. After the glass is poured into the waste dis-
posal canister, or alternatively, after the glass melting operation takes
place in the disposal canister, cool-down can be rapid until the glass
reaches a temperature of approximately 500-600°C, at which point the
glass is still plastic enough that cracking will not occur. Below this
temperature the cool-down rate, in combination with the shrinkage stresses,
will determine the amount of cracking that ultimately takes place in the
glass.
This effect has been quantitatively evaluated by BNWL, which indicates
that the relative surface areas of the as-cast cylinders, in this case
15 cm (6 in) diameter, increased by a factor of 20-25 for rapid water
quenching, a factor of 12 for air-cooling, and a factor of about 5 for
slow-cooling at a rate of about 50°C/hour.(39) There was essentially
no increase for a full-annealing cooling rate, e.g., less than l°C/hour.
147
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It has not been established what cooling cycle will be, used f.BNWE
indicates that cooling probably would be in the range from slow cooling
to free air convection, which might result in an increase of about seven
times in surface area. Whether this relative surface area increase is
the same for the larger-diameter cylinder, e.g., 30 cm (12 in) is not
known. Because of the inherently slower cool—down rate, the larger-
diameter cylinder would probably have a somewhat lower relative surface
area increase for a given mode of coaling. For a slow-cool or annealed
case, a soaking pit operation would be required and such a "storage"
operation would probably add to the problems of facility throughput.
In the as-cast condition, a 0.3 m (12 In) diameter by 3.0 m (10 ft)
high cylinder, has a surface-to-mass ratio of about 0.05 cm^/g. For
reasonably slow cooling, a multiplier of seven has been assumed, so the
surf ace-to-mass ratio would be 0,35 enr' /g.
Another factor that must be considered in setting the surface-to-
mass ratio is that of possible mechanical damage to the glass of the
cylinders. Very high impact velocities, e.g., 21 m/sec (70 ft/sec) can
increase glass surface by a factor of perhaps ten or more. On the other
hand, fairly nominal impacts, e.g.,1.5 m/sec (5 ft/sec), would not Increase
the overall surface area by more than about 10%. There is no reason to
assume that any canisters subjected to a high-impact force would be sent
to disposal; hence, the nominal kind of incident is the type that should
be considered. A 10% increase of the surface-to-mass ratio, as developed
above from the cooling analysis, would increase the surface to about
0.4 cm^/g. This surface-to-mass ratio is equivalent to that of a cube
of glass 4,7 cm on an edge containing approximately 332 g of glass. This
initial "particle size" has been used in the bulk leaching analysis
presented below in Section B-3.6.1.3.5.
B-3,6.1.3.2.2 Devitrification
Although there has been a great deal of discussion of possible de-
vitrification of glass, there is no basis at the present time to assume
that such devitrification will occur. Tests have been run on various
high-level waste glasses, at different temperatures, "spiked" with curium
in order to develop radiation damage effects more rapidly. Samples with
curium were prepared in 1974 and these samples have developed an equivalent
age of up to 2000 years from the standpoint of radiation damage.^ ' Many
characteristics have been measured, such as stored energy, density change,
leach rate, physical strength, etc. Relatively modest changes in leach
rates have been noted, not more than a factor of two higher, but the
radiation exposure does not seem to have any bearing on devitrification.
Similar results were obtained by the Karlsruhe Laboratories in Germany,
The studies indicate that devitrification processes are basically effects
caused by high temperatures, generally in the range of 700-900°C, but
some devitrification may occur in the 400-500°C range.
148
-------
Devitrification, if it were to occur, might increase the leach rate
by a factor of approximately ten. Specifically, effects of various de-
vitrification temperatures observed by Ross'3:0 indicate that the bulk
leach rate increased by a factor of six with the peak occurring at a
devitrification temperature of 700°C.
B-3.6.1.3.3 Calcine Leach Rates
Calcine leach rates are generally high, with values reported of
abput 0.1 g/cm -day.(11) This rate would result in total dissolution of
a calcine, even assuming that it had a surface-to-mass ratio as low as
that of the bulk glass described above, in less than 60 days. For all
practical purposes, then, calcine dissolution per se does not represent
a delay function in the waste disposal release scenario.
B-3.6.1.3.4 Spent Fuel Dissolution Rates
The major development work on the leaching of spent fuel was per-
formed by BNWL.^2) This study utilized particles of high-burnup light
water reactor fuel pellets and attempted the dissolution of these fuel
particles with Hanford ground water, distilled water, and deionized
water. Leaching of chemical elements (curium, plutonium, strontium/ and
cesium) was evaluated. Within the leach time of those studies, 150 days,
the leach rate constants that were established ranged from about 4 x
10~" g/em -day to about 7 x 10 g/ctrr-day. The lowest values were for
dissolution derived from curium removal measurements while the high rates
were based on cesium.
The data reported farther indicate that after one day, the "leach"
values are on the order of 100 times higher than after 150 days. These
data indicate that fuel assembly leaching might be less than that
of high-level waste glasses, certainly not the anticipated result.
In order to estimate the long-term reachability of spent fuel, it
is necessary to consider whether conditions may be present that will
cause the fuel to oxidize, i.e., to change from its initial ceramic
form to a simple, finely-divided, oxide mixture. If oxygen (air)
is available to spent fuel oxide pellets at temperatures much above 120-
150°C (250-300°F), the fuel rapidly oxidizes from U02 to l^Og, and in
so doing, forms a finely—divided powder that expands to the point where
fuel cladding is ruptured. The effect would propagate from the first
small pinhole to develop in the Zircaloy cladding. If this phenomenon
occurred, leaching of the fuel would be at least as rapid as with calcine.
How much effect water of varying chemistries would have on a dissolution
rate of fuel is not known, and therefore a reasonable single reference
value for spent fuel leaching cannot be set with any confidence. The
leaching values that can be hypothesized, based on non-oxidizing
conditions, could be similar to those of the high-level waste glasses.
149
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At the other extreme, if hot, oxidizing conditions are assumed, leach
rates could well resemble those of calcines, i.e., rapid.
B-3.6.1.3.5 Total Leach Times
B-3.6.1.3.5.1 Glass Leach Times
In order to establish the quantity of material that might be leached
under idealized situations for subsequent transport to the environment,
the conditions previously reviewed have been Integrated into an overall
removal function.
3
For a cube initially of side x cm, with density p g/cm , with a
specific leach rate constant of k g/cm^-day, it is easy to show that
after a time of 0 days, the side of the cube has shrunk to x cm, where
-------
.E
._
o>
QC
c
I
-------
.014
.012
03
CD
l_
OJ
Q.
"D
0>
u
ra
d>
.010
.008
"5 .006
O
4—
O
•(->
OJ
u
m
Q_
.004
.002
0
4,000
8,000
12,000
16,000
20,000
24,000
Time from Start of Leach, Years
FIGURE B-42 ANNUAL LEACH RATE VS. TIM1, k « 10"6 g/cm2-day
(CONSERVATIVE VALUEI
152
-------
O)
DC
CD
4-
O
*~i
I
o
Q.
100
80
60
40
20
0 L
0
0,8 1.2 1.6
Time from Start of Leach, Millions of Years
2.0
2.4
FIGURE B-43 PERCENT OF GLASS REMAINING (UNLEACHED) VS. TIME
k = 10"8 g/cm2-day (POSSIBLY ATTAINABLE VALUE)
153
-------
1
o
(O
(D
J2
O
"<5
o
"i™
o
<*-
o
4->
.00014
.00012
.00010
.00008
.00006
.00004
.00002
0
\
t\
2.0
0.4 0.8 1.2 1.6
Time from Start of Leach, Millions of Years
FIGURE B-44 ANNUAL LEACH RATE VS. TIME, k = W8 a/cm2-day
(POSSIBLY ATTAINABLE VALUEJ
2.4
154
-------
calcine containment is dependent primarily upon the canister in which it
is buried. In a salt repository, there is not much difference between
the corrosion resistance of carbon or stainless steel. Carbon steel has
perhaps a more uniform corrosion rate, but stainless steel is more sub-
ject to stress chloride cracking. Although the mechanisms are different,
both might be breached in a relatively short time. The failure point is
probably about 50 years; if the geologic phenomenon that could start
the leaching process occurs, the material is effectively available for
leaching and transport in a short period of time.
It would be desirable, if at all feasible, for the canister to pro-
vide some longer-term retention for a material like a calcine, which has
little integrity of its own. About 500-1000 years would be a suitable
period, since in this time frame the fission products are controlling
and are generally more soluble than the TRUs.
Titanium might be useful as a container for calcine in a salt
repository. The relatively standard titanium alloys have a corrosion
rate of about 0.025 mm (1 mil) per year, for a high temperature salt
wet/dry situation (see Appendix B-V). Some alloys of titanium, includ-
ing one with palladium, would have much better corrosion properties,
but would be substantially more expensive. It appears that 0.025 mm (1 mil)
per year would be acceptable; as a corrosion rate; therefore,using
"commercial" material and thickening the wall appropriately to last
500-1000 years would be a better approach than using a more expensive
canister. This approach should be explored to establish the optimum
cost/benefit. For the purpose of estimating hazards, a canister lifetime
of about 800 years could be assumed, but after canister failure, fairly
rapid release of the waste calcines must be assumed.
A "supercalcine" might be developed for most waste compositions
that would have very good leach resistance characteristics, but it
appears that the supercalcines must be specifically tailored to a given
waste composition requiring a degree of control not necessary for glasses.
Furthermore, because the procedures for tailoring and sintering are not
much different from those of glass production, there appears to be no
significant advantage to this approach.
B-3.6.1.3.5.3 Spent Fuel Assembly Leach Times
The containment in series is first, the canister; second, the zir-
conium fuel cladding; and third, the ceramic UOo. As indicated in the
prior discussion, the leach rate applicable to spent fuel assemblies is
highly speculative. Data in BNWL-2057^2) suggest that spent fuel
assemblies can be assumed to be similar in leach characteristics to the
glass previously discussed. Depending on the oxygen content of the
ground water, however, the U02 in a heated fuel assembly may be oxidized
to the UoOg state. If this were the case, it would have leaching
characteristics similar to that of the HLW calcine. In this case, the
155
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short-term integrity could be determined, similar to that of the calcine,
by the canister lifetime characteristics. Depending on the temperature
and the oxidizing nature of the ground water, a spent fuel assembly may
be assumed to have leach characteristics as low as those of glass or as
high as those of calcine.
B-3.6.1.4 Solubility Considerations
The leaching rate data given above will result, for any given ratio
of leachant flow to high-level waste weight, in a particular set of con-
centrations of all soluble species. For low flows, however, the very
low solubilities of the actinide compounds may limit their concentration
in the leachate.
Rai has correlated the solubilities of the actinide compounds.
His approach was to determine the most insoluble of all the feasible
compounds for each actinide, then to calculate from solubility products
the concentration of all soluble actinide-bearing ions in equilibrium
with that compound. These solubilities are a function of pH, since H
ion is usually involved in the equilibrium. Thus, in the dissolution of
PuQ2 to form Pu024+ ion,
PuO- + 2H+ + 1/2 02 = PuOj4"*' + HO (3)
the equilibrium PuC>2 concentration (solubility) ought to Increase with
the square of the IT*" ion concentration, or the log (Pu02"*~*") ought to de-
crease in accordance with twice the pH.
Solubility relations such as this are summarized in Table B-21 for
the four principal actinides, based on curves in BNWL-1983.v^O) Relation-
ships are given in saturated salt solutions and in non-salt solutions like
those found in rock.
When leach calculations are made, if the actinide concentrations in
the leachate should exceed the values calculated from Table B-21 , they
should be reduced to those values.
B-3.6.2 Reference Technologies in Other Media
In the absence of published information, it can only be assumed that
the same leach rates presented above for the salt environment would apply
roughly in a rock repository as well. The actinide solubilities would
be different, however, and the differences have been shown in Table B-21.
156
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TABLE B-21
SOLUBILITY LIMITS OF PRINCIPAL SPECIES FROM ACTINIDE DISSOLUTION
Actinide
In Sat'd NaCl Solution
In Water in Rock
Least Soluble
Solid
Most Soluble Species
in Solution
Least Soluble
Solid
Most Soluble Species in
Solution
Np
Pu
Am
Np02(OH)2
PuO,,
Am02OH
log [U02C1 ]= -0.7-(4)(pH)
log [Np02 ] = 4-pH
log [Pu02(OH)°2] = -11.5
log [Pu02Cl°] = -4.8-pH
log [AmOH} = -4.4 -(2)(pH)
log [AmCl ] = 3.3 -(3)(pH)
U02(OH)2
Np02(OH)2
PuOr
AmO OH
log [U02 ]=5 -(2)(pH)
W/C02 @ .0003 atm. in
alkaline medium,
log [U02++] < 11.2 -(2)(PH)
log [Np02 ] = 4 - pH
W/CO., same
log [Pu02(OH)°2] = -11.5
log
= -0.5-(2)(pH)
log [Pu02] = -5.3 - pH
w/CO-, above plus
log [Pu02(C03)(OH)2=K -27.2+(2)(pH)
log [AmOH ] = -4.4 - (2)(pH)
Effect of CO- unknown
Note: All logarithms are to the base 10. Solubilities in g mols/liter
Source: Based on data in Ames, L.L., et al, "A Review of Actinide-Sediment
Reactions...," BNWL-1983, Battelle Northwest Laboratory, February 10, 1976.
-------
Page Intentionally Blank
-------
B-4.0 COST CONSIDERATIONS
B-4,1 GENERAL
The purpose of this section of the report is to present data on the
comparative costs of the various reference cases discussed in Section 3.0.
Such information will be needed in order to evaluate the cost effective-
ness of alternative disposal techniques. Since only limited amounts of
cost data have been published, and because it was beyond the scope of
this study to do independent cost-estimating, the data presented should
be considered only as rough estimates. Nevertheless, they provide useful
perspective on the relative economics of various disposal techniques.
Two basic types of HLW are considered in this section—solidified
HLLW from reprocessing operations and spent fuel elements from a throw-
away cycle. Both types are assumed to be suitably packaged for non-
retrievable disposal in deep geologic formations (repositories). The
effect (on the cost) of incorporating retrievability into the design
of such a repository will also be shown.
The two waste paths considered are:
For Disposal as Solid HLW For Disposal as Spent Fuel
* Storage of spent fuel at
reactor sites and at interim
storage sites for a total
of 10 years.
* Transfer of spent fuel
to reprocessor.
* Reprocessing
* Prompt conversion of HLLW
to canned solid HLW.
* Transfer of HLW canister to
repository.
NA
* Burial of HLW canisters.
Storage of spent fuel at
reactor sites and interim
storage sites for a. total
of 10 years.
*
NA
NA
NA
Transport of spent fuel to
repository.
Canning of spent fuel.
Burial of spent fuel canisters.
Thus, the sequence for the solidified HLW disposal scheme is what
has generally been assumed in the past (but with delayed reprocessing),
with prompt conversion of the reprocessing wastes to canned solids, for
disposal at a Federal repository. The sequence for the spent-fuel dis-
posal scheme (the "throwaway cycle") assumes the identical combined ten-
year storage of spent fuel at the reactor site and at a centralized
Not Applicable
159
-------
interim storage facility, and finally burial at the repository. The
spent-fuel transport industry has evolved on the principle of shipping
uncanned fuel in special casks; the same approach would probably be
used to transport spent fuel from interim storage facilities to the
repository, where final canning would take place. Specialized equipment
and machinery are required with stringent quality control; therefore,
it seems more reliable and more cost-effective to perform canning at a
single site (the repository) rather than at multiple interim storage
sites or reactor sites.
For each approach, approximate cost figures will be given, with
the exception of the following items, which are beyond the scope of
this assignment:
• Reprocessing costs will not be given.
* Transportation costs will be estimated as a general
range only.
B-4.2 SPENT-FUEL STORAGE COSTS
Storage of spent fuel at reactor sites (using high-density storage
racks in lieu of the normal-density racks hitherto used) and at interim
facilities for a total of ten years can be performed for costs of about
$80-150/kg of heavy metal (HM)*, depending on the financing costs and
including an allowance for intermediate shipment at $10-20/kg HM. These
costs are the same, regardless of which of the two waste paths is chosen.
B-4.3 POST STORAGE/PRE-BURIAL COSTS
B-4.3.1 SolidHLW Disposal Path
B-4.3.1.1 Transportation Costs (to Reprocessor/Solldification Plant)
Transportation of ten-year-old spent fuel from interim pool storage
to the reprocessing site would cost about $8-18/kg HM. No cost is
assigned in this study to the reprocessing itself. Because the solidi-
fication facility is assumed to be at the reprocessing site, there is no
transport charge for HLLW movement.
B-4.3.1.2 HLLW Solidification Costs
The solidification process will probably take place at the repro-
cessing facility. Although it is possible that the reprocessing
facilities will also be located at, or adjacent to, the HLW repository,
location is a factor reflected only in charges for shipping solidified
waste, which will be about the same for all alternatives (see below).
Heavy metal means the uranium and plutonium in the fuel originally
charged to the reactor.
160
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The HLW,solidification facilities are assumed to handle the normal
throughput of a reprocessing plant, which is in the range of 4-8 MTHM/day.
A 5 MTHM/day (or 1500 MTHM/yr) size is chosen to be the same as that of
the AGNS facility and that proposed by Exxon,
In considering the relative costs of the competing solidification
processes, it should first be noted that capital costs dominate and
equipment cost is a relatively small fraction of the total investment.
The major portion of the capital investment is for the "hot cell" itself
plus associated auxiliary support systems, e.g., remote handling and
maintenance equipment, the off-gas treatment system, etc.
One of the key determinants of hot-cell costs is physical size,
especially area. Area depends upon the space needed for process equip-
ment and is particularly sensitive to the amount of redundancy required
for such equipment.' Thus, if one of the processes were judged to require
inherently more redundancy or a substantially greater cell area for a
given throughput, that process would clearly have an economic disad-
vantage relative to other processes with more compact layouts. In gauging
the amount of redundancy required, particular attention must be paid to
reliability and maintenance problems in a highly radioactive environment.
To date, only one detailed study of solidification costs has been
published, C^-*/ Cost data on total reprocessing plants have been
reported, however, and an attempt can be made to allocate a portion of,
total costs to the HLLW solidification functions.
Data have been published for two reprocessing plants involving HLLW
solidification—the proposed Exxon Nuclear Fuel Recycle Facility and the
AGNS plant. In addition, the S. M. Stoller Corporation (SMSC) and
Battelle Northwest Laboratories (BNWL) have independently developed such
cost data.
B-4.3.1.2.1 Exxon Nuclear Fuel Recycle Facility
The Exxon Fuel Recycle Facility PSAR indicates that the con-
centrated HLLW will be sent to a fluidized-bed waste calciner utilizing
aluminum oxide as a starting granular material and with a fluidizing
gas of superheated steam. The calcined wastes then flow to a mixing/
metering vessel, where they may be combined with an added material
suitable for glass-forming or fed directly to a waste-melting furnace.
The melt then flows into waste canisters that are air-cooled and are
then temporarily stored in a water pool beneath the process operating cell,
In the published data Exxon has not allocated costs to this cal-
cining/glassif ication operation, but has estimated the total Recycle
Reprocessing Facility capital cost at approximately $600 million' -^ in
1975 dollars. This includes, of course, all of the common auxiliary
service functions required for the operation of the entire facility —
161
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spent fuel assembly storage, -cooling towers, fire water system, roads,
security, chemical tank farms, sanitary wastes, emergency power,
ventilation system, administration facilities, etc.
In order to estimate roughly the fraction of capital costs that
should be allocated to the HLW solidification portion of the plant,
the various process cell areas in the facility have been analyzed; the
HLW solidification cell constitutes approximately 5% of the total remote
and contact cell area of the Exxon plant. If 60% of total facility
cost is directly associated with the reprocessing building itself, the
HLW solidification fraction would be 3% of the total project cost. If
a 10% per year escalation is assumed since 1975 to correct for inflation,
the 1977 direct capital figure for the waste solidification cell would
be about $23 million. An additional allowance must be included for
interest during construction, which if calculated at 8% per year would
add somewhat more than $6 million, for a total rounded capital cost of
$30 million (1977 dollars) for waste solidification. The equivalent
unit cost, at an annual fixed charge rate of 20%, is approximately
$4/kg HM. Process changes may result in somewhat more or less space
required in the reprocessing plant, but the resultant cost impact
should be relatively small. Changes attributable to equipment cost
variations are expected to be even smaller and relatively unimportant
as compared with the cell cost elements.
B-4.3.1.2.2 Allied General Nuclear Services
The AGNS plant at Barnwell, S.C. does not currently have a waste-
solidification step, because regulatory criteria were not defined when
the plant was constructed. The HLLW were to be concentrated and then
stored in stainless-steel tanks pending resolution of regulatory
criteria on solidification processes and facility design. AGNS has
announcedt^"/ total costs for upgrading the facility to include waste
solidification as well as other steps. The published data are insuf-
ficient to partition the costs among the various additional process
steps, however.
B-4.3.1.2.3 Western Reprocessing Study
In a preliminary cost analysis for the Western Reprocessing Study
Group, SMSC developed a value of approximately $120-130 million of total
waste treatment facilities associated with a reprocessing plant costing
approximately $600-650 million in 1976 dollars. The waste treatment
facilities included not only HLLW solidification, but also HLLW concen-
tration and storage, as well as the cost of all other waste treatment
facilities in the plant, e.g., facilities for treatment of intermediate-
and low—level TRU wastes, etc. It is estimated that on the order of one-
quarter to one-third of total waste-treatment costs would be allocated
to the HLW solidification step, or $30-45 million, a value roughly com-
parable to that inferred above from published data on the Exxon facility.
162
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Another item of information useful for rough comparisons is the
estimated cost of the HLW calcining facility at Idaho Falls for defense
wastes, approximately $65 million.(47) Although this facility is not
the same as one for solidification of commercial HLW, many of the
features are similar, e.g., off-gas treatment systems, support systems, etc.
B-4.3.1.2.4 Battelle Study
(43)
In BNWL-1667, a cost optimization was performed to determine -
the sensitivity of waste treatment costs to variables of waste type§
cooling time, and solidification process. The calculational code was
based on a 1370 MTU/yr reprocessing throughput, 20-year plant life, a
30/70 debt/equity ratio, 15% return on equity, 8% interest on debt,
and included adjustments for equipment depreciation and tax structure.
The code calculates cost sensitivity to varying storage times, given a
waste type and solidification process.
