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BNWL-1900 Volume 4 HIGH-LEVEL RADIOACTIVE WASTE MANAGEMENT ALTERNATIVES SECTION 7: SECTION 8: SECTION 9: WASTE PARTITIONING EXTRATERRESTRIAL DISPOSAL TRANSMUTATION PROCESSING May 1974 4% Battelie Pacific Northwest Laboratories Richland, Washington 99352 The Library of the NOV 4 1974 University ot Illinois at Urbana-Chamoaign Prepared for the U.S. Atomic Energy Commission under Contract AT(45-1):1830 Z O O NOTICE The report was prepared as an account of work sponsored by the United States Government. Neither the United States nor the United States Atomic Energy Commission, nor any of their employees, nor any of their contractors, subcontractors, or their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed, or represents that its use would not infringe privately owned rights. PACIFIC NORTHWEST LABORATORY operated by BATTELLE for the U.S. ATOMIC ENERGY COMMISSION Under Contract AT (45-1 )-1 830 Printed in the United States of America Available irom National Technical Information Service U.S. Department of Commerce 5285 Port Royal Road Springfield, VA 22151 Price: Printed Copy $10.60; Microfische $1.45 AtC-RL RICHLAND. WASH BNWL-1900 Volume 4 HIGH-LEVEL RADIOACTIVE WASTE MANAGEMENT ALTERNATIVES SECTION 7 SECTION 8 SECTION 9 WASTE PARTITIONING EXTRATERRESTRIAL DISPOSAL TRANSMUTATION PROCESSING May 1974 Editors K. J. Schneider A. M. Platt Section 7 Contributors J. W. Bartlett, Study Leader R. E. Burns, Study Leader L. A. Bray L. L. Burger J. L. Ryan Section 8 Contributors K. Drumheller, Study Leader C. L. Brown B. Griggs R. E. Hyland et al., NASA-Lewis J. S. MacKay, NASA-Ames D. R. O'Keefe, Gulf Energy & Environmental Systems Company Section 9 Contributors R. C. Liikala, Study Leader J. B. Burnham B. F. Gore B. R. Leonard, Jr. D. L. Lessor C. W. Lindenmeir T. I. McSweeney E. T. Merrill W. C. Wolkenhauer BATTELLE PACIFIC NORTHWEST LABORATORIES RICHLAND, WASHINGTON 99352 S/.Lf 1 1 1 STUDY CONTRIBUTORS - TOTAL REPORT BNWL-1900 Following is a listing of the primary study contributors. Unless otherwise noted, they are affiliated with the Pacific Northwest Laboratory of Battelle Memorial Institute. Overall Study Coordination K. J. Schneider A. M. Piatt Background on High-Level Waste and Its Management W. K. Winegardner, Study Leader G. Jansen Study Methodology and Safety Considerations D. E. Deonigi, Study Leader J. P. Corley, Study Leader J. B. Burnham T. I . McSweeney D. H. Denham D. A. Baker J. K. Soldat G. Jansen R. C. Routson Geologic Disposal Concepts K. M. R. D. W. J. J. J. L. G. D. E. G. J. W. J. E. L. S. Fe J. R. W. D. K. R. D. N. L. Ja H. B. A. W. Th E. N. J. H. ni x Schn Krei Hal 1 Till Wi ne Shef Kase Hart Ames nsen Stew Ekre Di nw Mytt ordar Wei r H i nr Schr Wood and eider, Study Leader ter, Study Leader ace son gardner f r ley art n, u: i d d i on , son , , Jr ichs oder cock Scisson Drilling Co ISGS-Den ver e, USGS-Denver USGS-Denver USGS-Denver , USGS-Denver USGS-Denver USGS-Denver Seabed Disposal Concepts R. W. Wallace, Study Leader D. D. Tillson, Study Leader W. H. Swift J . R . Divine P. J. Valent -\ C i v i 1 Engineering H. J. Lee lLaboratory, U.S. D. G. True (Naval Construction R. J. Malloy JBatta'.ion Center Ice Sheet Disposal Concepts R. W. Wallace, Study Leader D. D. Tillson, Study Leader W. H. Swift J . R . Divine Extraterrestrial Disposal Concepts K. Drumheller, Study Leader C . L . Brown B . Gri ggs R. E. Hyland et al., NASA-Lewis J. S. MacKay, NASA-Ames D. R. O'Keefe, Gulf Energy & En v i ronmen tal Systems Company Transmutation Elimination Concepts R. C. Liikala, Study Leader B. R . Leonard , Jr. W . C . Wol kenhauer D . L . Lessor E . T . Merri 1 1 T . I . McSweeney J . B . Burnham B. F. Gore C. W. Lindenmeir Waste Partitioning R. E. Burns, Study Leader J. W. Bartlett, Study Leader L. A. Bray L . L . Burger J . L . Ryan Waste Management Costs R. W. McKee, Study Leader J . B . Burnham S. A. Rao R. D. Spillman"! Automation Indus- N. F. Stark t tries, Inc., Vitro J Engineering Division Pol icy Conf 1 icts J . B . Burnham Public Response J. B. Burnham, Study Leader S. M. Nealey "i Human Affairs W. S. Maynardr Research Center, J Battelle Digitized by the Internet Archive in 2013 http://archive.org/details/highlevelradioac04paci BNWL-1900 FOREWORD This report is a comprehensive overview study of potential alternative methods for long-term management of high-level radioactive waste. The study includes a compilation of information relevant to technical feasibility, safety, cost, environmental considerations, policy conflicts, public response and research and development needs for: 1. Disposal in terrestrial locations a. In geologic settings on land b. In the seabed c. In ice sheets 2. Disposal into space 3. Elimination by transmutation (nuclear transformation of certain waste constituents into nuclides having less long-term toxicity). The study is limited to the management of high-level radioactive waste from nuclear power by variations of these alternatives. Consideration of alterna- tive types of electrical power generation are not within the scope of the study. In addition, evaluation of interim storage of radioactive waste in retrievable surface storage facilities is not part of this study. Disposal of waste in bedded salt deposits was studied extensively in other AEC programs, and the concept is included here as part of the overall matrix of geologic disposal techniques. To complement these studies, investigations were also conducted on waste partitioning (separation of radionuclides in radioactive waste into different elements or groups of elements according to their long-term toxicity or suita- bility for different disposal methods), and systems methodology was developed to assess the effects of radionuclides from waste introduced into man's eco- logical cycle, assuming some failure of the primary waste containment. Information pertinent to evaluating the various potential waste disposal techniques was developed without promoting any single disposal concept. The study is concerned with management of the waste and does not consider the potential for recovery of resources within the waste, including the heat. Concepts are developed only to the detail necessary to describe them for the overall investigation and in general are studied on a systematic, generic basis. This information can be used in comparing and assessing the various disposal concepts as a basis for decisions regarding their further study. VI BNWL-1 900 The evaluations of feasibility are not restricted to currently available technology. Rather, the study attempts to take into account technology which can be developed or is expected to be available at least within the next four decades. Indeed, most of the concepts studied are estimated to require 15 to 30 years for full implementation. The study includes most currently known waste management alternatives, but is not considered to be all-inclusive. As new data become available, and as new or varied concepts become evident (e.g., disposal in rocks in permafrost areas, isotopic dilution of selected materials, etc.,) comparable follow-on studies will be carried out. This investigation has been performed largely by a multiple-discipline technical staff at the Pacific Northwest Laboratory of Battelle Memorial Institute with significant input from a large number of consultants and out- side contributors. This wide involvement of persons was an attempt to assure up-to-date and accurate coverage of the broad scope of subject matter, includ- ing areas where there are diversities of opinions. This report is issued as nine major sections in four volumes: Volume 1 Section 1 Summary^ ' Section 2 Background and Data Base Section 3 Evaluation Methodology Volume 2 Section 4 Geologic Disposal Volume 3 Section 5 Ice Sheet Disposal Section 6 Seabed Disposal Volume 4 Section 7 Waste Partitioning Section 8 Extraterrestrial Disposal Section 9 Transmutation Processing Appendix material is included with its own respective volume. In general, metric system units are used in this report. Conversion fac- tors to English units are given in Appendix l.A. This section is almost identical to WASH 1297, High-Level Radioactive Waste Management Alternatives , US AEC Division of Waste Management and Transportation, May 1974. VI 1 BNWL-1900 ACKNOWLEDGMENTS - TOTAL REPORT This study, performed over a period of about 1.5 years, received significant support from many people who are not listed as key contributors. The contribu- tions of these persons are gratefully acknowledged. Although the total of such participants is too numerous to mention, the following list shows many of the major contributors. Program Guidance, Funding and Review F. K. Pittman U.S. A.E.C., Division of Waste Management and Transportation A. F. Perge U.S. A.E.C., Division of Waste Management and Transportation R. W. Ramsey U.S. A.E.C., Division of Waste Management and Transportation H. F. Soule U.S. A.E.C., Division of Waste Management and Transportation 0. J. Elgert U.S. A.E.C. , Richland Office R. B. Goranson U.S. A.E.C, Richland Office N. T. Karagianes U.S. A.E.C. , Richland Office R. D. Fogerson U.S. A.E.C, Richland Office Overall Review G. H. Daly U.S. A.E.C, Division of Waste Management and Transportation M. Skalka U.S. A.E.C, Division of Waste Management and Transportation R. D. Walton U.S. A.E.C, Division of Waste Management and Transportation W. K. Eister U.S. A.E.C, Division of Waste Management and Transportation A. F. Kluk U.S. A.E.C, Division of Waste Management and Transportation V. G. Trice U.S. A.E.C, Division of Waste Management and Transportation T. L. Dunckel U.S. A.E.C, Division of Waste Management and Transportation C L. Osterberg U.S. A.E.C, Division of Bio- medical and Environmental Research H. M. Parker Battelle, Pacific Northwest Laboratory C M. Unruh Battelle, Pacific Northwest Laboratory R. F . Foster Battelle, Pacific Northwest Laboratory Study Methodology and Safety Considerations Fault Tree Consultant P. A. Crosetti United Nuclear Industries Risk and Public Response Task Force S. S. Epstein, M.D. Case Western University J. McCarrol 1 , M.D. Los Angeles Medical Services Division S . M . Nea 1 ey Battelle, Human Affairs Research Center L . H . Ra ppoport Kansas State University L. A. Sagan, M.D. Palo Alto Clinic C . Starr Electric Power Research