The BNWL-1667 computational code gives results in mills/MWh, which
may not be directly comparable to other results expressed in $/kg, or
total capital costs, because the relations between mills/MWh, $/kg, and
capital costs are a function of the economic assumptions used for
different studies; nevertheless, certain observations may be made
regarding the results.
The BNWL-1667 data confirm that the waste solidification process
selected (assuming equivalent reliability) has very little effect on
total costs, although longer cooling times will decrease solidification
costs of any one system. Total costs of storage-plus-solldification
(Including canning) for five-year-old waste vary from 28 to 36 mills/MWh
(1972 dollars) over a range of waste solidification techniques from pot
calcination to spray melter solidification. Note that 28 to 36 mills/MWh
corresponds to about $7.4-9.5/kg, or, escalated to 1977 at 10% per year,
about $12-15/kg.
Based on the foregoing discussion, the cost of waste solidification
facilities should fall in the range of $30-60 million, or, $4-15/kg HM.
Cost variations due to alternative calcining or glassification processes
are likely to be small ($l-2/kg HM) and well within the uncertainty
band of these rough estimates of total waste solidification costs.
B-4.3.1.3 Canister Costs
Waste canister costs vary with the material and thickness chosen,
with carbon steel at $l/kg HM, stainless steel at $2/kg HM, and
titanium at $4-6/kg HM.
163
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B-4,3.1.4 Transportation Costs (to Repository)
Shipping coats are not expected to differ greatly between calcine
or glass, and both ought to be about one-quarter to one-third of the
cost for shipping the equivalent amount of recently-discharged spent
fuel (allowing for the higher density of waste permissible per unit
volume, since it is ten years old). This leads to a solidified-waste
transportation cost of about $3-8/kg of original HM (not per kg of
solidified waste). The low end of the range is, of course, more likely
if the reprocessing unit is located at or near the repository.
B-4.3.1.5 HLW Solidification Cost Conclusions
It is very difficult to compare different cost estimate sources
directly without a detailed comparison of cash flow assumptions. Thus,
the value of the cost data presented in this report lies not so much in
the absolute cost figures given, but rather in the following general
conclusions:
* Costs are largely insensitive to waste solidification
methods, assuming equivalent reliability.
• Cooler wastes are less expensive to treat, ship, and
dispose of, but these cost savings are offset by added
storage costs.
B-4.3.2 Spent-gue1 D i sp o s a 1 Path
B-4.3.2.1 Transportation Costs (to Repository)
Transportation of ten—year-old spent fuel from interim pool storage
to the repository is assumed to be $8-18/kg HM, the same as In Section
4.3.1.1.
B-4.3.2.2 Canning of Spent Fuel
It Is assumed that the fuel will be placed in gas-filled, sealed
outer canisters in a canning facility at or adjacent to the repository.
The major components of the spent-fuel canning cost are;
1. Fabricated canister component costs.
2. Canning facility fixed charges — the capital charges
associated with the loading, sealing, and testing
facilities necessary for the canning process.
3. Canning facility operating and maintenance charges.
164
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The fabricated canister components are estimated from $2/kg HM for
carbon steel, to $6/kg HM for stainless steel, to $15/kg HM for titanium.
Capital costs for a canning facility have not been published to date.
Based on the number of operating locations in a hot cell, such a facility
might be built to handle 3000 MTHM/yr at a cost of roughly $100 million.
Allowing for possible variations in this figure and in facility lifetime,
the equivalent unit cost is in the range of $5-10/kg HM.
Operation and maintenance charges for such a facility keyed to a
staff of about 150, are estimated to add another $l-2/kg of HM.
When the above cost ranges are combined, the resulting range of
total costs is $8-27/kg HM; most of this range of variation is due to
the $13/kg HM variation in the estimated costs of fabricated canisters.
B-4.4 REPOSITORY COSTS
B-4.4.1 Costs for Geologic Disposalin Salt
Order-of-magnitude cost estimates for geologic disposal in bedded
salt have been developed for the reference facilities discussed in
Section 3.5 of this report. The estimates (in 1977 dollars) for both
retrievable and non-retrievable facilities, designed to receive either
canistered HLW from a reprocessing cycle, or canistered spent fuel from
a throwaway cycle are summarized in Tables B-22 and B-23. The
estimated costs for mine excavation, waste handling, equipment design
and fabrication, and overall construction costs have no allowance for
contingencies.
Mining costs are based on a review of recent bid data for salt dome
production in Louisiana, as well as on recent cost experience in under-
ground mine construction. The mine general arrangement, as discussed
in Section 3.5, is similar to that evolved for OWI.Cl?) Each of the four
shafts is concrete-lined and sunk to a depth of 460m (1500 feet) through
hard rock (granite). The mine excavation costs cover all costs for
mining-related activities over the life of the facility, including sur-
face support facilities, mining machines, underground and aboveground
materials handling and hauling equipment, and mining manpower. Thus,
whether all mining is done prior to facility operation or concurrent
with facility operation, the overall mining cost component will not
change. (It should be noted that normal commercial mining accounting
practices would probably consider much of the mine excavation as an
operating expense rather than a capital item.) A unit excavation cost
of $65/nP ($50/yd3) for salt has been assumed.
Unit costs for canister holes have also been developed, based on
conventional mining data and the conceptual canister hole designs dis-
cussed in Appendix B-IV. For non-retrievable storage, costs for
HLW are $1000 to bore the hole, insert the waste, and backfill the hole.
165
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TABLE B-22 .
APPROXIMATE COSTS REFERENCE REPOSITORY IN BEDPED SALT -
HON-RETRIEVABLE STORAGE FOR EQUIVALENT OF 107.000 MTHM
Millions of (1977) Dollars
Cost Component
Surface facility
In-mine structures
Excavation
Low-level TRU
Intermediate-level TRU/
Cladding Hulls
HLW/Spent fuel
Excavation Sub-Total
Shaft costs
Canister holes
HLW
Intermediate-level TRU
Facility backfill
Facility decommissioning
Total Capital
HLW
125
60
85
314
503
902
18
36
314
30
10
1495
"Throwaway"
Case
125
60
50
127
1572
1749
18
494
163
58
10
2677
Average Annual Operating and
Maintenance Costs (over
10-year operating period)
Total (Undiscounted)
Cost for Capital + 10 years
of operation
30/yr
1795
30/yr
2977
Total (Undiscounted)
$17/kgHM
$28/kgHM
166
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TABLE B-23
ACTROXIMATE COSTS REFERENCE REPOSITORY IH BEDDEDSALT-
FULLY RETRIEVABLE STORAGE FOR EQUIVALENT OF 107,000 MTHM
Cost.Component
Surface facility
In-mine structure
Excavation
Low-level' TR0
Intermediate-level TRU/
Cladding hulls
HLW/Spent fuel
Excavation Sub-Total
Shaft Costs
Canister holes
HLW
Intermediate-level TRQ
Facility backfill
Facility decommissioning
Total Capital
Average Annual Operating and
Maintenance Costs (over
10-year operating period)
Total (Undiscounted)
Cost for Capital + 10 years
of operation
Total (Undiscounted)
Millions of (1977) Dollars
"Throwaway"
HLW Case
125
60
85
314
503
902
18
186
282
451
10
2034
30/yr
2334
125
60
50
127
1572
1749
18
1728
117
875
10
4682
30/yr
4982
$22/kg
$46/kg HM
167
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For retrievable storage, costs for HLW are about $5000 to bore the
hole, provide a one-inch thick carbon steel canister liner and
concrete shield plug, install the liner, and place the waste canister
and hole plug.
The facility design is still in the conceptual stage. Furthermore,
the NRG has not yet issued any design criteria directly applicable to
waste isolation facility design. It was therefore necessary to draw
heavily on experience in other nuclear facilities when developing cost
estimates for the reference waste burial facility structures and equip-
ment. The basic criterion in sizing the surface receiving facility and
the in-mine canister-handling stations is that the facility can physically
handle the required throughput of 350,000 canisters over the assumed
operating lifetime of ten years.
Necessary plant structures are discussed in Appendix B-IV, Briefly,
however, in-mine structures Include the HLW and intermediate-level TEU
receiving/transfer cells and the man and materials receiving station.
Surface facilities Include a canistered waste receiving building, can-
is tared waste transfer cells, an interim canistered waste storage
building, a canistered waste hoist cell and hoisting equipment, a low-
level TRU receiving building, an interim TRU storage building, a TEU
hoist facility, a man and materials hoist facility, a mine ventilation
system building, a site radwaste process facility and plant stack, and
site support facilities, such as an administration building, machine
shop, warehouse, security gate houses, and a general personnel staging ,
facility containing health physics and chemistry laboratories, a lunch
room, showers, lockers, etc. Nuclear facility and equipment cost data
from WASH-1230(48) escalajed to 1977 dollars, were used where applicable.
It was arbitrarily assumed for costing purposes that the plant radwaste
and ventilation systems, the interim canistered waste storage area, the
canistered waste handling cells, and the canistered hoist cell and
hoisting equipment will all be housed in the equivalent of seismic
Category 1 structures. Construction cost data were also taken from
R. S. Means'^' where applicable. Overall structure and equipment
requirements are similar for each of the respective reference facilities
in Tables B-24 and B-25. Annual operating and maintenance costs include
staff salaries and benefits (for all staff, except those associated
with the mining operation), facility maintenance, and costs for elec-
tricity, water, fuel, etc. Facility staff requirements are summarized
in Table B-24. Because canister handling is assumed to be a continuous
24-hour, 7-day operation, a rotating 4-shift operation is necessary for
personnel associated with canister handling (equipment operators,
mechanics, health physics support, and a skeleton maintenance crew).
Security will also operate on a rotating 4-shift basis.
For purposes of comparison of repository costs with other waste
disposal costs, a unit cost value ($/kg HM charged) has been included
for each reference facility in Tables B-22 and B-23. This number is
derived by summing the total facility capital costs and ten years
168
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TABLE-24
PLANT STAFF ESTIMATE FOR REFERENCE REPOSITORY
SURFACE FACILITY
General Staff
Administrative and Clerical
Security
Chemistry
Health Physics
Maintenance
Custodial
Canistered Waste Handling
Yard Crew
Cask Handling
Hot Cell Operation
Low-level TRU Handling
Transfer Hoist Operators
Radwaste/Laundry
Total Staff
10
21
5
21
19
Sub-Total 80
20
136
36
Sub-Total 192
3
13
4
Surface Facility Total 292
BELOW GRADE
Low-level TRU Handling
Canistered Waste Receiving
Facilities
Canistered Waste Transport
and Burial
Underground Maintenance
24
240
12
BELOW GRADE LEVEL 281
292
FACILITY TOTALS - Surface Facility
Below Grade 281
TOTAL 573
169
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TABLE B-25
SUMMARY OF HIGH-LEVEL WASTE DISPOSAL COSTS
(In Salt, Non-Retrievable)
Early storage of spent fuel (10 yr)
Transport to reprocessor
(Reprocessing)
Solidification of HLLW, inc. canning
Transport to repository
Canning
Burial of canisters
Total
$/kg BM (1977 Dollars)
"Throwaway"
HLW Case Case
80-150 80-150
8- 18
(Not considered)
6- 23
3- 8 8- 18
(Incl. in Solidif'n) 8- 27
17* 28*
114-216 124-223
Retrievability would add 30-60% to these values.
170
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of operating costs (assumed active operating design life) and dividing
by the total facility capacity (107,000 MTHM'charged). While this is
a simplified cost treatment, it is adequate for present purposes.
B-4.4.2 Cost Versus Depth
Cost versus depth was not examined in detail for this study, but
it may be assumed that deeper burial will be more costly. Shaft costs
will be higher; mining costs will be higher, because of the longer time
to transfer men and materials into and out of the mine; mine operating
costs will be higher for the same reasons. Stability criteria
for mines at greater depth may also have an adverse impact on the
general arrangement of the mine and on maximum permissible planar heat
density.
B-4.5 COST SUMMARY FOR HLW OR SPENT FUEL DISPOSAL IN SALT
The various elements contributing to total waste disposal costs
are summarized in Table B-25. The cost is dominated by the interim
storage costs. For comparison, total fuel cycle costs might be
roughly $1000/kg HM for a prompt reprocessing cycle. Hence, interim
storage alone can increase these costs by 8-15% depending upon the
type of storage facility.
B-4.6 COSTS FOR GEOLOGIC DISPOSAL IN HARD ROCK
Detailed cost estimates could not be developed for geologic dis-
posal in hard rock since suitable planar heat loading design criteria
have yet to be established for hard rock. Unit costs for excavation
have been examined and, based on preliminary findings, are expected to
be 30-100% higher for hard rock than for salt at the same elevation.
Preliminary indications are that hard rock heat input might be limited
to something on the order of 50% that of salt.(30,50) jn that case,
overall excavation costs in hard rock would be on the order of 2.6-4
times as high as for salt at the 460-meter mine depth assumed for this
study, and that overall facility costs for hard rock would be two to
three times as high as those for salt for the same waste stored.
B-4.7 NOTE ON THE DISPOSAL COSTS FOR LOW-LEVEL TRU WASTES'
Table B-26 shows estimates of costs expected for a geologic disposal
facility handling 140,000 m^ of low-level transuranic (TRU) wastes.
The surface and handling facilities will be much less expensive than for
HLW because shielding is not required. Excavation costs per unit of
waste volume will be much less than for HLW because thermal loading
restrictions will not be limiting. With lower radiation levels, annual
operating costs are lower.
171
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TABLE R-26
REFERENCE FACILITY COSTS (BEDDSD SALT) -
NON-RETRIEVABLE STORAGE OF LOW-LEVEL TRU WASTE
Cost Component
Surface facility
Below-grade facility
Mine excavation
Shaft costs
Backfill
Millions of (1977) Dollars
Non-Retrievable Retrievable
19
7
40
6
20
92
19
7
85
6
40
157
Operating coats/yr
Decommissioning
3/yr
5
3/yr
10
172
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Adding the figures In Table B-26 ($92 million Initial capital cost,
$3 milllon/yr operating costs for ten years, and $5 million decommis-
sioning costs) with no discounting yields a cost (1977 dollars) of $127
million, or $860/m3. An earlier 8tudy(51) which assumed a much smaller
facility (5000 m3) developed capital costs of $60 million and annual
operating costa of $3,4 million, for unit costs per cubic meter at least
20 times the above figure. Despite this wide range of cost estimates,>
for every cubic meter of TRU waste volume reduction, there is a potential
cost saving of at least $860 for a non-retrievable facility; for a
retrievable facility, using figures shown In Table B-26, the cost savings
work out to at least $1400/m3.
173
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Page Intentionally Blank
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GLOSSARY
ActInldes'. A series of elements In the periodic table, beginning with
actinium (element No, 89) and continuing through lawreneium (element
No. 103). The series includes uranium (element No. 92) and all of
the man-made transuranium elements. All are radioactive.
Alpha partie Lei A positively charged particle emitted by certain radio-
active materials. It is made up of two neutrons and two protons bound
together. It is the least penetrating of the three common types of
radiation (alpha, beta, gamma) emitted by radioactive material, being
stopped by a sheet of paper. It is dangerous to plants, animals, or
man only if the alpha-emitting substance has entered the body.
Background radiation; The radiation in man's natural environment,
including cosmic rays and radiation from the naturally radioactive
elements, both outside and inside the bodies of men and animals.
It is also called natural radiation,
Beta radiation; See Decay, radioactive.
jk?ijLlng water reactor(BWR) ; A type of nuclear power reactor that employs
ordinary water (E^O) as coolant and moderator and allows bulk boiM.ng
in the core so that steam is generated in the primary reactor vessel.
Breeder^ reactor: A reactor that produces fissionable fuel as well as
consuming it, especially one that creates more than it consumes. The
new fissionable material is created by capture (in "fertile" materials)
of neutrons from fission. The process by which this occurs is known
as breeding.
Calcine; (As a verb) To roast under oxidizing conditions. (As a noun)
The solid product of a roasting treatment under oxidizing conditions.
Canister: A metallic container for waste.
Cask; A container for transporting waste canisters or for spent fuel.
Chopper; Device for cutting spent fuel elements into smaller lengths.
Cladding; The outer jacket of nuclear fuel elements. It prevents
corrosion of the fuel and the release of fission products into the
coolant. Aluminum or its alloys, stainless steel, and zirconium
alloys are common cladding materials. Synonym: hull.
Curie; The basic unit to describe the intensity of radioactivity in a
material. The curie is equal to 37 billion disintegrations per second,
which is approximately the rate of decay of 1 gram of radium. A curie
is also a quantity of any nuclide having 1 curie of radioactivity.
B--I-1
-------
GLOSSARY (continued)
Named for Marie and Pierre Curie, who discovered radium in 1898. Ihe
prefixes milli-, micro- and nano- are frequently used and indicate
quantities of 10~3 curie, 10~6 curie and 10"^ curie, respectively,
Daughter: Ihe nuclide remaining after the radioactive decay of an atomic
nucleus. In turn, this daughter may be either stable or radioactive.
Decay heat! The energy released when radioactive nuclides change to
their daughter nuclides.
Decay? radioactive: The spontaneous transformation of one nuclide into
a different nuclide (or into a different energy state of the same
nuclide). The process results in a decrease, with time, of the number
of the original radioactive atoms in a sample. It involves the emission
from the nucleus of alpha particles, beta particles (electrons) or gamma
rays (electromagnetic radiation); or the nuclear capture or ejection of
orbital electrons. Also called radioactive disintegration.
Decontamination factor; The ratio of the amount of a given type of radio-
active material entering a process (or process step) to that which
leaves the process (or process step).
Disposal: Isolating the radioactive waste permanently in a form and
manner with no intent to retrieve it.
Element; One of the 103 known chemical substances that cannot be divided
into simpler substances 'by chemical means. A substance whose atoms
all have the same atomic number. Examples; hydrogen, lead, uranium.
(Not to be confused with fuel element.)
Fertile material; Material in which fissile isotopes can be produced
by neutron capture.
Fines; Small-size-range powder.
Fission: The splitting of a heavy nucleus into two approximately equal
parts (which are nuclei of lighter elements) , accompanied by the
release of a relatively large amount of energy and generally one
or more neutrons. Fission can occur spontaneously, but usually is
caused by nuclear absorption of neutrons or other particles.
Fission products: The nuclei (fission fragments) formed by the fission
of heavy elements, plus the nuclides formed by the fission fragments'
radioactive decay.
Fuel: Fissionable material used or usable to produce energy in a reactor.
Also applied to a mixture, such as natural uranium, in which only part
of the atoms are readily fissionable, if the mixture can be made to
sustain a chain reaction.
3-1-2
-------
GLOSSARY (continued)
Fuel assembly; A unit containing clad pieces of nuclear fuel for insertion
into the core of a nuclear reactor. An integral part of the fuel element
is the cladding provided to protect the fuel from corrosion by the reactor
coolant and to contain the fission products formed during irradiation.
Fuel bundle; Same as Fuel assembly (q.v.).
Fuel cycle: The series of steps involved in supplying fuel for nuclear
power reactors. It includes mining, refining, the original fabrication
of fuel elements, their use in a reactor, chemical processing to recover
the fissionatle material remaining in the spent fuel, re-enrichment of
the fuel material, and refabrication into new fuel elements.
Fuel element; Same as Fuel assembly (q.v.).
Fuel reprocessing: The processing of reactor fuel to recover the unused
fissionable material.
Fuel rod; A tube containing UOo or mixed oxide fuel; part of a fuel assembly.
Gamma radiation; See Decay, radioactive.
Glass ceramic: A ceramic material produced by the controlled devitrifi-
cation (crystallization) of a glass.
Glassification: Incorporation into glass.
Half-life; The time in which half the atoms of a particular radioactive
substance disintegrate to another nuclear form. Measured half-lives
vary from millionths of a second to billions of years.
High-level waste: The highly radioactive waste resulting from the
reprocessing of spent fuel to separate uranium and plutonium from
the fission products. The term includes the high-level liquid wastes
(HLLW) produced directly in reprocessing, and the solid high-level
wastes (HL¥) which can be made therefrom.
Hull: See Cladding.
Iji-can melting; A process in which solids are converted to a molten
glass in the very canister in which it is intended later to bury
the glass.
Iodine: An easily-volatilized fission product (q.v.).
Jon: An atom or molecule that has lost or gained one or more electrons.
By this lonization it becomes electrically charged. Examples: an
alpha particle, which is a helium atom minus two electrons; a proton,
which is a hydrogen atom minus its electron.
B-I-3
-------
GLOSSARY (continued)
Isotope; One of two or more atoms with the same atomic number (the same
chemical element) but with different atomic weights. Isotopes have
very nearly the same chemical properties, but different nuclear
(radioactive-decay) properties. Thus, for the element carbon, for
example, the isotope of atomic weight 12 (C-12) and the isotope of
atomic weight 14 (C-14) behave identically in chemical reactions;
but whereas C-12 is not radioactive, C-14 is radioactive, decaying
with a 5730-year half-life to stable nitrogen (N-14) with release
of a beta particle.
A gaseous fission product (q.v.)«
Leaching: Extracting material from a solid by contacting it with water
or with a solution.
Light water reac t or : A reactor in which ordinary water (I^O) is used as
the coolant. In such reactors the water is either allowed to boil
(boiling water reactor or BWR) or pressurized to prevent boiling
(pressurized water reactor or PWR) .
Loadings Amount of waste contained per unit of volume.
Low-level waste: Waste containing types and concentrations of radio-
activity such that shielding to prevent personnel exposure is not
required.
Matrix; The base material throughout which other materials are dispersed.
Metal matrix; An arrangement whereby aggregates of radioactive matter
are dispersed in a continuous metallic block.
Mil; One thousandth of an inch.
Mill; One thousandth of a dollar,
Mixed-pxide^ ^uel^ vcyjcle: A fuel cycle (q.v.) in which fuel containing
both uranium oxide and plutonium oxide is fed to the reactors. Such
a cycle requires reprocessing of spent fuel to recover the residual
uranium and the plutonium for fabrication of fuel elements.
Monolith: A solid material composed of a single phase, uninterrupted
by cracks or voids.
Natural uranium: Uranium as found in nature, containing 0.7% of U-235,
99.3% of U-238, and a trace of U-234. It is also called normal uranium.
B-I-4
-------
GLOSSARY (continued)
Nuclear criticality; The existence of all conditions needed to initiate
a nuclear fission chain reaction, as in a nuclear reactor or in an
atomic bomb.
Nuclide: A general term applicable to all atomic forms of the elements.
The term is often used erroneously as a synonym for "isotope," which
properly has a more limited definition. Whereas isotopes are the
various forms of a single element (hence are a family of miclides)
and all have the same atomic number and number of protons, nuclides
comprise all the isotoplc forms of all the elements.
Off-gas: The gas given off in any process step,
Plasma: A gas at a temperature sufficiently high that the gas molecules
become completely ionized. This occurs at temperatures of tens of
thousands of degrees C or more.
Plenum spring: A small spacing element placed inside the cladding end
cap of a fuel rod to maintain the fuel pellets in a stable configuration.
Plutonium? A heavy, radioactive, man-made metallic element with atomic
number 94. Its most important isotope is fissionable plutonium-239,
produced by neutron irradiation of uranium-238. It is used for reactor
fuel and in weapons.
Pressurized water reactor (PWR): A type of power reactor that employs
ordinary water (l^O) as coolant and moderator and is pressurized to
keep the exit coolant stream from boiling.
Pyrolysis: Thermal breakdown of a solid or liquid to form gases, in the
absence of sufficient air to burn these gases completely.
Rad_: A measure, applicable to any form of ionizing radiation, of actual
energy absorption, being defined as the amount of energy imparted to
matter by ionizing radiation per unit mass of irradiated material at
the place of interest. One rad corresponds to the absorption of 100 ergs
of energy per gram (100 ergs = 6.24 x 10^ million electron volts, Mev).
In interpreting standards of radiation protection, it is generally
permissible to consider 1-roentgen exposure to gamma or x-rays as
roughly equivalent to a dose of 1 rad in soft tissue.
RadiatioTi; The emission and propagation of energy through matter or
space by means of electromagnetic disturbances which display both
wave-like and particle-like behavior; in this context, the "particles"
are known as photons. Also, the energy so propagated. The term has
been extended to include streams of fast-moving particles (alpha and
beta particles, free neutrons, cosmic radiation, etc.). Nuclear
B-I-5
-------
GLOSSARY (continued)
radiation is that emitted from atomic nuclei in various nuclear reactions,
including alpha, beta, and gamma radiation and neutrons.
Radioactivity; The spontaneous decay or disintegration of an unstable
atomic nucleus, usually accompanied by the emission of ionizing
radiation,
Radioisotope: A radioactive isotope. An unstable isotope of an element
that decays or disintegrates spontaneously, emitting radiation. More
than 1300 natural and artificial radioisotopes have been identified.
Radionuclide: A radioactive nuclide. Thus, carbon-14 (C-14) is a radio-
nuclide because it decays radioactively to nitrogen-14 (N-14).
Radwaste; A contraction of the term "radioactive waste."
Recycle: The returning of uranium and plutonium (recovered in spent fuel
reprocessing) for reuse in new reactor fuel elements.
Rent; A dose unit which t'akes into account the relative biological
effectiveness (RBE) of the radiation. The rem ("roentgen ^quiv—
alent man") is defined as the dose of a particular type of radiation
required to produce the same biological effect as one roentgen of
(0.25 Mev) gamma radiation. A 1-rad dose of alpha particles is
approximately equivalent in its biological effects to 10 rads of
gamma radiation, and hence may be expressed as 10 rems. A milli-
rem (mrem) is one thousandth of a rem.
Roentgen (R); A measure of the ability of gamma or X rays to produce
ionization in air. One roentgen corresponds to the absorption of
about 86 ergs (100 ergs = 6.24 x 10' million electron volts, Mev)
of energy from X- or gamma radiation, per gram of air. The corre-
sponding absorption of energy in tissue may be from one-half to two
times as great, depending on the energy of the radiation and the
chemical composition of the tissue. The roentgen is thus more useful
as a measure of the amount of gamma or X rays to which one is exposed
than as a measure of the dose of such radiation actually received.
Ruthenium; A sometimes-volatilized fission product (q.v.)«
Salt jiake: The solid residue resulting from a concentration of high-level
liquid waste in underground waste storage tanks,
Saltdome: A geologic salt formation in which a plug of salt has been
thrust up through rock at some depth, leading to a subterranean
"cylinder" of salt which may be a mile or more in diameter and
several miles deep.
B-I-6
-------
GLOSSARY (continued)
Scoping analysis: An analysis made in order to determine the upper and
lower bounds of a,phenomenon.
Sintering; Agglomeration of particles, achieved by heating them almost
to their melting point.
Source terms: The amounts of specific radioactive nuclides issuing from
a process or from a process step.
Spailing; Breaking off of small pieces from a larger mass.
Sparge; Bubbling of a gas through a liquid,
Spent fuel; Fuel after its use in a nuclear reactor. It then contains
fission products, activation products and actinides, many of which
are radioactive. Synonym: irradiated fuel.
Springline: The line of intersection of roof and walls.
Sup ere ale ine: A crystalline ceramic waste form prepared by mixing
selected additives with HLLW and then calcining.
Throwaway fuel cycle: One in which the spent fuel is disposed of
directly rather than reprocessed.
Transuranic elements^ Elements with atomic numbers greater than 92,
They include neptunium, plutonium, americium, curium, and others.
Transuranic waste: Any waste material measured or assumed to contain
more than a specified concentration of transuranic elements.
Tritium: A radioactive isotope of hydrogen, of atomic weight 3.
Tritium (H-3) has a half-life of 12.3 years.
Uplift: The vertical rise on the surface of the earth caused by thermal
expansion in the case of a waste repository.
Uranium: A radioactive element with the atomic number 92 and, as found
in natural ores, an average atomic weight of approximately 238. The
two principal natural isotopes are uranium-235 (0.7% of natural uranium)
which is fissionable, and uranium-238 (99,3% of natural uranium) which
is fertile. Natural uranium also includes a minute amount of uranium-234.
Uranium is the basic raw material of nuclear energy.
Waste, radioactive: Equipment and materials (from nuclear operations)
that are radioactive and for which there is no further use.
B-l-7
-------
ABBREVIATIONS
AGNS Allied Gulf Nuclear Services
ARHCO Atlantic Richfield Hanford Company
BNFP British Nuclear Fuels Processing
BNWL Battelle Northwest Laboratories
BWR Boiling Water Reactor
CANDU Canadian Deuterium (Heavy Water) Reactor
DF Decontamination Factor
EIS Environmental Impact Statement
ERDA Energy Research and Development Administration
EPA Environmental Protection Agency
FP Fission Product
GE General Electric
HEDL Hanford Engineering Development Laboratory
HLLW High-Level Liquid Waste
HLW High-Level Waste
HM Heavy Metal (Uranium and Plutonium)
HTGR High Temperature Gas-Cooled Reactor
HWR Heavy Water Reactor
ICPP Idaho Chemical Processing Plant
ILLW Intermediate-Level Liquid Waste
INEL Idaho National Engineering Laboratory
LASL Los Alamos Scientific Laboratory
LMFBR Liquid Metal Fast Breeder Reactor
LTA Low Temperature Adsorption
LWR Light Water Reactor
MTU Metric Ton of Uranium
NRC Nuclear Regulatory Commission
ORNL Oak Ridge National Laboratory
ORGDP Oak Ridge Gaseous Diffusion Plant
PVC Polyvinyl Chloride
PWR Pressurized Water Reactor
TRU Iransuranic
WSEP Waste Solidification Engineering Project
o
A Angstrom unit (10~° cm)
Ci Curie
d/m Disintegrations per minute
GW Gigawatts (1 GW = 1Q6 kW)
GW-yr Gigawatt-year
kg kilogram
kW kilowatt
m Meter
mCi Millicurie
MCi Megacurie
uCi Microcurie
mR mllliRoentgen
mrem millirem
B-I-8
-------
ABBREVIATIONS (continued)
MT Metric ton
MTHM Metric ton of heavy metal (Uranium and Plutonium)
MWe Megawatts electrical
MWt Megawatts thermal
MWt-yr Megawatt thermal-year
nCi Nanocurie
B-I-9
-------
APPENDIX B-II
SOLIDIFICATION PROCESSES IN EARLY STAGES OF DEVELOPMENT
B-II.l SUPERCALCINE
ERDA 76-43 describes aupercalclne as a crystalline ceramic waste
form prepared by mixing selected additives with high-level liquid waste
(HLLW) and then calcining^-^ , The result is a mixture of thermodynami-
cally stable crystalline phases that are more refractory and leach-
resistant than ordinary oxide calcine formed without additives.
Work on the development of supercalcine is being conducted at
Pennsylvania State University, fixing HLLW in stable inert crystalline
phases. Waste streams containing 35-45% by weight of additives, usually
including Ca, Sr, Al, and SI, can be calcined by any of the previously
described methods. One nonradioactive engineering-scale run has been
made at BNWL. Figure B-II-1 shows two of a number of possible options
for processing waste as a supercalcine.
Several more years will be required to develop the full potential
of the supercalcine process for all HLLW constituents to be incorporated
into optimally stable crystalline forms.
If the supercalcine is part of a composite waste form, several Inte-
grated processing steps, such as pelletizing and application of coating,
must be developed. It may be ten years before a p-actical design can be
developed that would use supercalcine for the core of a multi-barrier
waste form.
Laboratory work is continuing in the following areas:
• Refinement of additives and formulations.
• Reduction of potential volatilization losses of radio-
active materials not currently included in crystalline
phases.
• Development of consolidation methods for supercalcine.
* Studies of leachability and thermal stability.
* Assessment of radiation effects.
B-II.2 SINTERING PROCESSES
If calcine is mixed with the proper flux or frit and sintered, the
product will have both a glass and crystalline phase and will be more
leach resistant, have high impact strength, be resistant to thermal
B-II-1
-------
MI IUU
PLUS -*
CALCINER
TAILOR-WADE INtRT~l
— X-RAY AMORPHOUS 4
ADDITIVES
SUPERCALCINE
I. FOR SOLIDIFICATION AS SUPERCALCiNE
2. USING SUPERCALCINE AS THE CORE OF A
MULT1BARR1ER WASTE FORM
HEAT IN CANISTER
TO CRYSTALLIZE PX
COATINGS
PELLET FORM IMG
(INCLUDES m-lVXfa
THERMAL TREATMENT!
! /
•
METAL
MATRIX
7t
STORAGE
CANISTER
FIGURE B-ll-1 SUPERCALCINE PROCESSES
OFF GAS
HLLW
BINDER -
GLASS FRII •
I. INEL PROCESS
2. PNl PROCESS
METAl MATRIX
(OPTIONAy
I,
STORAGE CAKISTtR
FIGURE B-ll-2 SINTERING PROCESSES
Source; Alternatives for Managing Wastes from Reactors and
Post-Fission Operations in the LWR Fuel Cycle.
E1DA 76-43, U.S. Energy Research and Development
Administration, May 1976,
B-II-2
-------
shock, and remain solid at temperatures up to 800°C. The product can be
formed as pellets or as large thick discs.
The Idaho National Engineering Laboratory (INEL) and BNWL processes
for producing these sintered glass ceramics (SGC) are illustrated in
Figure B-I1-2.
Frits used by INEL and BNWL differ somewhat . Product densities
range from 2.4-3,3 g/cm-'. Specific volumes of the reference waste would
be 30-55 liters/MTU. The leach rates of well-sintered products are simi-
lar to those of glass.
Various methods of preparing the material for sintering and various
post-treatments are under investigation. Work thus far has been on a
laboratory scale only; future work will concentrate on development of
process equipment and design verification. Experimental development may
be required for product forming, moving the containers through sintering
furnaces, and off-gas cleanup. Pilot-plant operation is scheduled for
1977. More work must also be done on product characterization. The
design of a full-scale demonstration plant may be started in 1979 or 1980.
B-II.3 METAL MATRICES
Metal matrices are monoliths with high thermal conductivity composed
of calcine, vitreous beads, or pellets cast in molten metal or embedded
in sintered metal. Alloys of lead, zinc, aluminum, or other low-melting
metals are used in casting; iron, iron alloys, or copper are used in sin-
tering. Waste constitutes up to 65% of the matrix.
The main advantage of these matrices is that they provide a high-
conductivity form to decrease centerline temperatures. This would tend
to make them advantageous for use with "young" wastes where the heat dis-
sipation problems are more severe. Their stability may be limited to
temperatures below the melting point of the metal to minimize slow oxida-
tion or reaction of the metal with the waste or canister. The matrices
vary in leach resistance, depending on the form of the waste incorporated.
Figure B-II-3 is a schematic illustration of some methods of forming
a metal matrix.
Cast matrices can be formed either by introducing the formed waste
near the bottom of the canister and letting it displace the molten metal,
or by forcing molten metal by pressure or vacuum through a canister of
waste particles. Sintered matrices can be formed either by mixing wastes
and metal powder, compacting the mixture into a pellet and sintering, or
by vibrating a metal powder throughout a canister loaded with formed
waste and sintering the mixture into a coherent mass.
Matrices have been formed and tested on a laboratory scale. Casting
of vitreous beads by molten metal displacement has been verified on a non-
radioactive pilot-plant scale.
B-II-3
-------
HLIW
MOLTEN
METAL ""
OFF GAS
^STORAGE CANISTER
MOLTEN METAL CASTING
METAL POWDER
METAL
P3WDER
STORAGE
I CANISTER
1. PNL PROCESS
VIBRATOR 2. 1 NFL PROCESS
POWDER METAL SINTERING
•OFF CAS
••CANISTER1ZATION
FIGURE 6-11-3 METAL MATRIX FORMATION
HILW —
GLASS FRIT
AND NUCLEATING ~»
AGENT
TH1N-
CONT
*j CALf
1
INER
r
CONTINUOUS MELT
r"3
VALLED
MNER
b—
^
t
ffiJ
+ff=^
r* urr-ufl^
TUNNEL KILN ^ r~£l
METAL MATRIX
lOPTIOKAU
STORAGE
CANISTER
FIGURE B-ll-4 GLASS CERAMIC PROCESS
Source: Alternatives for Managing Wastes from Reactors and
Post-Fission Operations in the LWE Fuel Cycle, 1KDA
76-43, U.S. Energy Research and Development Administra-
tion, May 1976,
B-II-4
-------
ERDA 76-43 lists the major development efforts required:
• Evaluation of matrix formation properties of new waste
compositions .•
* Development of best methods for use under remote oper-
ating conditions.
• Investigation of methods for casting fines or reducing
percentage of fines.
* Verification of methods of metering particles and molten
metal to the casting container.
* Determination of best process techniques for receiving
waste, canister hookup, and removal from matrix-forming
apparatus.
* Study of methods for compacting mixes for sintering
operations,
* Continuation of product characterization studies.
• Operation of non-radioactive pilot plant to verify a
conceptual process.
Design for a full-scale facility could start in late 1978 or early
1979.
B-II.4 GLASS-CERAMICS
If a glass is subjected to a controlled crystallization, a fine-
grained crystalline body with some residual glass phase may be produced.
This glass-ceramic is stronger than glass, is not as susceptible to
uncontrolled devitrification at high storage temperatures, and has leach
resistance similar to that of glass. A thin cross section (3 cm or
less) is needed to allow rapid cooling and to maintain uniform temperatures
during processing, in order to control nucleation and crystal growth.
One acceptable geometry would be a disc, which could be stacked into a
canister with a filler (probably metal), to aid heat transfer between the
discs and canister walls.
Because work thus far has been limited to laboratory scale develop-
ment, the process is only a concept at present. It is illustrated in
Figure B-II-4.
To develop the best compositions and characterize their behavior
during the waste fixation and storage process would take 2-5 years of
major laboratory research. Engineering development of the glass-casting
and heat treating processes would also require a major effort W.
B-II-5
-------
Because there is now no such major effort In the United States to
develop this process, although work has been done in Germany, it is
difficult to estimate when a production facility- could be designed. It
would probably be well after 1980.
B-II.5 COATED PELLEtS
Leach resistance of solidified wastes can be increased by forming
the waste into pellets and coating the pellets with a variety of mater-
ials, including pyrolytic carbon, SiC, Si<>2, and Al^O-j. The coatings
could be applied by chemical vapor deposition in a fluidized-bed or drum
coater, by plasma spray, or by an enameling process.
Because the coated pellets have poor thermal conductivity and their
coatings could chip during handling and transport, it may be desirable
to incorporate the pellets into a metal matrix.
The possible combinations of waste form, coating type, and composi-
tion of metal matrix are still being evaluated. Besides the process prob-
lems that must be overcome in determining which of the above choices is
best, the additional handling involved in moving materials from calciner
to pellet-former to coater to canister will create more complications.
The resultant complexity may affect process reliability,
ERDA 76-43 identifies the following major developmental efforts
required:
• Selection of pelletizing and coating process,
* Optimization of the matrix and characterization of the
product, and
• Design of equipment for remote operation.
It will probably be five years or more before conceptual design can
begin.
B-II.6 ION EXCHANGE
Sandia Laboratories is developing a method of disposing of high-
level radionuclides by first fixing them on ion-exchange media and then
compacting and sintering the media. A simplified flowsheet of the proc-
ess is shown in Figure B-II-5.
The HLLW is neutralized by the addition of NaOH and clarified by
centrlfugation. The clarified waste is then almost completely decon-
taminated by ion exchange with sodium titanate (ST). On a laboratory
scale, the effluent from the ST ion exchange has been reduced to less
than 10"*^ Ci/ml of total activity; most of the carryover radioactivity
B-II-6
-------
NaOH
HLLW-
j NEUTRALIZATION
CLIANUPREAGENTO
CENTRIFUGE!*
AND WASH 1
SUSPENDED
SOLIDS
SODIUM T1TANATE _,
ION EXCHANGE BED
LOADED ST
A
AS *-
— 1
4
PAN FILTER DRIER
A
CONSOLIDATION
BY PRESSURE
SINTERING-llOft, ~WOKq/c
, ZEOLITE BED
FOR CESIUM
i
LOADED
ZEOLITE
-» CANI5
n?
CliANUP OF Tc,
RuANONONIONlCS
CLEANUP SOLIDS
TO DISPOSAL
LOW LEVEL SODIUM
SALT SOU DS TO
DISPOSAL
SECONDARY WASTES
UNIQUE TO THIS PROCESS
FIGURE B-ll-5 ION EXCHANGE FIXATION
Source: Alternatives for Managing Wastes from Reactors and
Post-Fission Operations in the IMS. Fuel Cycle.
EKDA 76-43, U.S. Energy Research and Development
Administration, May 1976.
B-II-7
-------
is Ru» Cs» and Tc. Activity of alpha emitters was reduced to less than
10 Ci/ml. Cesium may be subsequently removed by ion exchange with a
zeolite, and remaining traces of such wastes may be removed with an
anion exchanger material and other reagents. The spent ST and zeolite
are removed as slurries, combined with the suspended solids removed in
the clarification step, dewatered and dried on a pan filter, and then
consolidated into cylinders by pressure sintering.
The major process effluents would be low-level dry sodium salt
solids and tritiated water. A small quantity of trace cleanup solids is
also expected,
The ion exchange process is in the bench-scale demonstration stage
and is expected to be ready for practical design of a full-scale plant
during 1981. The experimental program includes investigation of the
effectiveness of the ion exchanger, consolidation of the solidified
waste, and evaluation of the physical and chemical properties of the
resulting waste form.
B-II.7 GERMAN THERMITE PROCESS
A thermite process is being developed in Karlsruhe as a possible
alternative to the glass melting step in the German VERA process.
The thermite reaction produces a rugged, two-phase, ceramic—metal
material, with properties at least equivalent to glass.
A 2-kg batch process has been tested, and an apparatus has been
built for processing batches up to 20 kg to test the process on a larger
scale.
B-1I.8 GERMAN BOROSILICATE GLASS PROCESS
Work at Julich is directed toward perfecting a denitration, drum-
drier, in-can melting process for borosilicate glass.
Formaldehyde is added to the liquid waste until the pH rises to
about 3.0; then glass-forming additives (CaC03, 8102) are introduced,
which form a viscous slurry. The slurry is converted to a solid cake
that is continually scraped off the drier and dropped into a canister
where it is melted. This process has been demonstrated on a laboratory
scale for both radioactive and nonradioactive glasses.
B-II.9 GERMAN PHOSPHATE GLASS PROCESS
The Pamela process is being developed by Gelsenberg AG of Germany
in cooperation with the Eurochemic Company of Belgium. It is being sup-
ported by the German Ministry of Research and Technology.
B-II-8
-------
HLLW is denitrated with formaldehyde and concentrated with HgPOA
to form a phosphate glass. Originally a platinum melter was used; a
later version uses an electrically-heated ceramic melter. The glass
malt passes through a series of nozzles and the droplets are caught on
a rotating disc, where they solidify. The granulated glass is encapsu-
lated in a lead alloy using technology developed at Euroehemic. Vitri-
fication off-gas is fed back into the reducing atmosphere of the denitra-
tion unit, which retains any ruthenium that has volatilized.
Granules of phosphate glass that may contain about 30% by weight
of fission product oxides have Been favored so far. Work is under way,
however, to extend the process to the production of borosilicate glass
and borosilicate glass-ceramic products.
B-II.10 OTHER FOREIGN PROCESSES
Several solidification processes are currently being developed in
Russia, Italy, Japan, and Sweden. They combine features of the proc-
esses already discussed. They will therefore not be described in this
report.
REFERENCE
1. Alternatives for Managing Wastes from Reactors and Post-Fission
Operations in the LWR Fuel Cycle. ERDA 76-43, U. S. Energy Research
and Development Administration, May 1976.
B-II-9
-------
APPENDIX B-III
OFF-GAS TREATMENT TECHNOLOGY
This appendix discusses In detail the technology available for
various degrees of control of five gaseous materials generated during
solidification processes (see Section B-3.2.4 of the text):
1. Krypton, with its Kr-85 content;
2. Iodine and its compounds, with their 1-129 content;
3. Carbon compounds, with their C-14 content;
4. Hydrogen and water vapor, with their tritium
(H-3) content; and
5. Ruthenium compounds, with their Ru-106 content.
Each of these is presented in a separate section of this Appendix.
B-lIi.l KRYPTON
ERDA 76-43 gives annual Kr-85 production from a 1500-MT/yr LW1 re-
processing plant as 9.8 x 10^ Ci, or about 2 x 105 Ci/GW-yr.W These
figures (equivalent to about 6500 Cl/MT of fuel) are somewhat lower than
another estimate of 9700 Ci/MT.(2) Because Kr-85 has a moderately long
half-life (10.7 years), and is produced In large quantities, the U. S.
Environmental Protection Agency (EPA) has promulgated for the entire
fuel cycle Kr-85 discharge limits of 50,000 Ci/GW-yr,(3) Although the
need for these limits has been debated, since krypton is chemically
inert and does not present an Internal biological hazard, such con-
siderations are beyond the scope of this study.