Institute P . S 1 o v i c Oregon Research Institute N. E. Rasmussen Massachusetts Institute of Technology Geologic Disposal Concepts Re vi ewers W. S. Twenhofel U.S. Geological Survey, Denver R. K. Blankennagel U.S. Geological Survey, Denver G. D. deBuchannane U.S. Geological Survey, Wa sh i ngton , DC A. L. Boch Oak Ridge National Laboratory T. F. Lomenick Oak Ridge National Laboratory VI 1 1 BNWL-1900 W. C. McClain Oak Ridge National Laboratory J. 0. Blomeke Oak Ridge National Laboratory Consul tants R. F. Walters, Walters Drilling Co. P. F. Kerr H. A. Coombs University of Washington J. Gilluly R. L. Loofbourow G. C. Kennedy University of California Los Angeles Ice Sheet Disposal Concepts Consultant and Reviewer C. B. B. Bull Ohio State University Reviewers J . H . Zumberge University of Nebraska M . F. Meier U.S. Geological Survey E. J. Zeller University of Kansas Seabed Disposal Concepts Consultant and Reviewer M. N. A. Peterson Scripps Institution of Oceanography Reviewers and Technical Editorial Assistance W. P. Bishop Sandia Laboratories C. D. Hollister Woods Hole Oceanographic Institute Revi ewers D. A. McManus University of Washington J. S. Creager University of Washington A. J. Coyle Battelle Columbus Transmutation Concepts Revi ewers A. S. Kubo U.S. Military Academy West Point B. I . Spi nrad Oregon State University H. W. Lefevre University of Oregon C . J . Poncel et Carnegie-Mellon Institute J . L . Crandal 1 E. I. duPont de Nemours and Co. R. E. Hellens Combustion Engineering, Inc. D . G . Foster , Jr . University of California Waste Partitioning Amicon Corp., Cambridge, MA C. E. Armantrout U.S. Bureau of Mines L . E. Bruns Atlantic Richfield Hanford Co. W. W. Schulz Atlantic Richfield Hanford Co. C. R. Cooley Hanford Engineering Development Laboratory R. E. Lerch Hanford Engineering Development Laboratory G. L. Richardson Hanford Engineering Development Laboratory R. E. Leuze Oak Ridge National Laboratory D. F. Peppard Argonne National Laboratory H. C. Rathvon Exxon Nuclear T. H. Siddall Louisiana State University R. E. Sparks Washington University G. W. Watt University of Texas Report Editor J. A. Powell IX BNWL-1 900 CONTENTS TOTAL REPORT Abbreviated contents of the total 4-volume report are listed here. Detailed contents of each of the nine sections are listed at the front of each secti on . VOLUME 1 STUDY CONTRIBUTORS - TOTAL REPORT FOREWORD ACKNOWLEDGMENTS - TOTAL REPORT CONTENTS - TOTAL REPORT . 1.0 SUMMARY .... 1.1 INTRODUCTION . 1.2 HIGH-LEVEL RADIOACTIVE WASTE MANAGEMEN HIGH-LEVEL RADIOACTIVE WASTE . STUDY METHODOLOGY SAFETY CONSIDERATIONS .... DISPOSAL CONCEPTS - DESCRIPTION AND SY TECHNICAL FEASIBILITY .... RESEARCH, DEVELOPMENT, AND TIMING . WASTE MANAGEMENT COSTS. ENVIRONMENTAL CONSIDERATIONS . POLICY CONFLICTS PUBLIC RESPONSE SELECTED BIBLIOGRAPHY .... 1 1 1 1 1 1 .8 1 .9 1 .10 1.11 1 .12 1.13 STEM 2.0 BACKGROUND AND DATA BASE . 2.1 WASTE MANAGEMENT OPTIONS . 2.2 POWER PROJECTIONS 2.3 REACTOR PLANTS 2.4 FUEL REPROCESSING AND HIGH-LEVEL WASTE 2.5 HIGH-LEVEL WASTE CHARACTERISTICS . REFERENCES 3.0 EVALUATION METHODOLOGY . 3.1 SAFETY 3.2 CONCEPT COST ANALYSIS . 3.3 POLICY CONSIDERATIONS . 3.4 ENVIRONMENTAL CONSIDERATIONS 3.5 PUBLIC ATTITUDES AND PERCEPTION OF RISK REFERENCES i 1 1 v vii i x 1 .1 1 .1 1 .3 1 .6 1 .10 1.13 1 .22 1 .46 1 .63 1 .68 1 .71 1 .73 1 .74 1 .76 .1 .1 .5 ,7 .9 .1 3 .17 1 5 65 88 92 99 112 APPENDICES FOR VOLUME 1 BNWL-1900 APPENDIX APPENDIX APPENDIX APPENDIX APPENDIX APPENDIX APPENDIX APPENDIX APPENDIX APPENDIX APPENDIX APPENDIX APPENDIX APPENDIX APPENDIX APPENDIX APPENDIX APPENDIX 1 .A 1 .B 2. A 2.B 2 2 2 2 2 3 3.B 3.B 3.B 3.C 3.D 3.E 3.F 3.G CONVERSION FACTORS GENERALIZED MAPS OF THE UNITED STATES. NUCLEAR REACTORS BUILT, BEING BUILT, OR PLANNED NUCLIDES AND RCG PERMISSIBLE CONCENTRATIONS USED IN PREPARATION OF TABLES 2.8, 2.9 AND SUMMARY TABLES OF APPENDICES 2.C THROUGH 2.G LWR-U PLANT WASTE. LWR PLANT WASTE, PU FUEL FRACTION HTGR PLANT WASTE .... AI-LMFBR PLANT WASTE . GE LMFBR PLANT WASTE . SPECIMEN DOSE CALCULATION PROGRAM, 1 External Radiation Dose Factors 2 Adult Radiation Dose Factors - Ingestion 3 Adult Radiation Dose Factors - Inhalation RADIONUCLIDE TRANSFER FACTORS USED IN DOSE CALCULATIONS FOR SELECTED MEDIA . RADIOACTIVITY IN WASTE ACCUMULATED THROUGH YEAR 2000 ASSUMED RURAL POPULATION DISTRIBUTION FOR GEOLOGIC DISPOSAL ASSUMED METEOROLOGICAL DATA - WESTERN REGION QUESTIONNAIRE - PILOT PUBLIC ATTITUDE STUDY l.A.l 1 .B.l 2.A.1 B.l C.l D, E F G, 3.A.1 3. B.l 3.B.2 3.B.3 3. C.l 3.D.1 3.E.1 3.F.1 3.G.1 VOLUME 2 STUDY CONTRIBUTORS - TOTAL REPORT FOREWORD ACKNOWLEDGMENTS CONTENTS - TOTAL REPORT. 4.0 GEOLOGIC DISPOSAL . 4.1 DESCRIPTION OF GEOLOGIC DISPOSAL CONCEPTS . 4.2 SYSTEM REQUIREMENTS FOR THE CONCEPTS . 4.3 GEOLOGIC CONSIDERATIONS FOR SPECIFIC CONCEPTS 4.4 TECHNICAL FEASIBILITY 4.5 SAFETY 4.6 RESEARCH AND DEVELOPMENT NEEDS. 4.7 TIME REQUIREMENTS FOR COMMERCIAL OPERATION. 4.8 CAPITAL AND OPERATING COSTS .... 