When the EPA-proposed regulation is used and the less than 1% of
krypton that escapes at the reactor plant Is Ignored, the reprocessing
plant decontamination factor (DF) required to meet the 50,000 Ci/GW-yr
limit Is: 2 x 105/(0.5 x 1Q5) = 4.
Because krypton Is a noble gas, It is reasonable to assume that
essentially all the krypton Is given off as "off-gas" at the chopper and
dissolver steps and thus, Is readily removed into the off-gas stream.
According to EKDA 76-43, krypton will be present in the off-gas in
concentrations of about 0.003% by volume; xenon will be present in con-
centrations of 0.02% by volume. (*•) Thus, technology for krypton removal
must be capable of removing krypton present in the off-gas In small
quantities; the required decontamination factor to meet EPA standards Is
modest (about 4).
B-III-1
-------
Two technologies for krypton removal have been Investigated:
cryogenic distillation, and fluorocarbon absorption. Two technologies
for krypton storage have been Investigated: pressurized cylinder storage
and zeolite encapsulation.
B-III.1.1 Cryogenic Distillation
This process, identified in ERDA 76-43 as "currently available"
technology, Is the most commercially advanced of the two krypton removal
methods.(1) Cryogenics are used extensively for commercial gas separa-
tion. At the Idaho Chemical Processing Plant (ICPP), cryogenic distil-
lation has been used to obtain krypton from off-gas, though not at high
overall system efficiencies.
The boiling point of krypton (-157°C) is sufficiently higher than
that of oxygen (-183°C), nitrogen (-196°C), and carbon monoxide (-190°C)
to allow separation from these gases. However, gases will probably be
present with a boiling point higher than -157eC (N20, -90°C; N(>2, 21°C;
CC>2, -79°C; H20, 100°C) and may present freezing problems. Xenon, with
a boiling point of -107°C, and NO, with a boiling point of -152°C, will
tend to follow the krypton through the cryogenic process but can be
separated if desired.
At the cryogenic plant at Idaho, a catalytic converter (rhodium,
480°C) converted N£() to N£ + 02, and a cooler/drier (silica gel) removed
residual water and N02- Nonetheless, considerable difficulty was re-
ported with plugging of the cold trap and other process equipment due to
C02» H20 and NOX freezing, a difficulty that could probably be overcome
or the effects mitigated in a commercial-scale plant with increased
regeneration/redundancy design and surge volume.
The Idaho plant (a customized, uniquely designed unit) utilized a
continuous primary distillation column and periodic transfer of bottoms
(containing primarily noble gases with NO if present and some air
carryover) to a batch distillation column for final concentration of the
krypton. The equipment at the ICPP facility distilled krypton at an
efficiency of 97% and xenon at an efficiency of 98%. However, the whole
system efficiencies during operating periods were 52% and 63%, respec-
tively. These lower numbers should be viewed with the understanding that
the ICPP facility was designed to allow some krypton removal for com-
mercial use, not for environmental protection. Thus, high overall op-
erating efficiency was not a design objective: "The discrepancy between
overall system efficiency and cryogenic equipment efficiency is due
mainly to losses occurring during unanticipated process interruptions
and during startup-shutdown periods."W) Nonetheless, the efficiencies
achieved indicate that a distillation DF of ten is well within reason
for a commercial facility, although there is considerably less certainty
regarding the ability of a commercial facility to operate with suffi-
ciently high reliability to yield a system DF greater than ten. The
Idaho facility treated 240 liters/rain (8.4 standard ft-Vraln) of disaolver
off-gas at atmospheric pressures and 60°C
B-III-2
-------
The product of the ICPP krypton facility was typically high in con-
taminants (the product was shipped to Oak Ridge National Laboratory
[ORHL] for final purification). Of five runs reported, the average con-
tent was: 23% Kr, 60% Xe, 4% N2, 10% 02, 2% Ar, and detectable C02 and
NOX.(4) The highest krypton content was 51%, but this run contained
25% 02, a potentially undesirable storage, situation.
Considerable concern has been expressed regarding the potential
dangers of ozone production In a cryogenic facility. The main point of
concentration of the ozone appears to be in the primary distillation
column, where liquid oxygen can accumulate and be converted to ozone via
radiolysis (from the radioactive krypton). At cryogenic temperatures,
the ozone will be decomposed less easily. Sources at the Idaho labora-
tory suggested that the ozone problem in storage bottles is less sig-
nificant because of the higher rate of decomposition (confirmed by tests)
at or above ambient temperatures.
There are two options for dealing with the ozone problem in a cryo-
genic system. The oxygen may be removed ahead of the distillation proc-
ess, or the process may be designed so that residence times of liquid
oxygen with radiokrypton are minimized. At ICPP, the latter was done;
it is not clear that design of a commercial-scale facility would permit
this. Removal of oxygen and ozone has been performed with an 02/H2
recombiner; potential difficulties exist with this concept, including
the H2 explosion potential, and greater potential for freeze-plugging
of the system by reduction in the 02 level (NOX and noble gases are
less soluble in LN2 than in liquid oxygen).
B-III.1.2 Fluorocarbon Absorption
Somewhat less commercially developed, but with fewer apparent prob-
lems, is the fluorocarbon system, whereby a fluorocarbon ("R-12":
dichlorodifluoromethane) selectively absorbs impurities, which are then
stripped from the fluorocarbon.
At the Oak Ridge Gaseous Diffusion Plant (ORGDP) , pilot plant tests
at 210-620 liters/min (7.5-22 standard ft3/min) have been performed
with a fluorocarbon system that uses radioactive krypton in nitrogen
feed gas containing various amounts of 02, C02, nitrogen oxides, iodine,
and xenon- (5) The pilot plant consisted of a primary absorption/
stripping operation and peripheral equipment required for recovery,
purification, and recycle of solvent and some purification of krypton
product.
The advantage of the fluorocarbon system is that Kr, C02, N02,
and I2 all have substantially higher affinity for R-12 than the main
off-gas components 02 and N2. Xenon also will be selectively absorbed
by the R-12. This characteristic makes the system potentially very at-
tractive as an integrated Kr-85, C-14, 1-129 removal system.
B-III-3
-------
The testa at the ORGDP pilot facility yielded krypton DFs of
greater than 10^, Ct>2 removal DF's of greater than 10^, iodine DFs
(both 013! and 12) of greater than 10^, and N02 removal of greater than
100. The stripper product gas will not be pure Kr but will contain
quantities (depending on feed gas) of Xe, Ar, and C(>2. The iodine,
NC>2> and 1^0 may be removed in the solvent purification step. Figure-
B-III-1 shows the various gas exhaust points. A program is in progress
to determine ways of removing contaminants from the stripper off-gas,
ORGDP representatives state that argon may be desirable as a diluting
gas for krypton storage, though this Is not clear since only 6% of the
krypton Is expected to be Kr-85.'l>
Some 02, N2, and R-12 vapor will also be present in the product gas.
A cold trap could be used to trap the C02, Xe, and R-12, though not the
02, N2, or Ar. The C02 could then be removed from the cold-trap re-
generate in a caustic scrubber (see section on carbon-14) , Alternatively,
an inline COg caustic scrubber could be used. A reactive metal trap
(copper or manganese) may be tested for 02 removal, and a titanium metal
trap for N2 removal. The product purification features have not been
integrally tested at the pilot plant.
Iodine, N02, and t^O will remain in the R-12 through the stripper
and are removed in the solvent still. T^, CH-jI, NO^j and 1^0 leave the
solvent still as liquids or solids in solution, and several tests at
ORGDP have confirmed the removal of iodine at high DF's. Separation of
the iodine from N02 and 1^0 has not been developed at ORGDP, and it is
not clear that this is a simple step. Use of a solid adsorbent has been
suggested, but this technique has economic drawbacks for total iodine
removal (see B-III.2). Cold-freezing has been suggested, but It does
not appear this would be successful for CH3I. An alternative solution
may be removal of iodine prior to the fluorocarbon process.
The ORGDP pilot plant has accumulated over 6000 hours of operating
time. A reliability analysis has shown that equipment availability is
90% at the pilot facility; a program is currently underway to determine
what design features are necessary for a 99% availability at a demon-
stration facility.
Automatic control response to variations in feed gas flows and
composition has not been tested completely.
B-III.1.3 Low-Temperature Adsorption
Although this method Is not discussed in ERDA 76-43, low-temperature
adsorption (LTA) is worth mentioning in any discussion of noble gas
removal. At ICPP an LTA process for noble gas removal was investigated
but abandoned because of questions of efficiency and freeze-up problems
with contaminants such as t^O, (X^, NC>2» and ^0. General Atomic
Company has successfully used an LTA system for cleanup of helium streams
in a high-temperature gas-cooled reactor (HTGR) application, but H^O
B-II1-4
-------
Volatile Soluble
Components Kr, C02, Xe
Solvent
Recycle
i
Ul
High Boiling
Components
I2, CH3I,N02
H2O
Decontaminated
Vent Gas
FIGURE B-lll-1
SCHEMATIC DIAGRAM OF THE ORGDP SELECTIVE
ABSORPTION PILOT PLANT
Contaminated
Feed Gas
2
Ar
co2
H2O
NO2
NO
N20
CH3I
Kr
Xe
RuO^
CH,
lource; Stephenson, M.J. and R. J. Eby. Development of the 'FASTER' Process for
Removing Krypton-85, Carbon-14, and Other Contaminants. Presented at the 14th
ERDA Air Cleaning Conference, August 1976.
-------
at a concentration of more than a few parts per million rapidly plugs
the cold-traps.
B-III.1.4 Other Systems
OBNL has Investigated a number of krypton-removal systems and
^concluded; "Based on the preceding review of krypton recovery methods,
it appears that there are only two processes which are suitable for
adoption to a fuel reprocessing plant. These are fluorocarbon absorption
and cryogenic distillation. The other processes either suffer from
major technical problems or lack of development effort."'"'
B-III.1.5 Cylinder Storage
Because of the long history of commercial use of pressurized cyl-
inders for storage of gases, including highly corrosive gases, the tech-
nology for storage of krypton in pressurized bottles may be considered
to be available. However, some additional testing needs to be completed
on the specific application of steel cylinders to intermediate-term
(100-year) storage of the specific krypton removal product.
Steel-cylinder storage has the advantages of commercial experience
(including storage of toxic gases, with use of safety techniques such
as packless valves, valve caps, and ultra-conservative pressure margins
with non-venting cylinders), long life (at least 500 years has been
estimated from testing of 1000 cylinders), good thermal and radiation
resistance, low cost, and ease of recovery. Use of stainless steel
would add corrosion resistance.
The main disadvantage, the potential for sudden catastrophic failure,
is not a serious one since total failure of any one cylinder would not
result in significant public health hazard.
The remaining uncertainties regarding cylinder storage mainly in-
volve the internal corrosion effects of non-krypton contaminants in the,
stored gas: ozone (if present), oxygen, NOX, and water. An optimum
cylinder curie-loading must be determined on the basis of allowable heat
rate and resultant corrosion rates. The cryogenic distillation process
used at Idaho, for example, leaves substantial 02 in the product gas.
The contaminant corrosion problem would be minimized by additional
product purification, or by a probably more cost-effective approach,
monitoring of sample cylinder performance during storage life. Any
resultant problems (unexpectedly high corrosion rate) could be dealt
with by transfer to new cylinders, a relatively simple operation.
Mention has also been made of solid daughter (rubidium) buildup on
the cylinders as a disposal problem. W) For example, a 500-psi cylinder
with an initial loading of 128,000 Ci Kr-85 will, after 100 years, con-
tain 0.3 kg of rubidium (as a solid at ambient temperature).
3-1II-6
-------
Rubidium has chemical characteristics similar to those of sodium, potas-
sium, and cesium; it will react' vigorously with water and will ignite
spontaneously In air. Studies of sodium storage In LMFBRs suggest that
rubidium will not chemically attack the cylinders.
A cylinder test program Is currently in progress at Idaho, with
test specimens subjected to various conditions of pressure, radiation
(gamma), and temperature.
B-III.1.6 Zeolite
Zeolites are crystalline aluminosilicates with uniform arrays of
pores; some are available with 4 A pore size. A krypton atom at ambient
temperature has a diameter of about 3.5 A. -Laboratory tests with pure
krypton have shown that at elevated temperatures and pressures, krypton
will diffuse into the pores; at ambient temperatures, the krypton will
remain in the pore structure.W About 1.6 times more storage volume
would be required with zeolite than with 500-psi cylinders.(9) The
rate of release from the cylinder in an accident would, of course, be
less if zeolite encapsulation were employed.
The atoms of gther gases have diameters close to or less than
krypton (Ar: 2.9 A; CC^: 3.3 A; N2: 3.2 A; 02: 3 i), and, therefore,
presumably wguld also be trapped in the zeolite. Xenon is-slightly
larger (3.7 A) and could conceivably be discriminated against. There is
some question as to whether in a high-xenon environment, the xenon might
not interfere with krypton capture. The effect of rubidium buildup is
also not clearly defined, although sodium experiments suggest that
rubidium would not be a problem.
B-III.1.7 Other Storage Sy&>,ems
(9)
Clathrates have been investigated as krypton-stabilizing substances.
Water solubility, susceptibility to oxidation, and thermal instability
make clathrates appear to be unpromising.
B-III.1.8 Summary^ ^rypton Recovery and Storage
Despite the successful operation of a cryogenic distillation krypton
removal facility at ICPP, there are a number of potential difficulties in
applying the technology to a large commercial facility in order to meet
the 50,000 Ci/GW-yr limit. The main question is expected reliability and
derives from the freeze-up problems experienced at ICPP. Provision of
redundant trains and rapid regeneration equipment, or large gas surge
capabilities, or front-end treatment equipment for 1^0, C02, and NOX,
could probably increase overall system reliability.
B-III-7
-------
On a test scale, the fluorocarbon system appears to be more prom- i
ising than cryogenic distillation and has the added potential advantage
of economical removal of CC>2 and perhaps iodine. However, the technology
of fluorocarbon absorption is not yet fully developed, even at a test
scale, especially at the product purification steps. Nevertheless,
reliability analyses have been encouraging regarding the potential system
availability in commercial application.
Because cylinder storage is an existing, proven technology, with ease
of transfer of stored gas in the event of unexpected difficulties, this
appears to be the most reasonable technology to assume for krypton storage
for the time frame required for an isotope with short (10,7-yr) half-life.
B-III.2 IODINE
After nuclear fuel has cooled for several months, the only radioiso-
tope of iodine from a fuel reprocessing plant that is of concern is 1-129
(1.6 x 10? yr half-life). Estimated uncontrolled release of 1-129 from
a 5 Ml/day LWR fuel reprocessing plant is 3.8 x 10^ g/yr, which is more
than the estimated pre-1940 worldwide inventory of 2 x 10^ g.(lQ) This
estimated uncontrolled release of 1-129 from a 5 MT/day plant translates
to 60 Ci/yr, or about 1.2 Ci/GW-yr and is consistent with the estimate
given in Reference 2a.
The EPA has promulgated regulations limiting discharge of 1-129 from
the entire fuel cycle to 5 mCl/GW-yr,(3) Calculations show that less than
10~6 Ci/yr of 1-129 is released from a typical modern LWR.* Thus, It is
reasonable to assume that the only significant release of 1-129 is from
the reprocessing plant. The proposed regulation requiring removal of all
but 5 mCi/GW-yr means that only 5 mCi/1.2 Ci or 0.4% of the 1-129 may
be released from the fuel reprocessing plant. This decontamination fac-
tor of 250 may present a difficult engineering problem.
Estimates have been made that over 99% of the Iodine can be volatil-
ized into the process off-gas (at the dissolver step). 0.1) This estimate
is based on laboratory and hot cell tests at ORNTu with spent LWR fuel and
use of an air sparge on the dissolver, and establishment of optimum
temperatures, acidity, and residence time.** It cannot be assumed that
*I£ 0.1% failed fuel Is assumed, 1.2 mCi/GW-yr of 1-129 will leave the
fuel at an LWR, most of which will be trapped in managed waste streams.
**0n the basis of hot cell tests, it has been estimated that more than
99% of the radioiodine in the dissolver solutions could be removed,(12)
but it is questionable whether large-scale commercial facilities could
achieve this level of dissolver removal. Savannah River and ORNL have
achieved 99.5% in LWR fuel tests, but these tests used once-through
nitric acid, with no recycle system.
'B-III-8
-------
production-scale processes can achieve greater than 99% Iodine removal
at the dissolver, and even If 99% were removed and collected, over one-
half of the remaining 1% of the Iodine would have to be accounted for In
order to meet the proposed regulations.
Because of the volatility of Iodine, more than one-half the remaining
1% may enter the off-gas stream through evaporator/concentrator overheads
downstream of the dlssolver. Meeting EPA requirements may require treat--
ment of these overhead streams for Iodine removal. Because Iodine removal
efficiencies required to meet these requirements are much higher than has
ever been required before, there is little information defining where the
iodine will migrate after the dissolver, and this may pose problems.
Regardless of how the above potential regulatory problems are resolved,
very high iodine DFs are clearly desirable in the reprocessing off-gas
system.
In addition to the approximately 380 kg of 1-129 produced yearly
from a 5 MT/day plant, approximately 250 kg of stable 1-127 from fission
and from reprocessing chemicals will also be present in the off-gas
stream,(11) Thus, a design objective in potential iodine control systems
is the removal of approximately 600-650 kg of iodine per year at very
high decontamination factors (103 or better). Approximately 1-5% of
organic iodine (methyl iodide, CH-jI) may be present in the off-gas stream,
as well as HI and HOI.(2a-)
Four possible technologies have been identified for iodine removal:
caustic scrubbing, mercuric nitrate-nitric acid scrubbing (Mercurex),
nitric acid scrub (lodox), and chemlsorption by metal-loaded adsorbents.
B-III.2.1 Caustic Scrub
Caustic scrubbing has been used at both Idaho and Hanford for iodine
removal, but this method has been generally abandoned In planning for
iodine removal from commercial reprocessing plants, because of the very
low efficiency of removal of organic iodine forms (which may constitute
up to 5% of the iodine) and the low DF for elemental iodine (about 10). ^ '
In addition, high volumes of waste are generated (a 60 mVmln gas-flow
scrubber would produce 1800 ra^/yr of waste solution^)). There are no
active programs in the United States to develop caustic scrub for iodine
removal from reprocessing off-gas streams.
B-III.2.2 Mercurex*
The Mercurex process utilizes a mercuric nitrate-nitric acid scrub in a
packed tower, with periodic transfer of the scrub solution to a
concentrator-evaporator for precipitation of the iodine as Hg(IO ), .
*This discussion covers the process as developed in the United States.
Although work on this process has also been done in Europe, the status
of that work was not investigated.
B-III-9
-------
This process has been studied on a laboratory scale (30 llter/m or "^
1 ft-Vmin) at ORNL and at Savannah River, where elemental Iodine DFs
(at optimum Hg concentrations) of 2260-8400 and methyl iodide DFs of
10 and above were obtained.(12) x^e higher 1% DFs were at higher con-
centrations of HNOg, but higher nitric acid concentrations decreased ef-
ficiencies of 013! removal. (This confirmed similar results at OBNL.)
At Savannah River, a full-scale iodine control process is also in
operation using dilute solutions and DIs below the commercially acceptable
level. At the Allied Gulf Nuclear Services (AGNS) plant at Barnwell,
S.C., a Mercurex-type process has been installed that is conceptually
similar to a system used for iodine removal at a recovery facility in
Idaho. The Idaho process was run for several years at DFs comparable to
those expected at the AGNS plant (ten per scrubber, two scrubbers in
series). DF for elemental iodine is expected to be higher than 10; for
CH-I, less than 10.
The main disadvantage of the Mercurex process appears to be a high
mercury use (4:1 molar mercury to iodine), translating to 30 nH Of liquid
waste/yr and lack of commercial scale demonstration of either the
process or a solidification/recycle process system. The waste problem is
caused by high iodine carryover as the retained iodine builds up In the
scrubbing solution. The iodine carryover is in a form (unidentified) not
readily removed by a second Mercurex train in series. A solid adsorber
or lodox tower will apparently remove this Iodine form. In order to
minimize this carryover, mercuric iodide must be removed in large quantity
to maintain system efficiency.
The information available regarding testing of the Mercurex process
Indicates that there Is still some question regarding Its tolerance to
potential impurities in the off-gas (substances other' than 02, %, Ar,
H20). The AGHS off-gas control' system provides for NOX absorption between
the two iodine scrubbers; plans were to move this upstream of the Iodine
removal system if NOX interferences in iodine removal required it."a>'
B-III.2.3 lodox
The lodox process uses a highly concentrated nitric acid scrub proc-
ess in a bubble-cap tower to oxidize iodine to nonvolatile iodate (organic
iodine is oxidized to free iodine first). The resultant solution is
evaporated to a solid and results in a very low waste volume (0.4 m-Vyr).
The lodox process has been developed at ORNL through a pilot-scale plant.
DFs as high as 3 x 10^ were reported for seven plates at 30 liter/min
(1 standard ft-Vmin) (average DF of 4.4 per plate). Groenier and Hannaford
concluded that, "treatment of a 100 ft^/min off-gas stream to obtain a
DF of 10^ could theoretically be accomplished using a 20-inch diameter
bubble-cap column having 10 plates."(13) (DF of 3.2 per plate.) The
pilot-scale tests used a 7.6-cm (3-inch) diameter bubble—cap column. A
more conservative estimate has been that a DF of 10^ is achievable. (•*-)
B-III-10
-------
Work has also been performed with packed towers» hut the bubble-cap proc-
ess was stated to be preferred due to better control over a range of flow
rates. Results for CH-jI varied from no change from 12 efficiencies to
lower than 12 by as much as a factor of ten. High CC>2 levels did not
significantly affect results. Presence of certain organics (hexane,
notably) reduced results by a factor of five.
Apparently lodox results on a pilot-scale were very encouraging.
The main disadvantages of the lodox process are the highly corrosive
scrub liquid and the fact that nitrogen oxides (NOX) interfere with the
process by shifting the chemical equilibrium (this is not a severe prob-
lem at a few percent NOX and, in any event, is a system design problem,
not a fundamental one); the corrosion problem can probably be dealt with
satisfactorily (according to sources at ORNL, the U. S. Army has con-
siderable experience with high-concentration HHOg in stainless steel and
Zircaloy). The fact that the lodox process has not operated on a com-
mercial scale makes the success of its application to a reprocessing plant
uncertain. The lodox process (with solid adsorbent polisher) has been
selected for conceptual design of an LMFBR reprocessing system.