4.9 PUBLIC RESPONSE 4.10 POLICY CONSIDERATIONS 4.11 ENVIRONMENTAL IMPACT i ll v vi i ix 4.1 104 138 148 157 163 165 172 172 174 175 177 XI BNWL-1900 REFERENCES APPENDIX 4. A GLOSSARY OF GEOHYDROLOGIC TERMS APPENDIX 4.B MAJOR STRATIGRAPHIC AND TIME DIVISIONS APPENDIX 4.C CHEMICAL COMPOSITIONS (IN PERCENT) AND PHYSICAL HYDROLOGIC PROPERTIES OF PRINCIPAL ROCK TYPES APPENDIX 4.D MODIFIED MERCALLI INTENSITY SCALE . APPENDIX 4.E PROMINENT EARTHQUAKES IN THE UNITED STATES THROUGH 1970 APPENDIX 4.F DESCRIPTION OF SYSTEMS FOR COST BASES . APPENDIX 4.G TIMING REQUIREMENTS FOR RESEARCH AND BENEFICIAL OCCUPANCY OF A COMMERCIAL REPOSITORY . AND 4.177 4.A.1 4.B.1 4.C.1 4.D.1 4.E.1 4.F.1 4.G.1 VOLUME 3 STUDY CONTRIBUTORS - TOTAL REPORT FOREWORD ACKNOWLEDGMENTS - TOTAL REPORT . CONTENTS - TOTAL REPORT 5.0 ICE SHEET WASTE MANAGEMENT SYSTEM 5.1 CONCEPTS DESCRIPTION 5.2 TECHNICAL FEASIBILITY . 5.3 ESTIMATED RESEARCH AND DEVELOPMENT REQUIREMENTS 5.4 ESTIMATED TIME REQUIREMENTS FOR OPERATION . 5.5 ESTIMATED CAPITAL AND OPERATING COSTS . 5.6 PUBLIC RESPONSE, POLICY AND ENVIRONMENTAL IMPACT REFERENCES 6.0 SEABED WASTE MANAGEMENT SYSTEM .... 6.1 CONCEPTS DESCRIPTION 6.2 TECHNICAL FEASIBILITY 6.3 ESTIMATED RESEARCH AND DEVELOPMENT REQUIREMENTS 6.4 ESTIMATED TIME REQUIREMENTS FOR OPERATION . 6.5 ESTIMATED CAPITAL AND OPERATING COSTS . 6.6 PUBLIC RESPONSE, POLICY, ENVIRONMENTAL IMPACT REFERENCES i 1 1 v vi i x 1 2 17 22 29 29 33 38 1 1 17 27 35 35 38 40 APPENDICES FOR VOLUME 3 APPENDIX 5. A APPENDIX 5.B APPENDIX 5.C APPENDIX 5.D APPENDIX 6. A APPENDIX 6.B A PROPOSAL FOR THE ESTABLISHMENT OF A PERMANENT INTERNATIONAL HIGH-LEVEL RADIOACTIVE WASTE REPOSITORY IN ANTARCTICA THE GREENLAND ICE SHEET THE ANTARCTIC ICE SHEET SELECTED CHAPTERS FROM INTRODUCTION TO ANTARCTICA WASTE CANISTER MATERIAL FOR SEABED AND ICE SHEET ENVIRONMENTS FEASIBILITY STUDY OF NUCLEAR WASTE SUB-SEAFLOOR EMPLACEMENT A.l B.l C.l D.l A.l B.l XI 1 BNWL-1900 VOLUME 4 STUDY CONTRIBUTORS - TOTAL REPORT FOREWORD ACKNOWLEDGMENTS - TOTAL REPORT .... CONTENTS - TOTAL REPORT 7.0 POTENTIAL FOR WASTE PARTITIONING 7.1 BENEFITS TO BE GAINED 7.2 PARTITIONING PROCESSES 7.3 ANALYTICAL REQUIREMENTS 7.4 PARTITIONING COSTS 7.5 RESEARCH AND DEVELOPMENT REQUIREMENTS . REFERENCES 8.0 EXTRATERRESTRIAL DISPOSAL .... 8.1 EXTRATERRESTRIAL WASTE MANAGEMENT SYSTEM 8.2 TECHNICAL FEASIBILITY 8.3 RESEARCH AND DEVELOPMENT REQUIREMENTS . 8.4 TIME REQUIREMENTS FOR COMMERCIAL OPERATION 8.5 CAPITAL AND OPERATING COSTS 8.6 PUBLIC RESPONSE, POLICY AND ENVIRONMENTAL REFERENCES 9.0 TRANSMUTATION PROCESSING/ELIMINATION 9.1 TRANSMUTATION WASTE MANAGEMENT SYSTEMS . 9.2 TECHNICAL FEASIBILITY 9.3 ESTIMATED RESEARCH AND DEVELOPMENT REQUIREMENTS 9.4 ESTIMATED TIME FOR REQUIREMENTS FOR OPERATION 9.5 CAPITAL AND OPERATING COSTS 9.6 PUBLIC RESPONSE 9.7 POLICY CONSIDERATIONS 9.8 ENVIRONMENTAL CONSIDERATIONS 9.9 SAFETY ASPECTS OF ACTINIDE RECYCLE IN LWRs REFERENCES 1 1 i v vi i i x 7.1 7.1 7.3 7.10 7.10 7.11 7.14 8.1 8.1 8.11 8.53 8.58 8.59 8.62 8.72 1 1 10 25 30 31 36 36 37 42 44 APPENDICES FOR VOLUME 4 APPENDIX 8. A NASA EXECUTIVE SUMMARY - FEASIBILITY OF SPACE DISPOSAL OF RADIOACTIVE NUCLEAR WASTE . APPENDIX 8.B STUDY OF EXTRATERRESTRIAL DISPOSAL OF RADIOACTIVE WASTES PART II . APPENDIX 8.C AN EVALUATION OF SOME SPECIAL TECHNIQUES FOR NUCLEAR WASTE DISPOSAL IN SPACE APPENDIX 8.D FEASIBILITY OF USING AN ORBITING ACCELERATOR TO EJECT RADIOACTIVE WASTE PRODUCTS INTO SPACE 8.A.1 8.B.1 8.C1 8.D.1 XI 1 1 BNWL-1900 APPENDIX 8.E TERRESTRIAL AND STELLAR CONTAMINATION FROM SPACE APPENDIX 8.F APPENDIX 9. A APPENDIX 9.B APPENDIX 9 APPENDIX 9 APPENDIX 9 DISPOSAL OF NUCLEAR WASTES COST ESTIMATE DATA TRANSMUTATION BY ACCELERATORS . TRANSMUTATION BY FISSION AND THERMONUCLEAR EXPLOSIVE DEVICES TRANSMUTATION BY FISSION REACTORS . TRANSMUTATION BY FUSION (CTR) REACTORS. COMMENTS OF PEER REVIEWERS. 8. E.I F.l A.I B.l C.l D.I E.I BNWL-1900 SECTION 7: WASTE PARTITIONING Section 7 Contributors J. W. Bartlett, Study Leader R. E. Burns, Study Leader L. A. Bray L. L. Burger J. L. Ryan 7.iii BNWL-1900 CONTENTS LIST OF FIGURES LIST OF TABLES 7.0 POTENTIAL FOR WASTE PARTITIONING 7.1 BENEFITS TO BE GAINED. 7.2 PARTITIONING PROCESSES 7.2.1 Solids . 7.2.2 Separations 7.3 ANALYTICAL REQUIREMENTS . 7.4 PARTITIONING COSTS 7.5 RESEARCH AND DEVELOPMENT REQUIREMENTS REFERENCES l v i v 7.1 7.10 7.10 7.11 7.14 7.iv BNWL-1900 LIST OF FIGURES 7.1 Basic Waste Partitioning Options 7.2 Partitioning Approach Using Solvent Extraction . 7.3 Schedule for Research and Development of Partitioning 7.5 7.8 7.13 LIST OF TABLES 7.1 Partitioning Feasibility Study Conclusions: Adequacy of Existing Technology 7.7 7.1 BNWL-1 900 7.0 POTENTIAL FOR WASTE PARTITIONING High-level radioactive waste gen- erated in nuclear fuels reprocessing plants contains radionuclides having a wide range of hal f - 1 i ves- -f rom less than 100 days to millions of years. Current practice is to treat the high- level waste as a single entity through storage, solidification and perhaps ultimate disposal. Such "total waste" will remain radio- actively toxic for a time comparable to the time periods for major geolog- ical changes in the earth's crust. An alternative approach in manag- ing high-level radioactive waste in- volves separating the total waste into fractions of different half- lives. Short-lived fractions would then decay to become radi oacti vel y non-toxic in relatively short times-- times short enough