ORNL has recently been studying use of electrolytic scrubbing for
removal of iodine.(14) xhe principal advantage to this method is that
lower-concentration nitric acid may be used. Tests were performed on a
laboratory scale (1350 ml/min) with sample gas containing air, NOX» water
vapor, and Iodine (both elemental and organic). At about 100 ppm (vol)
013!, DFs of about 100 were obtained. For elemental iodine, 20-40 ppm,
DFs of about 600 were obtained. Best results for both l£ and CH-jI
appeared to be at 8M HN03; DFs ranged from 111 (2% NO present) to 685 for
I2» 5.6 (2% NO present) to 113 for CHoI, The results reported suggest
that NO in 1-2% quantities severely degrades the DF.(14) N02 also de-
grades performance: 1% N02 lowered CH^I DF from 100 to 10, and 12 DF
from 600 to about 150. Electrochemical reduction yields several reduc-
tion products: NO, ^0, H^ and ammonia.
Several methods of concentrating and storing iodine products from
the electrolytic process are being investigated; although there are no
conceptual problems with distilling off the nitric acid to leave iodine
residue, the technology has not been proved .
B-III.2.4 Chemisorptlon
Laboratory tests using silver-loaded zeolite (AgZ) have yielded DFs
of 1Q2 - 1Q5 for 12 and organic iodine in air streams. Tests at Idaho
National Engineering Laboratory (INEL) using N02 and 1^0 contaminants
yielded DFs reduced by 2-3. The chemlsorption technique has wide sup-
port because of Its simplicity and ease of waste handling; however, it
is Included in conceptual designs mainly as a "polisher" because of
concern over the high cost of silver.(") Annual silver cost for removal
of 600 kg (21,120 oz) of Iodine at 25% efficiency and $5/oz is about
B-III-11
-------
(21,120) ($5) iif|i . S356.000
This Is a possible alternative to more complicated, high-capital-
cost systems. Some concern has also been voiced regarding heat-removal
problems in large, commercial-scale beds, although for LVR fuel re-
processing, this would not be a problem because of the low residual amount
of radioiodine after a few months of fuel cooling.
The results of tests of AgZ at Savannah River with Iy and CIUI ,-»»
in a 7 liter/min air stream showed a DF of 10^ for !£, 25u for CHoI.
For PbZ a DF of 333 was achieved for 1 2 but only 1.8 for
-------
C-14 to be evolved from the.LWK. fuel cycle with that present as a
result of natural processes.
C-14, a radioisotope of carbon with a half-life of 5700 years, is
being produced continuously in the upper atmosphere by nuclear changes
in stable N-14 (caused indirectly by cosmic rays), Enough is produced
to support a relatively constant inventory of about 4.5 MCi (megacuries)
of C-14 in the atmosphere worldwide. This C-14 (about 1 MT) is inter-
spersed uniformly in the carbon contained in the CO of the atmosphere,
and ultimately in the carbon contained in every living thing, at a C-14
level of about 15 disintegrations/minute (d/m) per gram of total carbon.
When an organism dies and ceases to exchange with the atmosphere
(by photosynthesis, breathing, or eating of other life forms), its C-14
level drops, with a 5700-year half-life. This provides the basis for
the so-called radiocarbon dating technique used for determining the age
of ancient carbon-bearing relics by comparison of their present C-14
levels with an assumed value of 15 d/m per gram of carbon at the time
of their death in the past. Recent calibrations of this technique with
wood from the individual tree rings of bristlecone pine trees (thousands
of years old) show, however, that the C-14 abundance in the carbon of
the atmosphere (and thus of living things) has not always been constant;
4000 years ago, it was 61 higher than it is today, and 7000 years ago,
it was 10% higher than it is today. These changes may be associated
with changes in the earth's magnetic field,(17)
The total amount of C-14 that would ultimately be emitted, if
uncontrolled, from the operation of the 700 GW of U.S. reactor capacity
(see Task A Report) to be built by the year 2020 would be, at 28 Ci/GW-yr
(see below) about 700 x 28 x 30 = 0,6 MCi of C-14. If 40% of the C-14
emitted stays in the atmosphere, ^^and if the world total is triple
the U.S. figure, then the atmospheric inventory would, when all those
reactors shut down by 2040 or so, have risen by 3 x 0.6 x 0.4 or 0.7
MCi of C-14, or 16% of the natural C-14 background. Since the combus-
tion of fossil fuels (which have no C-14 left in them because of their
great age) should simultaneously increase the C0« content of the atmos-
phere by several percent of the present level, tne level of C-14 per
gram of carbon would rise by less than the maximum figure given above".
If one assumes a rise of 16% in the nominal human body content of
about 85 nCi of C-14, this should result in an additional annual
personal dose of about 0.2 mrem, or about l/500th of the present total
dose from all sources, or an increment equivalent to that received from
cosmic radiation in one hour of commercial air travel.^ '
Thus, as shown above, even 60 years from now, the effect on
personal background radiation would still be: a) at most a very small
quantity, and b) of the same order as that caused by previously-seen
naturally-occurring changes in C-14 levels. Although there are unre-
solved questions regarding the need for C-14 control in the near future,
the following discussion presents the current state of knowledge concerning
C-14 control technology.
B-III-13
-------
Quantities of C-14 are produced both at fuel reprocessing plants
and also at nuclear reactors. The major source of C-14 in nuclear fuel
probably comes from the (n,p) reaction with N-14, which is present as
an impurity in fuel; concentrations are of the order of 10 ppm» and some
values as high as 25 ppm have been reported.'-'-/ Production of C-14 in
reactor coolants is primarily from the (n,o) reaction yifb 0-17 (which
occurs in natural oxygen with an abundance of 0.037%). ' Although
other possible reactions mentioned are C~13(n,Y), N-15(n,d), and
0-16(n,He-3), the N-14(n,p) reaction in fuel and the 0-17(n»a) reaction
in coolant are believed to be the most significant sources of C-14,' '
Breakdowns of C-14 source terms have been estimated as follows:^ ^
Fuel
N-14(n,p)C-14 -»• 18 Ci/GW-yr (assuming 20 ppm N-14 by weight in fuel)
Q-17(n,a)C-14 -»- 4 Ci/GW-yr (from 0.037% 0-17 in the IIO
BWR
0-17(n,a)C-14 + 8.9 Ci/GW-yr (from oxygen in coolant)
N-14(n,p)C-14 -*• 0.26 Ci/GW-yr (assuming 1 ppm N_ impurity in coolant)
PWR
0-17(n,a)C-14 -»- 3.3 Ci/GW-yr
N-14 (n,p) C-14 -*- 0.09 Ci/GW-yr
Thus, according to these estimates, approximately 22 Ci/GW-yr are
produced from fuel, and would appear as reprocessing plant effluents,
and approximately 3.4 and 9.2 Ci/GW-yr would be released from PWR and
BWR reactor plants, respectively, (average of 6,3 Cl/GW-yr if it is
assumed that reactors are equally distributed between PWRs and BWRs) .
These numbers indicate that although reprocessing plant releases will
account for the majority of C-14 discharged from the fuel cycle, re-
leases from reactor plants are also significant and, if C-14 is deemed
to be worth controlling, these releases will also have to be dealt with,.
Table B-III-1 presents a range of estimates of C-14 production rates. ^
(19)
Actual measurements taken at a BWR showed gaseous releases of
16 Ci/3,000 MWt-yr, which is equivalent to about 16 Ci/GW-yr. These
studies showed the C-14 to appear mainly as CO (95%) , with the remain-
der being CO and hydrocarbons. Similar studies at a PWRw showed
8 Ci/GW-yr, of which 80% were hydrocarbons (CH , C.Hfi) and less than
5% were C02 and CO.
It has been estimated ' that the maximum concentration of CO,
in the dissolver off -gas is approximately 0.025%; C0_ in the off -gas is
approximately 0.03% (CO™ concentration in air). Thus, the maximum ^CO
concentration would almost equal the natural C0n concentration. This
2 fi ft \
estimate varies substantially from other estimates^0' that about 0.02%
B-III-14
-------
TABLE B-III-1
PRODUCTION OF CARBON-14 IN LIGHT-WATER REACTORS
Carbon-14 Production Rate (Ci/GW-yr)
BWR
Fuel
BWR
Coolant
0-17
N-14
Total
0-17
N-14
Total
BWR Sum
(Fuel Plus Coolant)
PWR
Fuel
PWR
Coolant
0-17
N-14
Total
0-17
N-14
Total
PWR Sum
(Fuel Plus Coolant)
ORP/TAD
76-3 (16)
4
18
22
8.9
.26
9.2
31
4
18
22
3.2
.09
3.3
25
Bonka
et.al. (21)
8.4
12.9
21.3
9.9
1.3
11.2
32.5
7.1
12.2
19.3
9.8
1.3
11.1
30.4
Hayes
et.al. (22)*
10.9
21.2
32.1
11.5
—
11.5
43,6
4.0
7.6
11.6
3.3
0.1
3.4
15
ERDA-1535
(.23)
20**
16
36
17**
6
23
Kelly
et al.. (24)
2.7
10.9
13.6
16
29.6
2.7
10.9
13.6
6
19.6
/ o o \
*The production rates presented by Hayes et a1. and Kelly et al.
for 1000 MWt were multiplied by 3.03 (33% thermal efficiency to roughly present
the values on a per GWe—yr basis for comparison purposes.
**Fuel and cladding production rates from ERDA-1535
as a fuel production rate in this table.
(23)
were added and identified
Source: Public Health Considerations of Carbon-14 Discharges
from the Light-Water-Cooled Nuclear Power Reactor
Industry. ORP/TAD-76-3, U.S. Environmental Protection
Agency, July 1976.
B-III-15
-------
of the C0_ flow is expected to be CO,,. The lattet'estimates have ,
been checked and appear to be reasonable: an air flow rate of 1.18 x 10
g/hr is given- which equates to 15.1 m /rain (550 ft^/min) estimated in
ERDA 76-43.^
(1.18 x 106 g/hr) (0.035 ft3/liter) .., .3. , .~ . 3
-TL—rr 7TTT %—TTn—j—7t!—\ = 5J4 It /min or 1X1 m
(1.29 g/lxter) (60 min/hr)
14
ORP/TAD-76-3 gives C0? flow rate of 0.017% of the'CO flow rate
of 592 g/hr.^15) This equates to 20 Ci/GW-yr:
.(0.00017) (592 g/hr) (7200 hr/yr) (1.37 Ci/g 14CO,) _ ,n r./ra_
r- f\ i^erl ^ "~* fcW V»X/ V/W Vj.
50 GW
This is compatible with other estimates within the range of uncertainty,
supporting the estimate in ORP/IAD-76-3.
14
The concentration of C0? in CO- is not important from the stand-
point of suitability of removal technology;'what is important is the
concentration of C0_, and the removal technology selected must be
capable of reducing the 0.033% of C0« in air (0.07% by some estimates
if dissolver additions are considered and a very low air sparge is used)
down to a level which will allow effective removal of C-14.
Estimates have been made that from 95^99% of the C-14 in the
nuclear fuel will be released at the reprocessing plant dissolver as
C0? or CO,with the remainder "likely to be incorporated into carbonic
acad". There do not, however, appear to be any empirical data
supporting this assumption, although it is a reasonable one in view of
the conditions at the dissolver: 90*0,8 M nitric acid, and excess 00
in solution from the U02 and M0^. The 95-99% figure should not be
relied on as definite. The fate of-the 1-5% of C-14 remaining is un-
certain but OKP/TAD-76-3 states that.most likely it will be discharged
with the high-level waste stream. >-Lb-J A third source of uncertainty is
how much C-14 will form CO,, and how much will form CO. This is an
important point, because significant CO will require the installation
of a CO •*• CO- converter—probably' not a- difficult problem, but one that
must be accounted for.
The following discussion of methods of CO. removal from dissolver
off-gas streams is narrowly focused on'the assumption that significant
C-14 will appear as CO™ in the off-gas to make control of this C0_ the
main problem; this assumption is*a reasonable one but ORP/IAD-76-3 warns:
"In summary, the need for research on C-14 in reprocessing
facilities exists in the following areas: amounts of C-14
produced in the fuel, chemical forms of the C-14 as it is
evolved from the dissolver, unexpected chemical behavior
elsewhere in any system, partitioning between the off-gas"
and liquid process systems, possible process pathways and
-------
reactions in those systems , probable decontamination factors
for removal systems, and costs of collection systems."' °^
Three systems have been identified for possible use in removal of
CO, from the dissolver off-gas system: caustic scrubbing, molecular
sieve adsorption, and fluorocarbon absorption.''' The first two have
been used for C0_ control in other applications but not specifically
for dissolver off-gas application; the third method h«s been demonstrated
on a small scale for Kr-85 control from simulated dissolver off-gas (see
Section on Kr-85 control).
B-III.3.2 Caustic Scrubbing
This technique for removal of C0« from an air stream depends on
the reaction of hydroxides such as NauH with acidic gases (CO., NO, N0?)
to produce carbonates and nitrites. The system block diagrams shown in
EKDA 76-43 ' ' provide a two-stage scrub: first a pump utilizing NaOH
as a working fluid and sealant, then a caustic spray tower. By addition
of calcium hydroxide (Ca(OH).) to the stream, CaCCL (calcium carbonate)
is precipitated out. This calcium carbonate may tnen be dried and
packaged.
ERDA 76-43 suggests that annual output of CaCO would be 16-60 kg,
but this appears to be in error. As shown previously, approximately
592 g/hr of CCL is generated, which equates to 3
(592 g/hr) (7200 hr/yr) = 4262 kg/yr of CO-, which translates to:
(4262) (4 3) = 9686 kg/yr CaC03
a much larger quantity than stated in ERDA 76-43, but still within a
range that may be easily managed for ultimate disposal.
Use of caustic scrubbing to remove CO. from air is a common
industrial application. However, the process has never been tested for
dissolver off -gas application, and, therefore, the suggestion '^ that
99% removal rate may be achieved is speculative.
B-III.3.3 Molecular Sieve Adsorption
In this process, the above-described caustic scrub steps for
isolation of the C0~ stream are replaced by adsorption on Type 5A
molecular sieve, fnis is accomplished by -removal of NO and N0_ with
dilute HNO_, destruction of N«0 on a rhodium catalyst, drying on molec-
ular sieve 3A, and C0_ removal on molecular sieve 5A. In order to
collect the carbon, it is proposed that the molecular sieve be regener-
ated (by heating in a vacuum or a gas purge), then the regenerate gas
scrubbed to produce CaCO«.
B-III-17
-------
As with caustic scrubbing, the above-described technology (except
acid-circulating and sealed pumps) has been demonstrated or is con-
ceptually straightforward; the difficulty lies in that no integrated
system for dissolver off-gas application has been demonstrated.
B-III.3.4 Fluorocarbon Absorption
The section on krypton control discusses fluorocarbon absorption
at length and points, out that other gases besides krypton will be
absorbed by R-12, including CO-, and that pilot facility tests showed
C07 decontamination factors of over 10 . The stripper off-gas from the
krypton-oriented process contains Kr, Xe, AT, and CO-. How the C0_
would then be removed from the product gas is not clear, since storage
as a compressed gas is not suitable because of the long half-life of
C-14. Caustic scrubbing could be used to yield calcium carbonate, pre-
sumably at some saving over direct scrubbing of the off-gas, since
smaller volumes of gas are involved after the absorption process.
The fluorocarbon system has been tested conceptually and found to
be effective at CCL removal, although final CO- removal/isolation has
not been tested. tThe fluorocarbon system is explored in more detail
in the discussion of krypton.)
Figure B-III-2 summarizes the three systems.
(8)
B-III.3.5 Conclusions
The requirement for control of C-14 from the nuclear fuel cycle by
EPA appears possible. A large percentage (about 20%) of C-14 production
from the fuel cycle comes from power reactors, and, therefore, meaning-
ful control of C-14 could be accomplished only by control at reactors
as well as at reprocessing plants. Because reprocessing is not being
done at present in the United States, reactor plant off-gas technology
will likely provide a much greater technical base for C-14 control than
presently exists and may provide this base before reprocessing plant
control is required, thereby considerably easing the technical uncer-
tainty for U.S. reprocessors.
Some uncertainties remain in the areas of C-14 production sources,
rates, and chemical form from the reactor and from the' reprocessing
plant. These questions must be answered before an effective control
technology can be developed. Once the sources, rates, and forms of
carbon that must be controlled are more clearly understood, flowsheets
may then be developed on the basis of these quantified factors. Questions
must then be answered regarding acceptability of each proposed method
for the postulated effluent stream composition and percentage content
of C0~. If substantial carbon forms other than C09 are present, either
alternate processes must be provided, or these forms must be converted
to C0£.
B-III-18
-------
RELEASE AS
"*
C0
FIGURE B-lll-2 ALTERNATIVE TECHNOLOGIES FOR THE MANAGEMENT OF C-14
Source; Proceedings of the International Symposium on the
Management of Wastes from the LWR Fuel Cycle.
CONF-76-0701, U.S. Energy Research and Development
Administration, July 1976.
-------
Three technologies have been Identified by EHDA as possible
candidates for commercial application at fuel reprocessing plants for
removal of CO,, from dissolved off-gas streams: caustic scrubbing,
molecular sieve adsorption, and fluorocarbon absorption. None of these
systems has been proved for application to reprocessing plants,
although CO- removal from air by caustic scrubbing is a common commer-
cial process, and insofar as reprocessing plant application is similar
to removal from air, the technical problems will be minimized.
Finally, the possibility should not be overlooked that the most
cost-effective method of C-14 control could be reducing the nitrogen
content of the fuel during the fabrication process. No definitive work
appears to have been done in this area, nor is it expected that any
would be done until the C-14 control question has been better defined.
B-III.4 TRITIUM
B-III.4.1 Background
Current reprocessing plans call for release to the environment of
essentially all of the tritium produced at a fuel reprocessing plant.
There are no absolute limits on tritium releases from the fuel cycle,
but tritium releases must meet effluent concentration limitations set
by the Nuclear Regulatory "Commission in 10 CFR 20. Controls to meet
concentration limitations do not require absolute retention, but rather
dilution.
It is estimated that more than 981 of the tritium produced in ,,,.
LWRs (from ternary fission, with a fission yield of approximately 10~*)
will remain in the fuel to the reprocessing plant. Because of the low
production rate, however, the tritium will be present in very dilute
quantities; concentration in the dissolver is estimated to be about
2 ppm. ^"' Estimates of curie—quantities of tritium vary according to
the fuel design (plutonium fission produces more tritium than uranium^ ...
fission).(25J yalues vary from 415 Ci/MT of fuel<25»2a) to 490 Ci/KT '
to 700 Ci/MT. Under existing fuel reprocessing plant designs, con-
trol of this tritium is impossible except by collection and storage of
large volumes of waste water.
The distribution of tritium in a fuel reprocessing plant is not
well understood, as witnessed by large variations in the estimates in
various references. The following estimates have been given for tritium
pathways.
from
Substantial tritium will remain in the hulls, estimates ranging
10%( ' to 10-50%. (*•' Some tritium will be released in the dis-
solver off-gas as HT, estimates ranging from very little (!%'') to
substantial (10%'"-'). During a zirconium fuel processing campaign at
ICPP, 10-451 of the contained tritium was released from the dissolver
B-III-20
-------
(2.7) (2.B)
as HT. Another estimate Is that'.2Q| will leave at thesiissolvecv
The remaining tritium will remain In the aqueous streams and, unless
controlled, will ultimately be discharged to the environment in liquid
waste streams or as vapor. The uncertainty regarding distribution was
discussed by Musgrave, who concluded:
"Data are extremely sparse; on the chemical state and distri-
bution ratios for tritium in any of these streams. While
the majority of the tritium will eventually appear in
effluents as the oxide or element, its combination with
organic materials will present many additional control
problems if release to the environment is to be completely
avoided."
This organic exchange (which transfers the tritium from the first
extraction cycle contactor to the low-level areas of?the,reprocessing
stream) was estimated to be minimal (less than 1%). * However,
Hall and Ward considered this to be a significant pathway in recycle
schemes:
"...The uranium-plutonium bearing organic stream...also
transfers 0.004 g/1 water and 0.14 molar nitric acid across
the imaginary isolation boundary separating the two process
sides. This stream, an unavoidable cross-over path from the
leacher-feed solution to the low level process side, accounts
for 6 to 20% of the daily tritium input being transferred-.
across the boundary [in the reference recycle scheme]."
Given the above unknown regarding the precis^ behavior of tritium
at fuel reprocessing plants, it is not surprising that considerable
differences exist in proposed means of dealing with tritium. These
tritium-control proposals fall into two broad categories: head-end
processes (voloxidation and pyrochemical techniques) and process-stream
controls (recycle and/or isotopic separation). A third possibility,
retention of the entire water effluent (including, presumably, the water
removed from gas streams) seems practical (economic) only if an on-site
geologic method, such as dumping into deep, stagnant formations, can
be used.
B-III.4.2 Voloxidation
The voloxidation process would provide a front-end kiln to heat
chopped fuel elements to drive off tritium and some other fission
ducts and oxidize U0? to U.,0R. Testing of the voloxidation process
has shown that over 99% of tne tritium will become volatilized (as HTO)
at temperatures ranging from 450-650SC. Conceptual design using the
voloxidation process envisions passing off-gas from the chopper
(tritium would be emitted as HT) through an oxidizer (Ni-Cr-Pd catalyst,
400-60Q°G) to convert HT to HTO, then joining with the voloxidizer
B-III-21
-------
off-gas and cooling to condense the HTO, which may then be removed by
a drier—molecular sieve arrangement. The off-gas stream, which would
contain large quantities of noble gases and iodine, would then be
routed to the off-gas system for treatment. Murbach et al. report
a different flowpath, joining the chopper/voloxidizer off-gas upstream
of the catalytic oxidizer.^*3^
Substantial problems and uncertainties exist with the voloxldation
concept. Testing has shown that U0« -»• U.OR conversion is not always
complete, depending upon residence time, kiln rotational speed, oxygen
content of/the feed gas, and fuel pellet composition and fabrication
technique. Testing has shown consistently good and predictable
results on tritium removal, however, indicating that 99% release may be_ ... ^
readily achieved in the 450°C area. Below 350°C no oxidation occurs. '
One of the most troublesome problems with voloxidation is the
behavior of other fission products. Tests have shown that the process
releases krypton and xenon, and volatilizes ruthenium, cesium, cerium,
antimony, and niobium, as well as significant quantities of uranium and
plutonium. The release of extraneous fission products is not complete.