to be within man's ability to control the waste storage. Long-lived fractions — significantly reduced in mass, volume and radio- active decay heat output--could be considered for other treatment, e.g., transmutation to non-radioactive or short-lived nuclides, extraterres- trial disposal, or special terres- trial disposal. Extraterrestrial disposal and transmutation processing are discussed in Sections 8 and 9, respectively, in this volume. The basic guideline for this study was to obtain a comprehensive, evalu- ative overview of separations tech- nologies and their applicability for removing the actinides and long-lived toxic fission products from high- level waste. Major results sought from the review were identification of separations concepts which show promise for high-level waste separa- tions and identification of areas where research and development is required. 7.1 BENEFITS TO BE GAINED It is potentially possible to seg- regate high-level radioactive waste into numerous fractions based on var- ious properties such as half-life, decay heat generation, and mobility in man's immediate environment. Con- sideration of the various approaches to radioactive waste management being evaluated and discussed in this report, and the waste segregation needed to pursue these approaches, led quickly to the conclusion that a waste segregation based on half-life contributes toward these approaches. Separation of waste into two frac- tions, one containing nuclides which will decay to negligible radioactive toxicity in the order of 1,000 years, i.e., having nuclides with half-lives of less than about 100 years, and a second fraction containing the long- lived nuclides, would provide waste packages fitting the needs or capa- bilities of several of these waste management alternatives. Initial considerations were based on the concept that the short-lived fraction of the waste should be decon- taminated from the long-lived frac- tion to the extent that the short- lived fraction, solidified to a cal- cine or glassy form and aged for 1,000 years, would contain no more than 1 to 2 n Ci/g of any long-lived 7.2 BNWL-1 900 nuclide. This short-lived fraction would, after about 1,000 years time, represent no significant radiological toxicity to man and could be released to man's immediate environment with no more control than that required from a chemical viewpoint. Storage of waste in manmade structures for this length of time is well within reason. Also, 1,000 years is a short time period compared to the times re.quired for major naturally occurring geological changes in the earth's crust. Disposal of the short-lived fraction in the earth's crust prior to 1,000 years would present little likelihood that the waste would be returned to man's en- vironment prior to decay to negli- gible radioactive toxicity because of naturally occurring geologic changes. Emphasis was given to separating the actinide elements from the short- lived waste f rac ti on -s i nee these elements have long-lived isotopes and a high toxicity to man. However, it became apparent that some elements other than the actinide elements also have radionuclides which present sig- nificant toxicity to man after the wastes have aged a thousand years. Significant among these are radio- nuclides of I, Tc, Sm, Sn, and Ni (Ni enters the waste as a result of some dissolution of non-core fuel element components). The decontamination fac- tors required for removal of these elements from the short-lived frac- tion of high-level waste to meet the stringent criterion stated above range from about 100 for I to greater than 10 for actinide elements. By comparison, Purex fuel reprocessing plants achieve a separation of U from fission products of 10 7 , a separation of Pu from fission products of more than 10 8 , a separation of U from Pu of more than 10 , and a separation of Pu from U of 10 6 . They achieve 99.5 to 99.9% recovery of U and Pu. If the "dilution" of waste within a geologic formation, the low leach rates of many solidified waste forms and the sorption of radionuclides in the soil are taken into account, it may well be that the concentration of long-lived radionuclides in the short- lived fraction could be considerably higher than the above criterion with- out posing a significant risk to man through terrestrial disposal of the short-lived fraction. Consideration of factors such as these is a part of the methodology for Failure Mode and Radiological Pathway Analyses discussed in Sec- tion 3 of this report series. Re- sults of future analyses using these methodologies may be used to define better the separations requirements for any given waste repository con- cept; different separations require- ments may be indicated for various re- pository concepts even if the same criteria are used. The risk analyses have been com- pleted for one sample case only and its use for determining partitioning requirements is not established. For the repository concepts analyzed to date, preliminary example use of the risk evaluation models indicated that only the actinide elements need to be removed from the short-lived waste fraction. Also indicated was that re- moval of the actinides from the short- lived fraction by a decontamination 7.3 BNWL-1 900 factor (DF) as low as 100 may be adequate. Moreover, the primary acti- nide element of direct concern is Am, and U-233, U-234 and the daughters of these isotopes are the dominant risks for time periods greater than 100,000 years. Since Pu-238 is a major source of U-234, Pu partition- ing may be required to reduce the risk from U-234. 