Goode suggests that about 50% of the krypton/xenon will be released
from the shearing step, and an additional 25% will be released by vol-
oxidation. Iodine removed by voloxidation varied from 40% at 450°C
to 74% at 650°C (higher than. 650°C temperatures reduced iodine release).
Thus, at the likely operating temperature of 450°C, less than one-half
the iodine will be emitted at the voloxidizer. Release of other elements
was investigated by heating fuel rod segments to 450eC for 3 hr, then
750°C for 7.4 hr. The following releases occurred: Nb-95, 9.4%;
Ru-106, 43.2%; Sb-125, 25.8%; Cs-137, 29.1%; Ce-144, 3.3%; U, 2.2%;
Pu, 3.4%.
The release of xenon and krypton is probably not a problem, since
these inert gases will simply pass through the tritium recovery system
and into the off-gas control system. The other elements, however, are
a very significant potential problem, because of their tendency to
plate out at cold spots. Testing showed that most of the elements
released (other than inert gases) plated out either on the burner walls
or burner head, or were trapped in the molecular sieve. Because of the
volatility and subsequent plateout of fission products, there are signi-
ficant questions concerning maintainability and waste handling aspects
of the voloxidation process. Clearly, the system must be absolutely
leak-tight, at least to the molecular sieve.
Another question that arises from the volatillty-plateout question
covers iodine control. Meeting the 99.6% retention requirements of
40 CFR 190 for 1-129 will certainly be complicated if some of the iodine
is left in the voloxidation system.
A final question regards safeguards requirements. Present U.S.
regulations (10 CFR 70) require periodic physical inventories to verify
B-III-22
-------
the plant material balances to within 1% for plutoniura and 0.7% for
U-235. Such requirements may not be met if 2-3% of the uranium and
Plutonium are retained in the voloxidizer. Physical location and assay
of this uranium/plutonium would be virtually impossible without major
disassembly of the highly contaminated voloxidizer system.
Other potential voloxidizer problems have been listed by ERDA 76-43;
1. The UCL •+• UoOo oxidation is exothermic and a difficult control
problem results.
2. Handling the U_00, a fine powder, will present difficulties.
j o
3. Nuclear criticality considerations (although this is listed
as a potential problem, it does not appear to be any more
difficult than criticality considerations in other transfer
systems).
4. Cleaning and maintenance of the voloxidizer system.
5. Potential for thermal failure of the high-temperature kiln.
In addition, there is the hazard of possible ignition of the
Zircaloy cladding and fines by the high-temperature oxidizing gases.
Although extensive data from laboratory-scale and hot-cell tests
demonstrate the fundamentals of the voloxidation process (and identify
some of the potential problems), engineering development has .been
limited to tests with unirradiated U0_ in a rotary calciner.^,, A tri-
tium-control flowsheet has been developed for the AGNS plant along
with a cost estimate of $34 million in 1973 dollars.
At Savannah River a series of laboratory-scale tests are planned
(starting January 1977) of head-end tritium removal technologies ,(26)
which will involve voloxidation of irradiated LWR fuel rod sections.
These tests will include trapping of off-gases in order to determine
the composition of the streams. These experiments should provide
substantial additional data for evaluating the voloxidation process.
Estimates of waste quantities from the voloxidation method are
less than 100 liters/day for a 5 Ml/day plant.
B-III.4.3 Pyrochemical Processing
Pyrochetnical processing consists of first decladding the fuel by
selective melting of the cladding. With stainless steel, this is
feasible (melting point, 1450°C), but with zirconium-based cladding,
the required melting point (1840°C for zirconium) may be too hot. UC
melts at about 21008C. Following the cladding melt, the fuel is
B-III-23
-------
reduced In a solution of reductant alloy (Zn»Ca» Mg) and a salt
(e.g., CaCI™) at 800-9QQ°G. Salt Is recovered and recycled, and the
reductant alloy is distilled off for recycle, leaving U and Pu to be
sent to the acid dlssolver. An alternative process is to precipitate
the U and Pu through addition of calcium nitride.(1)
Fission product tritium and noble gases are both released during
decladding and reduction phases. Iodine is thought to remain in the
salt phase, *• ' but this seems highly speculative in view of the high
temperatures required. Fission product volatility is stated to be a
potential problem.
Most work done on pyrochemical processing has been done at Argonne
National Laboratory, and has been directed toward the LMFBR. According
to ERDA 76-43, "although considerable development work has been carried
out Involving various concepts, a. complete process has not been demon-
strated for LWR of Individual processes that appear to have potential
in a Purex head-end step."(1)
Because of the considerably higher temperatures required In the
pyrochemical process than in the voloxidation and the substantially more
complicated chemical systems, voloxidation will probably prove to be
more favorable than pyrochemical methods.
B-III.4.4 Process-Stream Treatment
Two techniques have been proposed for tritium control once It enters
the aqueous streams at a fuel reprocessing plant; isotopic separation
and water recycle. These two concepts should correctly be discussed
together, since Isotopic separation is normally considered to be supple-
mentary to recycle.
At one end of the recycle-separatlon spectrum is total recycle
without separation. In this case, tritium will build up to a high-con-
centration equilibrium (addition from fuel equal to removal by waste
streams and decay). At the opposite end of the spectrum is no recycling
but separation of tritium from the effluent stream. A middle ground,
and the most favored option, is recycle with bleedstream separation to
keep in-plant concentrations tolerable.
The obvious problem with total recycle without separation is that
tritium concentrations eventually become sufficiently high to present
severe In-plant safety problems. Even a small leak could be quite
serious to operating personnel. This results in the need for costly
gas control systems. On the other hand, the problem with no-recycle
separation of tritium from the effluent stream is that existing processes
for tritium separation from water are uneconomic at high-volume flowrates.
Likewise, storage of the entire aqueous waste stream would be uneconomic
without some degree of tritium concentration to reduce storage volume.
B-III-24
-------
General Electric (GE) has performed an important study of tritium
control by water recycle,(29) ^n which seven cases of recycle process
options were examined and material balances were established for each
case in order to determine the probable tritium distribution and effec-
tiveness of tritium retention. The process systems were, for simplicity,
divided into three segments: the leacher (dissolver), the high-level
system, and the low-level system (determined by total radioactivity,
not just tritium), 'This tritium equilibrium model is shown in Figure
B-III-3. The seven cases are briefly described in Table B-III-2, and
the resultant balances are given in Table B-III-3.
Recycle design considerations demand extensive recycle of water and
acid, removal of water from vent gases, extensive personnel protection
efforts (seals, containment, ventilation control), and avoidance of
contaminated gas contact with hydrophilic substances such as concrete.
GE's reference recycle scheme, Case 1, utilized a modified Purex
process and is arranged to maximize the use of recycle water throughout
the plant and minimize aqueous inputs. The recycle water systems are
segregated into a high-activity system (feed preparation, HA wastes)
and a low-activity system serving the remainder of the process building.
As may be seen in Figure B-III-3, tritium output is from two points
only: leacher off-gas, and tritiated wastes from the LA system. Table
B-III-3 shows that Case 1 results in high LA concentration relative to
HA. This is undesirable and is changed in Case 2 by removal of all of
the tritium from the LA side, which reduces concentrations on both sides
of the process. In Case 3, new acid is added to the LA side only, the
addition resulting in further dilution of the LA side but pushing the
HA side back up. Case 5 combines Cases 2 and 3, but no significant
improvements are realized.
Case 7 turns out to be the most advantageous, for here all aqueous
crossover from HA side to LA side is eliminated, and the only transfer
is through the organic stream from the high-activity column. The tri-
tium concentration on the low side is 0.63 Ci/liter, the lowest of all
seven cases. However, even with the relatively low tritium concentra-
tion, 0.63 Ci/liter, the report estimates that: "this degree of tritium
isolation would probably not be sufficient to significantly reduce the
tritium concentration in the low level process side to the extent that
the personnel protection...could be substantially reduced."^'
Although cost determination was not a goal of the GE study, an
estimate was made that recycle would increase plant cost by 20-25%.
The GE study was selected as a base from which to discuss recycle
because it points out several very important points regarding recycle:
1. Contrary to some studies, it may not be reasonable to expect
near-complete tritium isolation to the high-level side. The
GE study suggested that personnel protection problems would
B-III-25
-------
Of fgas to 1,?,3,6
Tritium ""*'
Removal
Fuel Input-
Aqueous Input
16
Leacher System
23
I H
'"I A
C
0
L
U
M
N
!
24
13,19,20
High Side
Organic Crossover
4,5
Aqueous
Crossover
7,8,9
Aqueous-
Organic
Crossover
10,11,12
Tritiated
Waste
10,14; IS
I
I 10,17,18
Low Side
FIGURE B-lll-3 TRITIUM EQUILIBRIUM MODEL
Source: Hall, N.E, and G.N. Ward. Tritium Control by
Water Recycle in a Nuclear Fuel Reprocessing .
Plant. NEDG-11342, General Electric Company,
San Jose, California, June 1975.
B-III-26
-------
TABLE B-III-2
DESCRIPTION OF POSSIBLE FLOWSHEET VARIATIONS
Case Number Deviation from Reference Flowsheet
1 None
2 No tritiated water removed from leacher; all
water removed from low level side. Same
quantity of water is removed.
3 New acid added to low level side. Equivalent
amount of acid is transferred from the low
level side to the high level side,
4 Recycle uranium transferred from low level
process side to HA feed tank at feed concen-
tration, 1.5 molar U instead of the 2.5
molar U in reference flowsheet.
5 Combination of Case 2 and Case 3.
6 No tritium water removed from leacher; all
water removed from low level side. Minimize
total water removed.
•7 Assumes that tritium is only transferred from
the high level side to the low level side by
the HAP stream.
Source: Hall, N.E. and G.N. Ward. Tritium Control by
Water Recycle in a Nuclear Fuel Reprocessing
Plant. NEDG-11342, General Electric Company,
San Jose, California, June 1975.
B-III-27
-------
TABLE B-III-3
ANALYSIS OF FLOWSHEET VARIATIONS
w
M
M
M
1
M
oo
Variable
Number
0
1
4
8
11
14
17
19
6
13
10
3
2
15
Units
Ci/day
Fraction
Evaporated
Fraction
Carried in HAP
Liters/day
Liters/day
Liters/day
Total liters
Total liters
Ci/liter
Ci/liter
Ci/liter
Liters/day
Ci/day
Ci/day
Case 1
6000
0.065
0.0137
11884
1352
432
8.8xl05
3.5xl05
1.33
1.13
7.89
1950
2592
3408
Case 2
6000
0.000
0.0137
11884
1352
2382
8.8xl05
3.5xl05
0.95
0.75
2.52
0
0
6000
Case 3
6000
0.065
0.0137
20344
23384
432
8. 8x10 5
3.5xl05
2.60
2.41
2.13
1950
5082
918
Case 4
6000
0.065
0.0137
12726
2194
432
8.8xl05
3.5xl05
1.55
1.35
6.87
1950
3032
2968
Case 5
6000
0.000
0.0137
20343
23384
2382
8.8xl05
3.5xl05
3.28
3.08
2.52
0
0
6000
Case 6
6000
0.000
0.0137
11884
1352
432
8.8xl05
3.5xl05
2.19
1.99
13.89
0
0
6000
Case 7
6000
0.065
0.0137
0
1352
432
8.8xl05
3.5xl05
2.90
2.70
0.63
1950
5667
273
Minimum time to tritium
equilibrium (days)
Percent daily tritium
input released
265
6.2
206
721
1.7
396
5.4
874
2.0
549
10.9
776
0.54
Source: Hall, N.E. and G.N. Ward. Tritium Control by Water Recycle in a Nuclear Fuel Reprocessing
Plant. NEDG-11342, General Electric Company, San Jose, California, June 1975.
-------
occur on the LA side even with no water transfer between the
two sides, simply due to carryover from the high-activity
column.
2. A rather obvious conclusion, but one which is important in
understanding the balance between recycle and separation, is
that as tritium levels reach equilibrium in the aqueous streams,
the concentration in waste streams also increases, allowing for
either more economical storage or more economical separation."
3. The conclusion reached by Hall and Ward was that, "based on
the findings of this study, tritium control by water recycle
in a nuclear fuel reprocessing plant is attainable and appears
to merit further study."(29)
(27)
A second study ignores the difficulties identified above
regarding contamination of the low side but takes the recycle concept
a step further and examines alternatives employing concurrent separation.
Isotopic separation of tritium from water streams may be performed
with a number of different technologies (discussed later), all of which
have the common characteristic of being more economic as volume through-
put decreases and as required decontamination factors decrease. There-
fore, in a recycle system the higher the allowable tritium concentration
in the process streams, the more economic will be the removal.
The first system examined by Musgrave envisions total recycle with
periodic removal of the entire Inventory to storage, or a bleedstreara
removal to storage.^ ' This concept has already been discussed and
probably is not optimal because of the significant problem of protecting
in-plant personnel. However, the concept should not be dismissed until
the economics of alternative systems have been compared.
The most potentially attractive scheme presented by Musgrave (who
also presents a once-through full-stream separation concept, which is
probably not economic because of the cost of separation) is operation
using recycle. The level of tritium contamination in the process streams
would be controlled by removing a bleedstream (low flowrate, high con-
centration) for separation. Depleted recycle water would then be
returned to the plant streams (see Figure B-III-4).
The great potential advantage to this concept is the ability to
balance the recycle concentration and separation efficiency to achieve
optimum economics. As the tritium concentration in the recycle water
Increases, the costs of plant maintenance and personnel protection
would be expected to increase (probably as a series of step functions).
Conversely, as tritium concentration increases, the cost of separation
would decrease. It is likely that the optimum cost would lie between
the two extremes of no recycle (zero tritium concentration in recycle)
to no separation (high tritium concentration in recycle).
B-III-29
-------
MAKEUP..
INPUT
EXTRACTION
PRODUCT
WASTE
RECYCLE
SOLIDIFIED
WASTE
BLEED
ISOTOPE
SEPARATION
STORAGE
DEPLETED RECYCLE
FIGURE B-lll-4 ONE PROPOSED TRITIUM SEPARATION SCHEME
Source; Musgrave, B.C. Tritium Distribution in the
Nuclear Industry. ICP-1041, Allied Chemical
Corporation, Idaho Falls, Idaho, January 1974.
B-III-3Q
-------
The above reasoning is consistent with the GE study because as
tritium increases in the high level side, it will spill over into the
low level side, and this will increase the maintenance/personnel costs
accordingly.
The AGNS,plant at Barnwell will release all tritium as vapor out
the stack. *••'"•' Of the total tritium production rate (estimated at
1.8 x 10~2 Ci/sec, or 1555 Ci/day), about 20% will come off at the
dissolver, and the remaining 80% will flow with the acid-water mixture
from the dissolver, to the high-activity waste concentrator, then to
the low-activity waste concentrator, and acid fractionator, issuing in
the overheads from the acid fractionator.v^8) These overheads will be
collected, condensed, sampled, and either recycled or discharged after
dilution to acceptable activity levels, if required. Thus, the AGNS
design provides for acid-water recycle from the fractionator back to
acid makeup, provided tritium contamination is not excessive. In
addition, water recycle occurs at several other points in the plant
design: GPD receiver, service concentrator feed tank, and 1UD/2UD
surge tank.C28, Table 7.1-lt
As already discussed, recycle may cause an increase in tritium
concentration in process flowpaths, and this could also be the case at
AGNS. This concentration would build up in the water recycle with the
acid, and also in tritium exchange with the acid. If the concentration
becomes sufficiently high, one might also expect carryover of tritium
in the solvent, from water-acid carryover, as well as from tritium
exchange with the organic solvent. At AGNS, however, tritium buildup
will be avoided by release out of the stack when concentrations become
excessive. The expected concentration of tritium in the water dis- fja\
charged from the acid fractionator is lower than that,in the GE study
by almost five orders of magnitude (8 Ci/liter vs 10" Ci/liter), a
difference suggesting that although water-acid recycle is planned, very
little (if any) tritium recycle is expected. How much actual acid-water
recycle is feasible with these conditions remains to be seen.
B-III.4.5 Isotopic Separation
Considerable disagreement exists in the published literature regard-
ing the practicality and most promising technology for removal of tritium
from water streams. The tritium separation technologies are generally
based on heavy-water production techniques and include the following;
1. Catalytic exchange
2. Fractional distillation of water
3. Distillation of hydrogen
4. Electrolysis of water
5. Reversible electrolysis
6. Laser enrichment
B-III-31
-------
Catalytic exchange has been used to produce heavy water on a large
scale and thus the method is well understood. The process utilizes
the exchange (in the case of tritium):
HT + H20 * HTO + H?
and occurs when a vaporized mixture is passed over a catalyst. Equili-
brium shifts to the right with lower temperatures.
Because the conventional catalytic exchange method requires vapor-
ization of the water, efforts are being pursued to develop an anti-
wetting hydrophobic catalyst so exchange may occur between liquid water
and gaseous hydrogen in direct contact. The Canadian effort^-*2) ±n
this field has progressed from the laboratory stage to the engineering
stage.
(25)
Electrolysis of water is simple and compact and gives high
separation factors, but is quite expensive because of high electricity ,,..
requirements. A technique of reversible electrolysis is being developed
that could reduce the power requirements by as much as 80%.' ^ Originally,
reversible electrolysis was never developed to commercialization pos-
sibly because it required expensive palladium membranes. Recently,
work in Yugoslavia has produced a technique using a carbon and poly-
ethylene electrode that may be successful for commercial application.
Laser enrichment (by which, theoretically, very high separation
factors are possible) is mentioned in ERDA 76-43 as being under devel-
opment but still at the basic research level.' ' Ribnikar and Pupezin
discuss the first four possible separation techniques and favor cata-
lytic exchange or distillation processes from an economic viewpoint.' '
Of the separation techniques listed, all but laser enrichment and
reversible electrolysis have been applied in the production of heavy
water and are well understood for this application. However, there is
considerable debate regarding their applicability to tritium removal.
This disagreement centers mainly on economic questions and ultimately
comes to the question of whether an economic mix of recycle and separ-
ation may be achieved. There is little question that full-stream (no
recycle) separation is uneconomic:
"The light isotope separation methods, although theoretically
possible, are not practical. For example, over 1 tonne per
hour of water is vaporized in the waste concentrator. The
requirements to electrolyze this quantity of water to obtain
the mixture of hydrogen isotopes for separation is impressive.
Further, the number.of separation stages required would also
be very large..."
"For elimination of releases, complete recycle, isotopic
removal, or a combination of these would be necessary. Iso-
topic removal will be very difficult; there is no proven
B-III-32
-------
system, now operational, to accomplish the separations
required. Existing or proposed processes would be arranged
and cascaded in a manner to accomplish the very high through-
put and large isotopic separation factors required, but it
seems unlikely that an economical process will be developed
to handle this job."^27^
Waste releases from a recycle/separation process would be relatively
small:
"Initial calculations indicate that isotopic distillation of
overhead from evaporated high-activity waste (HAW) would
concentrate 95% of the tritium from processing one ton of
uranium for storage in 5 to 6 gallons.,."(26)
B-III.4.6 Waste Disposal
Since the above processes have tritiated water as their end-product,
disposal of this waste water would not be difficult conceptually. Solidi-
fication as cement appears to be the simplest and cheapest, although
even cheaper methods would perhaps be found in disposal in stagnant
aquifers.
Although EKDA 76-43 mentions a wide variety of possible techniques
for long-term tritium disposal, the existence of at least one cheap,
simple method obviates the need for detailed discussion of more exotic
methods.
B-III.4,7 Conclusions
Tritium control technology is still at the laboratory/hot-cell
test stage and will not be commercially developed for many years.
Sufficient proof-of-principle has occurred to give some hope that tri-
tium control is at least possible, although the economics of this tech-
nology are highly uncertain. There is great need for detailed economic
optimization studies to single out the one or two most promising systems,
followed by detailed cost studies to demonstrate whether tritium control
is economically feasible. Particularly interesting would be a study of
recycle/separation optimization to demonstrate whether such a scheme -
which is conceptually appealing - would in fact be economic and if so,
at what recycle and separation rates.
B-III.5 RUTHENIUM
The following discussion of ruthenium control is abstracted from
ERDA 76-43.(1)
B-III-33
-------
During high-temperature solidification of waste from a fuel
reprocessing facility, ruthenium may volatilize and require removal
from the off-gas system. According to ERDA 76-43 the reprocessing of
1 MTU of fuel irradiated to 25,000 MWD/MTU would feed 2.5 x 105 Cl of
Ru-106 into the solidification facility (based on one-year cooled waste)
However, the quantities of ruthenium passing Into process off-gases
during waste solidification by the different types of processes are not
well understood. The Final Safety Analysis Report for the AGNS Separ-
ations Facility lists the expected stack release for total Ru-106 as
4.6 milllcurles per day (mCi/day); this release is expected to meet all
standards and specifications promulgated by state and Federal agencies.
The only data available on the release of volatile ruthenium during
calcination are from laboratory measurements on simulated LWR wastes
during fluid bed calcination. The mode of heating affects the quantity
of gaseous ruthenium generated and passed Into the off-gas system.
Studies at INEL, using a simulated LWR commercial waste, show that of
the ruthenium fed into a fluidized bed calciner with indirect heating,
the amount leaving in the gaseous state varies from 6 x 10~ % at 570°C
to 3 x 10~2% at 500°C, and 100% at temperatures below 375°C. When in-
bed combustion heating is used, the same volatility, 2 x 10 %, Is
observed at 500°C and at 550°C.
Equipment components that can be considered for removal of both
particulate and gaseous ruthenium include cyclones, quench tanks, ven-
turi scrubbers, adsorbers and HEPA filters. Of these, the most effec-
tive in removing volatilized ruthenium appears to be the adsorber. Of
a number of adsorber materials studied, silica gel (Davidson Grade 40)
was found to be the most suitable. It has a high DF for volatilized
ruthenium (1000), has a high capacity (exceeds 5 g Ru/liter), adsorbs
at a moderate temperature, is not-attacked by the off-gas, and can
easily be regenerated.
The treatment of off-gases that contain volatilized ruthenium from
other waste solidification processes is essentially the same as for the
fluidized-bed calcination process, but few significant data are avail-
able on ruthenium behavior during solidification of LWR wastes by pro-
cesses other than fluidized-bed calcination. Any differences will be
directly related to the quantity of ruthenium volatilized. Some pro-
cesses may release large quantities of ruthenium, which require more
elaborate off-gas cleanup. Preliminary laboratory data, for example,
indicate that rotary-kiln calcination might volatilize ruthenium in
quantities that would require DF and adsorber capacity requirements of
500 to 5000 times those needed for fluidized-bed calcination.' ' For
such a situation, a method must also be provided to prevent ruthenium
plateout on pipe and component walls, which could lead to plugging. This
might be done by maintaining the wall temperature between 150°C and 300°C.
An alternative to removing ruthenium from the off-gas is to use
additives for holding ruthenium in the solid form. For high pH systems
at high temperatures, for example, formation of aluminosilicate in the
melt immobilized the ruthenium.
B-III-34
-------
APPENDIX B-III
REFERENCES
1. Alternatives for Managing Wastes from Reactors and Post-Fission
Operations in the LWR Fuel Cycle. ERDA 76-43, U.S. Energy Research
and Development Administration, May 1976.
2. Allied General Nuclear Services, Barnwell Nuclear Fuel Plant
Safety Analysis Report, Docket 50-332 (Rev. 3/21/64).
2.a. Ibid (Rev. 3/21/69).
3. U.S. Code of Federal Regulations, Title 40, Part 190.
4. Bendixsen, C.L. and F.O. German. 1974 Operation of the ICPP Rare
Gas Recovery Facility. ICP-1057, Idaho Chemical Programs, Allied
Chemical Corp., March 1975.
5. Stephenson, M.J. and R.J. Eby. Development of the 'FASTER1 Process
for Removing Krypton-85, Carbon-14, and Other Contaminants from
the Off-Gas of Fuel Reprocessing Plants. Presented at 14th ERBA
Air Cleaning Conference, August 1976.
6. Murbach, E.W., W.H. Carr and S.H. Gray. Fission Product Gas
Retention Process and Equipment Design Study, ORNL-TM-4560, Oak
Ridge National Laboratory, Oak Ridge, Tennessee, May 1974.
7. Foster, B.A. and D.T. Pence. An Evaluation of High Pressure Steel
Cylinders for Fission Product Noble Gas Storage. ICP-1044, Idaho
Chemical Programs, Allied Chemical Corp. , 1974.
8. Proceedings of the International Symposium on the Management of
Wastes from the LWR Fuel Cycle. CONF-76-0701, U.S. Energy Research
and Development Administration, July 1976.
9. Foster, B.A., D.f. Pence and B.A. Staples. Long Term Storage Tech-
niques for Kr. In: Proceedings of the Thirteenth AEC Air Cleaning
Conference, San Francisco, Ca., 1974. Vol. 1, U.S. Energy Research
and Development Administration, CONF-740807, March 1975.
B-III-35
-------
APPENDIX B-III (continued)
REFERENCES
10. Pence, D.T. and B.A. Staples. Solid Adsorbents for Collection and
Storage of Iodine-129 from Reprocessing Plants. In: Proceedings of
the Thirteenth AEC Air Cleaning Conference, San Francisco, Ca.
CONF-740807 U.S. Atomic Energy Commission, August ,1974.
f
11. Thomas, T.R. Separation and Retention of Iodine. In: Proceedings
of the International Symposium om the Management of Wastes from the
LWR Fuel Cycle, Denver, July 1976.
12. Thompson, C.H,, J.A. Kelley. Evaluation of Methods for Retention
of Radioiodine During Processing of Irradiated 237 Np. DP-1373,
E.I. duPont deNemours & Co., Savannah River Laboratory, S.C., June 1975,
13. Groenier, W.S. and B.A. Hannaford. An Engineering Evaluation of the
lodox Process; Part 2: Removal of Iodine from Air Using a Nitric
Acid Scrub in a Bubble-Cap Column; Part 3: Correlation of Mass
Transfer Data; Part 4: Process Sensitivity to Impurities. ORNL-
RM-4701, Oak Ridge National Laboratory, Oak Ridge, TN, May 1975.
14. Mailen, J.C. and D.E. Homer. Removal of Radioiodine from Gas^Streams
by Electrolytic Scrubbing. Nuclear Technology, September 1976.
15. Magno, P.J,, C.B. Nelson, and W.H. Ellett. A Consideration of the
Significance of Carbon-14 Discharges from the Nuclear Power Industry.
In: Proceedings of the Thirteenth AEC Air Cleaning Conference, San
Francisco, Ca., CONF-740807, March 1975.
16. Public Health Considerations of Carbon-14 Discharges from the Light--
Water-Cooled Nuclear Power Reactor Industry. Technical Note ORP/TAD-
76-3, U.S. Environmental Protection Agency, July 1976.
17. Renfrew, C., Carbon-14 and the Prehistory of Europe. Scientific
American 225(4): 63-72, October 1971.
18. Yeates, D.B., A.S. Goldin, and D.W. Moeller. Natural Radiation in
the Urban Environment. Nuclear Safety 13(4): 275-286, July and
August 1972.
14
19. Kunz, C.O. C Gaseous Effluents'from BWRs, American Nuclear
Society Annual Meeting, June 8, 1975.
B-III-36
-------
APPENDIX B-III (continued)
REFERENCES
14
20. Kunz, C.O. C Gaseous Effluents from PWRs, Health Physics
Soc., Knoxville, Tennessee, October 12, 1975.
21. Bonka, H., K. Brifssermann, G. Sch-^arz, Umweltbelastung durch
Radiokohlenstoff aus Kerntechnisehen Anlagen. Reaktortagung,
Berlin, April 1974,
22. Hayes, D.W. and K.W. MacMurdo. Carbon-14 Production by the
Nuclear Industry Savannah River Laboratory, Aiken, S.C.
DP-MS-74-33, to be published.
23. Final Environmental Statement—Liquid Metal Fast Breeder Reactor
Program. ERDA-1535, U.S. Energy Research and Development
Administration, December 1975.
24. Kelly, G.N. et al. The Predicted Radiation Exposure of the
Population of the European Community Resulting from Discharge of
Krypton-85, Tritium, Carbon-14, and Iodine-129 from the Nuclear
Power Industry to the Year 2000. Commission of the European
Communities, Luxembourg, Doc.V/2676/75, September 1975
25. Ribnikar, S.V. and J.D. Pupezin. Possibilities of Tritium Removal
from Waste Waters of Pressurized Water Reactors and Fuel Reprocess-
ing Plants. In: Proceedings of the Thirteenth AEC Air Cleaning Con-
ference, CONF-740807, San Francisco, California, August 1974.
26. Savannah River Laboratory Quarterly Report. Light Water Reactor
Fuel Recycle. DPST-LWR-76-1-2, E.I. Dupont de Nemours & Co.,
April-June 1976.
27. Musgrave, B.C. Tritium Distribution in the Nuclear Industry.
ICP-1041, Allied Chemical Corporation, Idaho Falls, Idaho,
January 1974.
IS. Barnwell Nuclear Fuel Plant Separations Facility Final Safety
Analysis Report. Allied General Nuclear Services, 1974.
29. Hall, N.E. and G.N. Ward. Tritium,Control by Water Recycle in a
Nuclear Fuel Reprocessing Plant. NEDG-11342, General Electric
Company, San Jose, California, June 1975.
B-III-37
-------
APPENDIX B- III (continued)
REFERENCES
30, Goode, J,H. Voloxidation, Removal of Volatile Fission Products
from Spent LMFBR Fuels. ORNL-TM-3723, Oak Ridge National Labora-
tory, Oak Ridge, Tennessee, 1973.
31. Unger, W.E. LMFBR Fuel Cycle Studies Progress Report for November
1970. ORNL-TM-3250 No. 21 Oak Ridge National Laboratory, Oak Ridge,
Tennessee, 1970.
32. Stevens, W.H. and F.P. Blackstein. Heavy Water Production Using
Hydrogen-Water Exchange. AECL Research and Development in Eng.,
Winter 1973, Atomic Energy of Canada, Ltd., AECL-4606, Winter 1973.
B-III-38
-------
APPENDIXB-iy
DETAILS OF REFERENCE DISPOSAL-SITE FACILITIES
This appendix describes details of the reference disposal site dis-
cussed in Section 3.5.1 of this report.
B-IV.l SURFACE FACILITIES
B-IV.1.1 Canned Was te Receiving Building
This building must receive approximately 350,000 canisters of high-
level waste (HLW), cladding hulls, and intermediate-level TRU waste over
the life of the facility, and must be sized accordingly. For a ten-year
operating life this corresponds to a throughput of approximately 100
canisters (8-10 rail transportation casks) per day for a seven-day-per-
week operation. The canister receiving building must provide cask-handling
capability — i.e., for rail car cask unloading, cask decontamination as
necessary, cask cool-down, and cask repair. An optimized HLW cask, which
is shipped in a vertical orientation and is capable of being unloaded
without removal of the cask from the rail car, will reduce cask-handling
time. It is anticipated that average rail cask turnaround time will be
on the order of 12-24 hours. Alternate forms of packaging for intermediate-
level TRU, such as shielded bulk shipment, will greatly reduce the service
demand on the canister-receiving facility.
B-IV.1.1.1 Surface Canister-Handling Cells
Canister-handling cells must be designed to handle the canister
throughput discussed above. The canister—handling cells must have the
capability to remove, temporarily store, and replace the shipping cask head;
unload canisters from the shipping cask; inspect individual canisters for
damage; repair or package damaged canisters; and transfer acceptable
canisters either to the canned waste shaft hoist cell for subsequent
transfer to the mine or to the interim surface-storage facility. All
canister handling must be done remotely because of the high radiation
levels of the canisters. The in-cell equipment must be specifically
designed for remote maintenance, or the various cells must be segregated
in order that a cell with faulty equipment may be isolated, and decontam-
inated if necessary, to give personnel access for hands-on maintenance.
Because of the sealed design of the canisters, relatively low contamination
and radiation levels are anticipated in the cells when waste canisters are
not being handled.
B-IV-1
-------
B-IV. 1.1,2 Interim Waste Cani'stef'Storage-
An interim canister-storage facility should be constructed to accept
an overflow of canisters if there are problems in the canister hoist
system or in the mine. Canisters would be separated according to waste
type and placed in dry storage vaults. Cooling would be provided either
by a network of cooling coils distributed throughout the storage vault
or by forced air circulation around the canisters. Final heat removal
from the cooling system would be via a heat exchanger for either option.
An interim storage capability of 3-6 months of normal canister receipt
operations has been suggested.
B-I?.1.1.3 Canister Hoist Cell/Hoist Equipment
Banisters are transferred from the canister-handling cells to the
canister-hoist cell and loaded one at a time into the canister hoist cage
for transfer to the mine. High-speed hoisting equipment will be located
outside the cell to permit hands-on maintenance.
B-IV.1.2 Low-Level TRUReceiving Buildjlng
"3
This facility must be designed to receive approximately 140,000 m
over the life of the facility. The receiving building should be designed
to permit receipt of low-level TRU waste either by rail or truck. Most
shipments will probably be made by truck with an average shipment con-
taining approximately the equivalent of 96 208-liter (55-gallon) drums
or 20 m3 of waste. For a facility that will be filled in ten years, the
average throughput will be 2.5-3 truckloads per day for a five-day-per-
week operation, and truck unloading time should average approximately two
hours. Trucks will be unloaded using conventional freight-handling equip-
ment, such as fork lifts or overhead cranes. If cladding hulls and
intermediate-level TRU wastes are shipped in bulk package form instead
of in canisters, these containers will be received in the low-level TRU
facility, with the design adjusted to handle the increased throughput.
The capacity of the low-level TRU interim storage building must also be
increased accordingly.
B-I¥.1.2.1 Low-Level TRU Interim Storage Building
This building will be used to store temporarily any backlog of
received waste that might result from problems with the hoist equipment
or in the low-level TRU portion of the mine. This building will resemble
an ordinary warehouse and should have a capacity of 3-6 months' accumula-
tion of low-level TRU.
B-IV-2
-------
B-IV-1.2.2 Low-Level TRU Hoist Facility
This facility will contain nigh-speed hoisting equipment and a
transfer cage loading area.
B-IV.1.3 Man and Materials (M&M) Hoist Facility
This facility will house the high-speed M&M hoisting equipment. It
will also serve as a staging area for personnel and materials transfer
into the mine, the fresh air inlet for the mine ventilation system, and
the transfer point for the mined materials storage and reclamation system.
B-IV. 1.4 Mine Vent ilation Sy s tem
The mine ventilation system will provide a continuous, once-through
source of pre-filtered fresh air into the mine through the ventilation *
compartment of the M&M shaft. It will also distribute the air throughout
the mine, as needed, from least-contaminated to potentially most-contaminated
areas, and then exhaust the air through the ventilation exhaust shaft into
the radwaste treatment system. The ventilation system must be designed
to treat all ventilation exhaust in the event of an abnormal occurrence
in order to ensure the safety of the general public, A detailed dis-
cussion of a typical mine ventilation system^suitable for use in the ,-.
reference facility has been developed for the Office of Waste Isolation.
The facility ventilation system can also be used to transport waste heat
out of the mine storage facilities during any period for which retriev-
ability of wastes is desired.
V-IV. 1.5 Site Radwaste Process_^aci_lit:y
The facility will process, package, and dispose of all radioactive
wastes generated at the waste burial site. All potentially contaminated
exhaust gases from the various on-site waste-handling cells, the chemical
laboratory, and the mine ventilation system will be passed through a
series of filters in the gaseous radwaste treatment system prior to
release from the plant stack. The radioactivity released from the stack
will be constantly monitored to ensure that it meets appropriate regula-
tory guidelines. Liquid, radwaste, mostly in the form of laboratory wastes,
decontamination solvents, and laundry wastes will be processed into a
solid waste form for burial. Solid wastes will include contaminated tools
and equipment, miscellaneous health physics and laboratory supplies,
solidified liquid radwaste, spent filters from the gaseous radwaste treat-
ment system,radiation-protective clothing that can no longer be used, etc.
Solid radwaste will be suitably packaged for ultimate burial in the LLW
portion of the mine facility.
B-IV-3
-------
B-IV.1.6 MinedStorage and Reclamation System
For the non-retrievable reference facilities in this study, a minimum
of 0.9 x 106 m3 (1.2 x 106 yd3) of salt for an HLW facility and 1.8 x 106
m3 (2.3 x 10" yd3) of salt for a throwaway-fuel-cycle facility must be
removed from the mine during construction, stored for the duration of the
facility operating period, and then returned to the mine as backfill
during facility decommissioning. For a fully-retrievable facility, such
as the one discussed at the end of this section, the corresponding volumes
are 14 x 106 m3 (18 x 106 yd3) for the HLW and 26 x 106 m3 (34 x 106 yd3)
for spent fuel. The mined materials storage and reclamation system will
consist of: 1) a rail loop transportation system connecting the M&M
shaft with the designated mined materials storage area; 2) hopper-type
rail cars to transport the mined salt; and 3) salt-handling equipment for
rapid loading of hopper cars at the M&M shaft and piling of the unloaded
salt at the designated storage location. The stored salt should be pro-
tected from the weather during the storage period to prevent salt contam-
ination of the environment.
B-IV.1.7 Site Support
These facilities include an administration building, machine shop,
warehouse, personnel staging area — including lunch room, locker room,
and showers —, health physics and chemical laboratory facilities, and
a site security system including perimeter fences, gate houses, etc.
B-IV.1.8 Shaft Data
Four concrete-lined shafts will be required for the reference facil
ities assumed in this study. The following shaft dimensions have been
suggested;(1)
Man and Materials Shaft — 7.9 m (26 ft) i.d.
Canister Waste Shaft — 1.8 m (6 ft) i.d,
Low-Level TRU Shaft — 3.0 m (10 ft) i.d.
Ventilation Exhaust Shaft — 6.1 m (20 ft) i.d.
A typical shaft-digging sequence is detailed, and suitable operating
criteria for hoisting equipment are outlined.
B-IV.2 MINE COMPONENTS
The design developed for the Office of Waste Isolation is for retriev-
able storage.^' To increase mine stability while minimizing overall mine
volume, the TRU fields have been located 21 m (70 ft) above the HLW burial-
field .
B'IV'4
-------
B-IV.2.1 Receiyjjig Stations .••..•; . .'YA'..Y
B-IV.2.1.1 M&M Receiving Station
This station is located at the bottom of the M&M shaft. All per-
sonnel, materials, and equipment entering and leaving the mine do so
through this station. This station serves the following functions:
* Provides a salt transfer point to the surface during mine
excavation; .
• Serves as mine operations control, including a personnel staging
area, offices, toilets, communications system, etc.;
* Controls ventilation by means of intake air distribution louvers;
• Handles all materials entering the mine, including many large
and heavy objects, such as disassembled parts of the canred
waste placement vehicles. This station must have the handling
equipment and facilities to receive and assemble such components;
* Connects, through the elevator system, the main (HLW) level of
the mine with the TRU storage areas 21 m (70 ft) above the main
mine level.
B-IV.2.1.2 High Level Waste Receiving Station/Transfer Cell
This facility is a remote-handling cell at the base of the canned
waste hoist shaft. Equipment in the cell must be capable of receiving
canned waste from the surface, temporarily storing canisters (10-15 max-
imum) prior to burial, and transferring canisters to the waste placement
vehicle. Waste canisters will be loaded into the waste placement vehicle
through ports in the top of the transfer cell. The number of loading
ports is determined by the required waste placement rate. Each waste
placement vehicle will be capable of handling an average of three canisters
per normal eight-hour shift, and one port can efficiently service three
vehicles.
B-IV.2.1.3 Intermediate-Level TRU Waste Receiving Station/Transfer Cell
This remote canister-handling cell is located at the intermediate-
level TRU storage level of the mine, approximately 21 m (70 ft) above the
HLW receiving station. The canned waste shaft passes through the TRU
transfer cell; canned intermediate-level TRU is removed with remotely
operated equipment at the intermediate-level TRU receiving station, while
canned HLW continues down the waste shaft to the HLW receiving facility.
(Facilities with separate HLW and TRU shafts are also being considered.)
The TRU transfer cell must provide the same functions for intermediate-
level TRU canisters as the HLW transfer cell does for HLW.
B-IV-5
-------
B-IV.2.1.4 Low-Level TRU Waste Receiving Station
This is a modest transfer area located at the base of the low-level
TRU shaft. It must be large enough to .permit efficient handling of con-
tainers with a fork lift vehicle.
B-IV.2.2 Mine Level Machine; Shop/Equipinent_Service Area
An area is set aside in the mine to service the waste placement
equipment. This area should also include a machine shop in order that
routine repairs may be made in the mine rather than requiring the time-
consuming transfer of equipment to a surface facility. The service area
should be connected by a series of corridors to the HLW and TRU, storage
areas in the same service facility in order-to make it possible to service
equipment from these areas.
B-IV.2.3 HLW Storage Area
The storage area for the reference HLW facility is made up of 1112
burial tunnels. Adjacent tunnels are parallel to each other with 23.8-m
(78-ft) center-to-center spacing, and are connected by a series of main
and branch access corridors. Each burial tunnel is essentially identical
in design, measuring approximately 171 m (560 ft) long x 5.5 m (18 ft)
wide x 5.5 m (18 ft) high. The height and width dimensions are set by
spatial requirements of handling equipment and may have to be modified
slightly when a final design is evolved for a canned waste placement
vehicle. Thirty-two canister holes are bored in the tunnel floor on 5,3-m
(17.5-ft) center-to-center spacing. Each tunnel has only one entrance,
and each exhausts to the ventilation treatment system from the dead end
of the tunnel. This arrangement makes it possible always to circulate
air from a less-contaminated to a potentially more-contaminated area of
the mine. Canisters are transported to the burial tunnel with the canister
placement vehicle discussed later and are placed In canister holes, working
from the dead end of the tunnel toward the open end. After the canister
has been lowered into the burial hole, the hole is backfilled with crushed
salt. When all canister holes in the burial tunnel are full, the tunnel
is backfilled as soon as convenient with salt removed from a burial tunnel
under construction.
Two rows of burial tunnels are located back-to-back (dead end to dead
end) in order to minimize the number of branch connecting tunnels and the
amount of excavation required for the network of ventilation exhaust
tunnels around the HLW storage area.
Main and branch corridors lead from the HLW receiving-transfer cell
to the HLW burial field. Approximate dimensions are 9.1 m (30 ft) wide
and 5.5 m (18 ft) high, with actual dimensions set by the requirements of
the handling equipment.
B-IV-6
-------
B-IV.2.4 Canister Holes .,,..,:
Each burial tunnel will have 32 canister holes bored along the cen-
terline. HLW canister .holes will be 0.46 m (18 in) in diameter x 6*1 m
(20 ft) long. After a canister is in place, the remainder of the hole
will be backfilled with crushed salt to provide radiation shielding. The
crushed salt will be put in with the canister transport still in place
over the canister hole to provide shielding for the operators. Once the
hole is backfilled, the dose rate at the floor of the burial tunnel will
be less than 10 mR/hr. Spent-fuel canister holes will contain up to
four canisters per hole. Each hole will be 0.46 m (18 in) in diameter
and 23.2 m (76 ft) deep. Prior to loading, each canister hole will be
fitted with a shielded isolation valve discussed in the following section.
The loading sequence will be as follows:
(1) Position the canister transport vehicle over the shielded
isolation valve.
(2) Open the isolation valve and lower waste canister No. 1 into the
empty canister hole and release canister.
(3) Backfill hole with a premeasured volume of crushed salt to
approximately 0.6 m (2 ft) above the in-place canister. This
backfill of salt will provide a floor on which the next canister
can rest.
(4) Close isolation valve and return to the waste receiving/transfer
cell for canister No. 2.
(5) Repeat steps 1 through 4 for canisters Nos. 2, 3, and 4.
(6) After placement of canister No. 4, backfill to the top of the
canister hole.
The multiple loading of four spent-fuel canisters in one hole pro-
duces less heat per location than a single HLW canister. The alternative
to multiple loading of spent-fuel canisters is boring approximately four
(actually 3.6) times as many holes in each standard burial tunnel. In-
creasing the number of canister rows in the tunnel would require widening
the tunnel, thereby increasing overall facility costs.
B-IV.2.5 Canned Waste Placement Vehicle
Canned wastes, including HLW, spent fuel, cladding hulls, and inter-
mediate-level TKU are highly radioactive and must be transported in
specially designed, heavily-shielded vehicles. Each vehicle must provide
the following functions:
B--IV-7
-------
* Provide a fine positioning capability for the transport vehicle
in centering the transport vehicle cask opening over the transfer
cell port and the canister burial holes.
* Provide shielding to operating personnel during the canister
loading/unloading sequences through the use of a movable shield.
Shielding must be effective against both neutron and gamma radia-
tion.