7.2 PARTITIONING PROCESSES An overall assessment was made of the technical feasibility of chem- ically separating high-level radio- active waste generated in nuclear fuel reprocessing plants into frac- tions with different long-term bio- logical toxicity. Major emphasis was given to process concepts which would result in isolating the acti nide ele- ments from the remainder of the waste. This type of separation is called waste partitioning in this study. Less emphasis was given to the iso- lation of long-lived fission products (i.e., I, Tc , Sm and Sn) and Ni from the remainder of the waste. Separa- tion of this kind is called waste fractionation. The reasons for this relative emphasis and the interest in removing these elements from the waste were developed in the Sub- section 7.1. The results of this study are summarized here. Details of the study are presented in report BNWL-1 776 . d) Considerations of the requirements for separating actinide elements from high-level radioactive waste identi- fied three major problems or areas for study. One problem concerns the solids which are always present in high-level waste and in the process streams in a fuel reprocessing plant. The extent to which such solids con- tain actinide elements requiring re- moval to meet partitioning objectives and the extent to which they inter- fere with the performance of parti- tioning processes must be defined and dealt with. A second problem is the choice of the most advantageous pro- cessing options to achieve actinide element partitioning needed for var- ious waste disposal alternatives. Past experience and literature per- tinent to the kinds of separations required in a waste partitioning program is wery extensive. Consid- eration of this extensive background, its application to partitioning needs and choice of the most promising ap- proaches comprise a major task. A third area concerns the requirements for analytical control of waste par- titioning. The detection and measure- ment of very small amounts of acti- nide elements in the short-lived waste fraction with its very high beta-gamma radioactivity is a formid- able problem; the question of whether adequate analytical control tech- niques exist or can be developed is pertinent. The study sought to make as com- plete a review as possible of the technology pertinent to the three major problems identified above. An extensive literature review was made. Also input from recognized experts in areas pertinent to the problem was sought. Approximately 25 experts out- side BNW were contacted during this study. Included were individual con- sultants, personnel from other AEC 7.4 BNWL-1900 installations and personnel from in- dustrial firms with acknowledged ex- pertise in specific separations tech- niques. The contributions of these experts are reflected in the content of the report. Two basic concepts for achieving actinide element partitioning were defined. In one, the partitioning would be accomplished within the fuel reprocessing plant by suitable modifi- cation to improve the uranium and plu- tonium recovery and to isolate other actinide elements. In the second, the partitioning would be done on the waste as normally produced by the fuel reprocessing plant. The former is discussed as "repropartitioning" and the latter as "adjunct partition- i ng" in reference 1 . Since waste composition will de- pend on many factors (e.g., reactor type, fuel exposure history, cooling time), the study sought to make the evaluations independent of waste com- position. Separations processes will, in practice, have to accommodate waste of various types. However, most information currently available is based on waste from LWR fuel with relatively low burnup in comparison with exposures anticipated for the future. Although this study at- tempted to identify the problems which should be generic to all types of waste, the results are potentially limited by the poss i bi 1 i ty >tha t present information cannot be extrap- olated reliably to waste from the mixed nuclear power economy of the future . The chemical separations technol- ogies shown in Figure 7.1 were evalu- ated. Physical separation technol- ogies (such as gaseous or liquid diffusion or gas centrifuging) which use the larger and heavier molecules of the actinides compared with those of the smaller fission products as a basis for separation, may also have some potential for partitioning, but study of these concepts was beyond the scope of this investigation. Figure 7.1 illustrates the fact that suspended solids or colloids in the waste are potentially a significant problem which could limit or define the feasible separations technologies Solids are expected to be present in the aqueous waste; if treatment of them is needed, operations such as shown in Figure 7.1 would be needed prior to separations. As shown in Figure 7.1, the possibility that sepa- rations would have to be done on a solid waste phase introduces the need to consider pyrometal 1 urgi cal , slag- ging, and other solids- based separa- tions techniques within the scope of this study. Rapid decisions on the feasibility of concepts were sought via consulta- tion with the team of acknowledged experts mentioned earlier. Those con- cepts which appeared to have little chance of achieving the degree of separation needed for partitioning, which are inherently costly or which would require very extensive Research and Development for application to the partitioning problem were deempha- sized in favor of the more promising concepts . 