* Provide internal, remotely-connected capability for moving cani-
sters into the transport vehicle from the transfer cell and for
lowering canisters into the canister burial holes. For spent
fuel and TRU the hoisting/lowering capability may be as much as
24 m (80 ft).
* Provide motive power for waste transfer. The most probable power
alternatives are diesel and electric. Round-trip distances per
canister may be as great as 4 miles for HLW and 8 miles for spent
fuel.
* Provide shielding and, if necessary, cooling capability during
transport. Shielding must be effective against bdth neutron and
gamma radiation.
• Provide capability for remote viewing of the canned waste handling
activity through the use of a radiation-tolerant television camera
system.
• Provide crushed salt backfill capability.
* Provide constant, portable radiation detection capability.
The above functions may be provided by one or several pieces of
equipment. The transfer vehicle will probably resemble a spent-fuel
shipping cask in a vertical position mounted on a motorized bed set on
either rubber tires or tracks. An internal hoist located in the top of
the cask portion of the vehicle would raise and lower the canister. A
thick shielded gate valve would be located at the bottom of the cask to,
provide shielding from radiation coming through the opening in the bottom
of the cask during canister transport. The transport vehicle would be
5 m (16 ft) high and weigh 30-45 MT, for an HLW facility. The length and
width dimensions will depend on the ultimate equipment design, mode of
propulsion, etc. but are expected to be no more than 4.6 m (15 ft) long
by 3.0 m (10 ft) wide. The transport vehicle for spent fuel will be 6 m
(21 ft) high and weigh 35-55 MT. The base dimensions will be similar to
the HLW machine. The interface between the transfer cell or the canister
hole and the transport vehicle can be accomplished either by: 1) using
a separate shielded isolation valve; 2) lowering a shielded collar sus-
pended from the bottom of the transport vehicle; or 3) lowering the cask
itself until it touches the floor. An isolation valve of some sort will
be necessary to close the port over the transfer cell when the transporter
B-IV-8
-------
is not In place for loading. Such a valve will also be necessary if mul-
tiple canisters are to be loaded into spent-fuel and TRU canister holes.
Furthermore, an isolation valve must be developed to facilitate canister-
hole shield-plug installation if retrievable storage is to be used. It
is probable that a combination of an isolation valve and a shield-collar
or lowered-cask technique will be used to provide personnel shielding
during canister transfer.
Canister transport vehicle work has been done for Project Salt
Vault.(2) Moreover, the functional requirements for canister handling
are very similar to those encountered in High-Temperature Gas-cooled
Reactor (HTGR) refueling experience.
B-IV.2.6 Intermediate-Level TRU Wast.e Storage Field
The intermediate-level TRU facility layout in Reference 1 has been
assumed for this study. Canned cladding hulls and TRU are received at
the intermediate-level TRU receiving station/transfer cell, loaded into
canned waste transporter vehicles, transferred to the TRU canister storage
field, and lowered into canister holes. The procedure is identical to
that for HLW disposal. Burial tunnels are 171 m (560 ft) long x 11.6 m
(38 ft) wide x 5.5 m (18 ft) high, and contain six parallel rows of cani-
ster holes. Canisters are located on a square array with 1.2-m (4-ft)
center-to-center spacing. Each burial tunnel can store 790 canisters;
398 rooms are required to accommodate the 313,500 canisters to be buried
in the intermediate-level TRU field. There are salt pillars 12.2 m (40
ft) thick between adjacent rooms, aligned in parallel on 23.8-m (78-ft)
center-to-center spacing. When all canister holes in a room are filled,
the burial tunnel is backfilled with salt being removed from the con-
struction of new tunnels.
The reference arrangement described above may not be the best con-
ceivable design. Because TRU waste produces relatively low levels of
decay heat, multiple canister loading per hole or stacked warehouse-type
storage of shielded bulk shipping containers are both possible. Both
approaches should be investigated before deciding on a final TRU canister
storage field design.
B-IV.2.7 Lpw-Level_TRU Waste_ Burial Field
Low-level TRU gives off low levels of both heat and radiation and
can therefore be handled with conventional freight-handling equipment.
TRU drums are picked up at the low-level TRU receiving station with a
diesel or electric powered fork lift vehicle and transported to burial
tunnels where units are stacked, warehouse-style. Each burial tunnel
measures 171 m (560 ft) long x 11.6 m (38 ft) wide x 5.5 m (18 ft) high.
When a burial tunnel is filled to capacity, it is backfilled with salt
from tunnels under construction in another part of the mine.
B-IV-9
-------
The above array may not represent the best storage for TRU drums.
The size of the low-level TRU storage field can be decreased by placing
some (possibly all) of this waste in HLW burial tunnels before they are
backfilled with salt.
REFERENCES
1. Parsons, Brinckerhoff, Quade & Douglas, Inc. Waste Isolation Facility
Description - Bedded Salt. Document Y/OWI/SUB-76/16506, September
1976.
2. Bradshaw, R. .and W. McClain, Project Salt Vault: A Demonstration of
the Disposal of High-Activity Solidified Wastes and Underground Salt
Mines. ORNL-4555, Oak Ridge National Laboratory, April 1971.
B-IV-10
-------
APPENDIX B-V
CORROSION OF METAL CONTAINERS
B-V.l TEMPORARY STORAGE OF CONTAINERS IN WATER
If the container Is temporarily stored in water prior to burial, it
must resist corrosion. The water will be deionized; its chemistry will
have to be monitored and it will have to be kept free of chlorides that
could enhance corrosion of some container materials, such as stainless
steel. There is also the remote possibility that corrosion could be
enhanced by the products formed by radiolysis of water (oxygen and perox-
ides) , A report from Sweden mentions the radiolysis effect on corro-
sion. ^
The container corrosion problem has been discussed with personnel
at Battelle Northwest Laboratories, which is planning to make full-sized
containers of various metals, including carbon steel, stainless steel,
and Inconel, and expose them to water. Chlorides will be introduced as
incursions to the water in a worst-case situation. Possible harmful
effects resulting from sensitization of the stainless steel or from
radiolysis of the water will be found from these tests. Radiolysis prod-
ucts of water are not believed to constitute a problem, except if there
are crevices in the metal.
B-V.2 LONG-TERM STORAGE IN SALT BRINE ENVIRONMENT
Data have been compiled on corrosion rates of alloys that may be
used as container materials and are given in Tables B-V-1 through B-V-5.
If the container is exposed to hot brine in a salt deposit, the alloys
having high resistance to general corrosion or pitting corrosion are
titanium-nickel (Ti Code-12) or titanium-palladium alloys. If the con-
tainer is subjected to hot salt above the boiling point of water in an
atmosphere containing water and oxygen, the alloy may be subjected to
pitting attack. This attack may occur with long-time exposure at temper-
atures above several hundred degrees Celsius. The mechanism of attack
has been described in a recent paper.'- ' Hot salt pitting has been dis-
cussed with several experts on corrosion at Titanium Fabrication Corpor-
ation (titanium alloys), and at the International Nickel Company (stain-
less steels and nickel-based alloys). These discussions indicated that
titanium alloys may not be resistant to pitting in hot salt, but there
is no information available on pitting times. Inconel 625 or Hastelloy
C would be the alloys most resistant to pitting in hot salt, but infor-
mation on hot salt pitting of these alloys below 30Q°C is also not avail-
able. Given the lack of information, It is recommended that tests be
carried out to determine the pitting resistance of selected alloys, such
as Inconel 625 or Hastelloy C, in hot salt.
B-V-1
-------
In Tables B-V-1 through B-V-5 some corrosion fates are given as 0.
This simply means that the corrosion was below the detectable limit for
the experiment in question. To gain an idea of the limit of deteetability,
it is suggested that in each case the reader look at the reference number
in the last column of each table, and scan the tables for other corrosion
rates reported by the same reference. If no such data are found,, reference
must be made to the original paper.
B-V-2
-------
TABLE B-V-1
CORROSION OF TITANIUM ALLOYS IN SODIUM CHLORIDE SOLUTIONS
Alloy
Ti
Ti
Ti
Ti
w Ti-6Al
5 yrs 4
5
5
6
6
No pits 7
None 8
Crevice corrosion 8
under Teflon gasket
9
-------
w
TABLE B-V-1
(Continued)
Time on Corrosion Rate.
Alloy
Ti-.2Pd
Ti
Ti-.15Pd
4
Ti
Ti-50A
Ti-50A
Ti-50A
Ti-50A
Ti
Ti
Ti
Ti-2Ni
Ti
Corrodent
25% NaCl
Sat. NaCl
NaCl Brine
23% NaCl in contact
with Teflon
Sea water, natural
25% NaCl
25% NaCl
25% NaCl
23% NaCl pH 1.4-1.5
23% NaCl pH 1.2-1.3
Sea water, 1.5x cone.
3.5% NaCl, pH = 1.0
1M NaCl pH = 3.0
Temperature
Boiling
Ambient
93°C
(200°F)
Boiling
2-121°C
(35-250°F)
2-82°C
(35-180°F)
82-149°C
(180-300°F)
Above 149 °C
(300°F)
Boiling
Boiling
116°C
(240°F)
Boiling
149°C
(300°F)
Test mm/yr mils/yr
48 hrs. 0.
0.
1.
— —
— —
— —
— —
— —
7 days 0.
7 days 0.
6 yrs. No
of
500 hrs. No
of
246 hrs.
0.
0.
5xlO~5 0.0006
—
—
—
—
__
0.
64-0.71 25-28
Pit ting- Crevice
Corrosion
—
—
__.
Crevice attack
No pitting
No pitting
Pits possible in
metal-to-metal
crevices
Pits
—
—
corrosion failures in 90 miles
22-gauge tubing
stress corrosion
cracking at 90%
Reference
No.
9
10
10
10
11
11
11
11
12
12
13
14
0.2% offset yield
—
Extensive crevice
corrosion
18
-------
TABLE B-V-1
(Continued)
I
Ul
Alloy
Corrodent
Temperature
Time on Corrosion Rate,
Test mm/yr Mils/yr
Fitting-Crevice Reference
Corrosion No.
Ti
Ti
Sat. NaCl pH 7.5
Sat. NaCl pH >12
107°C
(225°F)
98°C
(208°F)
0.0036
0.0036
0.14
0.14
SI. crevice attack
None
19
19
Source: Based on references listed above.
-------
TABLE B-V-2
w
I
CORROSION OF INCONEL ALLOYS IN SODIUM CHLORIDE SOLUTIONS
Alloy
625
625
625
625
825
Welded or
Plain 718
600
600
600
Corrodent
Sea water, natural
0 velocity
Sea water, natural
5 m/sec. velocity
Sea water, quiet
Sea water, 0.6 m/sec
Sea water, quiet
Sea water, quiet
Sea water
Saturated Salt Brine
pH =, 7.5
Saturated Salt Brine
pH >12
Temperature
Ambient
Ambient
Ambient
Ambient
Ambient
Ambient
Ambient
107°C
(225°F)
98°C
(208°F)
Time on Corrosion Rate,
Tesu mm/yr mils/yr
16 wks. <.003 <0.1
2 wks. 8xlO~4 0.03
7 yrs. <.003 <0.1
5.9 yrs. <.003 <0.1
3 yrs. 0.003 0.1
6 mo. 0.003 0.1
2.7 yrs. 0.002 0.08
185 days 0.027 1.05
125 days 0.134 5.27
Pitting-Crevice Reference
Corrosion No.
—
Incipient crevice
beneath fouling
0.28 mm (11 mils)
beneath Teflon
.0002 mm (,007 mils)
max. under barnacles
—
Max. pitting — 1.6
mm (63 mils)
Pitting
Severe pitting and
crevice corrosion
5
15
15
15
16
17
19
19
Source: Based on references listed above.
-------
TABLE B-V-3
w
i
CORROSION OF STAINLESS STEELS IN
SODIUM CHLORIDE SOLUTIONS
Alloy Corrodent
304 Sea water, fouling
316 Sea water, fouling
304 14% NaCl + 12% KC1
Time on
Temerature Test
Ambient 160 days
Ambient 160 days
25°C (77°F) 38 days
Corrosion Rate,
mm/yr mils/yr
0.038 1.5
<.003 <0.1
0.008 0.3
Pitting-Cr«vice
Corrosion
Pits perforated
samples
—
Pits .03-. 13 mm
Reference
No.
3
3
(1-5 3
316 14% NaCl + 12% KC1
304 Cone, sea water,
1.04-1.63x partly
deaerated
316 Cone, sea water,
1.04-1.63x partly
deaerated
304 Sea water, quiet
316 Sea water, quiet
304 4% NaCl aerated
pH 7.0
316 4% NaCl, aerated
25°C (77°F) 38 days
0.008
0.3
24-41°C 134 days 0.0002 0.008
(75-105°F)
24-41°C 134 days 0.0002 0.009
(75-105°F)
mils) in crevice
Pits .03-.13 mm (1-5 3
mils) in crevice
6
Ambient
Ambient
90°C
(194°F)
90°C
(194°F)
3 yrs.
3 yrs.
0.15 5.8
24 hrs. 0.01 0.5
max. pitting 3.0 mm 15
(119 mils), avg. pit-
ting 1.1 mm (42 mils)
max. pits 1.8 mm (72 15
mils), avg. pits 1.3
mm (50 mils)
av. max. pit depth 17
= .56 mm (22 mils)
62 pits/sq dm
No pits
17
-------
I
CO
TABLE
B-V-3
(Continued)
Alloy
304
316
316
316
310
Corrodent
Sea water
Sea water
Saturated salt brine
pH >.5
Saturated salt brine
pH >12
Sea water
Temperature
Ambient
Ambient
107°C
(225°F)
98°C
(208°F)
Ambient
Time on Corrosion Rate,
Test mm/yr mils/yr
1 yr. 0.008-
0.01
1 yr. 0.0001
185 days 0.01
125 days 0.97
42 months 0.008
.3-. 5
0.002
0.5
38
0.3
Fitting-Crevice
Corrosion
Pitted
No pits
Stress and crevice
corrosion
Stress cracking.
Crevice (corrosion
Pit depths — max.
Reference
No.
17
17
19
19
20
316
Sea water
Ambient
42 months 0.005
4.1 mm (163 mils),
avg. 1.8 mm (70 mils)
0.2 Pit depths—max. 20
1.6 mm (62 mils),
avg. 0.8 mm (32 mils)
Source: Based on references listed above.
-------
TABLE B-V-4
CORROSION OF MILD
0)
\
VO
Steel
Alloy
Carbon
steel
2.7 Ni
Mild
Mild
Mild
Mild
Carbon
Carbon
Corrodent
Sea water
Sea water
Synthetic sea water
aerated
Synthetic sea water
aerated
3% NaCl
25% NaCl
Buried in Tidal
Marsh
Sea water
Temperature
Ambient
Ambient
24-29 °C
(75-85°F)
60-6 3 °C
(140-145 °F)
Ambient
Ambient
Ambient
Ambient
STEEL IN SODIUM CHLORIDE SOLUTIONS
Time on Corrosion Rate,
Test mm/yr mils/yr
1 yr 0.104
1 yr 0.097
3 months 0.157
3 months 0.239
—
7.8 yr 0.198
15 yrs 0.08-.
4.1
3.8
6.2
9.4
—
7.8
12 3.0-4.8
Fitting-Crevice Reference
Corrosion No.
— — 3
3
— 3
3
Relative corr. rate 4
= 2.1
Relative corr. rate 4
= 0.4
Max. pit depth 7
= 3.3 mm (128 mils)
Max. pit depth = 17
Carbon Sea water
Ambient
Mild Vel. = 1.5 m/sec; 0_= 82°C
1 ppm (180°F)
Sea water, 83% salinity
42 months 1.60 6.3
30 days 1.52 60
1.1-6.1 mm (43-240
mils)
Max. pit depth 20
2.1 mm (81 mils)
avg. 1.5 mm (61 mils)
— 21
Source: Based on references listed above.
-------
TABLE B-V-5
<
i
Alloy
Ti-A-55
Comm. Zr
Inconel
Nionel
Mild
Steel
Ti-A-55
Comm. Zr
Inconel
Nionel
Mild
Steel
Ti-A-55
COMPARISON OF THE
Corrodent
55% CaCl2
saturated
55% CaCl2
saturated
55% CaCl
saturated
55% CaCl-
saturated
55% CaCl
saturated
73% CaCl 2
73% CaCl2
73% CaCl2
73% CaCl2
73% CaCl2
Saturated NaCl
CORROSION RATE OF
Temperature
113°C
(235°F)
113°C
(235°F)
113°C
(235°F)
113°C
(235°F)
113°C
(235°F)
177°C
(350°F)
177°C
(350°F)
177°C
(350°F)
177°C
(350°F)
177°C
(350°F)
21 °C
(70°F)
ZIRCONIUM
Time on
Test
84 days
84 days
84 days
84 days
84 days
49 days
49 days
49 days
49 days
49 days
208 days
WITH OTHER ALLOYS IN CHLORIDE SOLUTIONS
Corrosion Rate, Fitting-Crevice
mm/yr mils/yr Corrosion
0.
0.01
0.76
0.74
1.32
0.56
>3.0
0.08
0.03
2.90
0.
0.
0.4
30.
29.
52.
22. Severe pitting
>118. Sample dissolved
3. Slight pitting
1. Moderate pitting
114.
0.
Reference
No.
22
22
22
22
22
22
22
22
22
22
22
-------
TABLE B-V-5
(Continued)
Alloy
Comm. Zr
Mild
Steel
Ti-A-55
Comm. Zr
Mild
Steel
Ti-A-55
Comm. Zr
Mild
Steel
Zr
Zr
Corrodent
Saturated NaCl
Saturated NaCl
Saturated NaCl
Saturated NaCl
Saturated NaCl
Saturated NaCl
Saturated NaCl
Saturated NaCl
20-25% NaCl
Sea Water
Temperature
21°C
(70°F)
21°C
(70°F)
60°C
(140°F)
60°C
(140°F)
60°C
(140°F)
71°C
(160°F)
71°C
(160°F)
71°C
(160°F)
Boiling
Ambient
Time on
Test
208 days
208 days
169 days
169 days
169 days
163 days
163 days
163 days
—
15.5 wks
Corrosion Rate, Fitting-Crevice Reference
mm/yr mils/yr Corrosion No.
0.0001
0.08
0.0002
0. 0002
0.20
0.
0/0003
0.66
<.008
0.003
0.004
3. Moderate pitting
0.006
0.007
8. Slight pitting
0.
0.01
26.
<0.3
0.1
22
22
22
22
22
22
22
22
3
4
Source: Based on references listed above.
-------
Page Intentionally Blank
-------
Appendix B-V
REFERENCES
1. Buried Nuclear Waste Must Be Enclosed in Gold to Resist Corrosion.
Anti-Corrosion, June 1977.
2. Meyers, J.R. and J.A, Hall. Hot Salt Stress Corrosion Cracking of
Titanium Alloys: An Improved Model for the Mechanism. Corrosion,
33: 7, July 1977.
3. LaQue, P,L. and H.I. Copson, Corrosion Resistance of Metal and
Alloys. Reinhold Publishing Corp., N.Y., 2nd ed., 1963,
4. Uhlig, H.H., Corrosion and Corrosion Control. John Wiley & Sons,
Inc., N.Y., 1965.
5. Davis, J.A. and G.A. Gehring. The Effects of Velocity on the Sea
Water Corrosion Behavior of High Performance Ship Materials.
Materials Performance, 14 (4): 32-39, April 1975.
6, Schrieber, C.F. T.D. Boyce, B.D. Oakes, and F.H. Coley. Behavior
of Metals in Desalting Environments—Sixth Progress Report.
Materials Performance 14 (2): 9-14, February 1975,
7. Sanderson, B,T. and M. Romanoff. Performance of Commercially Pure
Titanium in Corrosive Soils. Materials Protection, 8 (4): 29-32,
April 1969.
8. Bomberger, H.B., and L.F. Plock. Methods Used to Improve Corrosion
Resistance of Titanium. Materials Protection, 8 (6): 45-48, June
1969.
9, Greco, B.C.. The Catastrophic Corrosion Behavior of Titanium
in Chemical Processes. Materials Protection, 6 (10): 22-27,
October 1967.
10. Titanium Design Data Book, Titanium Metals Corp. of America, N.Y.
11. Feige, N.G. and R.L, Kane. Experience with Titanium Structures in
Marine Service. Materials Protection and Performance, 9 (8): 13-16,
August 1970.
12. Millaway, E.E.. Titanium: Its Corrosion Behavior and Passivation.
Materials Protection, 4 (1): 17-21, January 1965.
B-V-13
-------
Appendix B-V (Continued)
REFERENCES
13. Newton, E.H., J.D, Birkett, and J.M. Ketteringham, Survey of
Materials Behavior in Large Desalting Plants Around the World.
ADL Report to the Office of Saline Water, U.S. Department of the
Interior, Washington, D.C., March 1972.
14, Sedricks, A.J., J.A.S. Green, and D.L. Novak. Electrochemical
Behavior of Ti-Ni Alloys in Acidic Chloride Solutions. Corrosion,
28 (4): 142, April 1972.
15. Huntington Alloy Products Div., The International Nickel Co., Inc.
Resistance to Corrosion. Huntington, W. Va., 1970.
16, Hohman, A.E. and W.L. Kennedy. Corrosion and Material Selection
Problems on Hydrofoil Craft. Materials Protection, 2 (9): 59,
September 1963.
17. Uhlig, H.H., Ed. Corrosion Handbook. John Wiley & Sons, Inc., N.Y.,
1948.
18. 'Griess, J.C., Jr. Crevice Corrosion of Titanium in Aqueous Salt
Solutions. Corrosion, 24 (4): 97, April 1968.
19. Turnbull, J.M. Salt Refinery Corrosion Caused by Varying
Evaporation Conditions. Corrosion 16 (7); 11-16, July 1960.
20. LaQue, F.L. and A.H. Tuthill. Economic Considerations in the
Selection of Materials for Marine Applications. Trans. Soc. of
Naval Architects and Marine Engineers, 69, New York, N.Y., 1962
(reprinted by the International Nickel Co., Inc., N,Y.)
21. Schrieber, C.F., B.D. Oakes, A.L. Whitted, and F.H. Coley,
Behavior of Metals in the Desalination Environment, 5th Progress
Report. Materials Protection and Performance, 12 (8): 32,
August 1973.
22. Gegner, P.J. and W.L. Wilson. Corrosion Resistance of Titanium and
Zirconium in Chemical Plant Exposures. Corrosion 15 (7): 341t-350t,
July 1959.
B-V-14
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