7.5 BNWL-1 900 HIGH-LEVEL WASTE LIQUID SOLIDS COLLOIDS SLAGGING PYROMET- ALLURGY HOMOGENIZE TO SOLID i i. w VOLATILITY SOLID FRACTION OTHER? SEPARATE SOLID LIQUID DISSOLVE SOLIDS, COLLOIDS LIQUID FRACTION EXTRACTION ION EXCHANGE DIALYSIS CHROMA- TOGRAPHY SORPTION ELECTRO- PHORESIS OTHER? SHORT-LIVED PRODUCT FRACTION LONG-LIVED PRODUCT FRACTION(S) FIGURE 7. 1 . Basic Waste Partitioning Options 7.2.1 Solids Solids will play a prominent role in the development of waste partition- ing processes. Some of these solids are present in the dissolved fuel solution prior to solvent extraction in a reprocessing plant. Although the literature contains some infor- mation on quantities and composition of dissolver solution and high-level waste solids, most of the data are based on laboratory-scale studies. Very little data are available based on plant-scale operation with power reactor fuels. Plutonium has been identified in dissolver solution solids, but little is known about the occurrence of other actinide elements on these solids. At present there is essentially no information useful to waste partition studies on the con- tent of high-level waste solids. Re- search and Development on the content of dissolver and waste solids, on the separation of the solids from the liq- uid and on the processing of the sol- ids to remove actinide elements is needed. Potential procedures for partition- ing waste solids (e.g., solidified waste or solids removed from waste slurries) as shown in Figure 7.1 were evaluated. It was concluded that such separations as slagging, volatil- ity processes and pyrometa 1 1 urgi cal processes present formidable problems 7.6 BNWL-1900 with relatively low probability of achieving the degree of separation needed for waste partitioning. The separations achieved in most of these processes are not clean. They pre- sent major problems in terms of high- temperature operation, control of volatile components, possible pro- duction of large quantities of con- taminated waste and relatively high operating hazard. In general, major Research and Development including extension of basic knowledge would be required to develop feasible and ef- fective partitioning processes. These processes were deemphasized early in the evaluation review. If removal of actinide elements from solids separated from waste slurries is required (which remains to be determined from future studies on the actinide element content of such solids and the results of the Failure Mode and Radiological Pathway Analyses studies), development of pro- cedures for accomplishing this will be required. Fusion-leaching proce- dures offer some promise. If ade- quate separation of actinides from the waste solids is impractical to achieve, the solids could be con- sidered for combining with the acti- nide waste fraction. If further study shows long-lived radionuclide content of the solids to be ade- quately low, the solids may be added to the short-lived waste fraction. 7.2.2 Separations In considering the application of known separations processes to parti- tioning of waste, it should be remem- bered that the objective is essen- tially the reverse of that for a fuel reprocessing plant. In the latter a high degree of separation of U and Pu and perhaps Np from all other con- stituents of the irradiated fuel (and from processing equipment corrosion products) is sought. While high re- covery of U and Pu (i.e., separation of these two elements from the remain- der of the constituents) is desirable, it is not economically practical to achieve a degree of separation compa- rable to that required for waste par- titioning. On the other hand, the objective of waste partitioning is to achieve a high degree of separation of all of the actinide elements from the remainder of the irradiated fuel constituents. Separations processes which have been developed for nuclear materials have sought a reasonable recovery and high purification of a desired constituent rather than a very high degree of removal of the desired constituent from the waste. As shown in Figure 7.1, many basic liquid-phase partitioning concepts were considered. A consensus was quickly established, however, among the staff and consultants who parti- cipated in the study, that solvent extraction and ion exchange offered far greater promise for effectiveness and flexibility than the alternative concepts. This consensus was con- firmed by communication with experts on the other concepts (e.g., flota- tion, dialysis, etc.). These other concepts are at this time not suffi- ciently developed, or are not suffi- ciently effective for use, or are beset by process problems (e.g., mem- brane degradation) that would pre- clude their use. For these reasons, major attention was centered on 7.7 BNWL-1 900 solvent extraction and ion exchange. Use of these techniques for separa- tions such as would be required by partitioning has been widely explored. Their use would therefore represent relatively straightforward extension of existing technology, maximum com- patibility with present operations, and minimum investment in future Re- search and Development programs. A summary of separations feasibility conclusions is given in Table 7.1. Detailed evaluation of problems which could limit practical applica- tion of solvent extraction and ion exchange (e.g., generation of large volumes of low-level liquid waste or contaminated waste ion-exchange resin) was beyond the scope of this work. This evaluation can be done with maximum cost effectiveness after actual partitioning requirements are better defined. A large volume of literature on solvent extraction and ion exchange (abstracted and referenced in Refer- ence 1) was reviewed with respect to application to the separations needed for a waste partitioning pro- gram. This survey of solvent extrac- tion indicates that the approach can be used to accomplish the partition- ing and/or fractionation that might be needed. Extraction of the acti- nides, Sm, Tc and probably Sn, can be achieved. Tracer experiments have shown that the concentrations in the remaining waste can be reduced to low levels at least for the actinides, and conventional techniques employing multistage extraction should accom- plish adequate removal and separation Nickel may be a special problem and every attempt should be made to keep it out of the feed. However, selec- tivity to provide a small package of TABLE 7.1 . Partitioning Feasibility Study Conclusions: Adequacy of Existing Technology ( a ) Actinide Separation (DF] 10-100 1,000-10,000 Yes 10 6 -10 8 Solvent Extraction Yes Possibly Ion Exchange Yes Possibly No Other Separation Possibly No No Techniques Analytical Capabil ity Yes Possibly No a. Existing separations technology needs adaptation to the objectives of partitioning. b. This study was concerned primarily with the adequacy of existing technology for obtaining adequate separation of actinide elements from the short-lived waste fraction. Adequate technology exists for obtaining needed purity of the separated actinide fraction. 7.8 BNWL-1900 these elements has not been adequate- ly demonstrated. Complications from radiation damage to the solvent are of some concern, although for a ten- year-cooled waste these do not appear serious at first consideration. Ca- pability of partitioning without pro- ducing major new waste streams either contaminated with the problem ele- ments or so large in volume that they create new storage problems remains to be demonstrated. Research is re- quired to determine the best solvents to make the separations at the de- sired points. A several-step process possibly involving more than one sol- vent may be required. Such an approach is shown in Figure 7.2. A logical approach is to follow one or more solvent extraction steps with another technique such as ion exchange . This review has generally assumed that waste which has been stored in the (nitric) acid form would be the feed for solvent extraction. From a materials standpoint, this is the preferred situation, although other acid systems are not ruled out at this time. Information reviewed in this study indicates that ion-exchange technol- ogy that can meet low-to-modest par- titioning or fractionation needs is available. Some experience has been obtained with each of the elements that might have to be isolated. Process problems rather than sepa- rations effectiveness are most likely to restrict the practical use of ion exchange for partitioning or frac- tionation. The major problem is that lanthanides would be absorbed with Am and Cm with consequent high radiation AQUEOUS WASTE U, Pu,Np,Am,Cm,Lan. SOLVENT SOLVENT SOLVENT EXTRACTION SOLVENT U.Pu.Np STRIP SOLUTION STRIP AQUEOUS Am, Cm, Lan. SOLVENT EXTRACTION SOLVENT STRIP SOLUTION AQUEOUS U,Pu,Np AQUEOUS, WASTE RESIDUE SOLVENT Am, Cm, Lan. STRIP SOLVENT Lan. STRIP SOLUTION AQUEOUS , Am, Cm AQUEOUS Lan. ► STRIP SOLVENT FIGURE 7.2 . Partitioning Approach Using Solvent Extraction 7.9 BNWL-1 900 dose to the exchangers and need for further operations to remove the lanthanides. Thus, potential process problems include corrosion, addition of extra components and volumes to the waste, gassing of the resin bed, handling and disposal of contaminated waste resin, solids formation, short resin bed life, and a need to avoid high radiation levels. Storage of waste for up to ten years prior to parti- tioning appears mandatory if problems from high radiation levels are to be avoided. Many of the ion-exchange tech- niques have not been demonstrated on a scale corresponding to process operations. Extensive scale-up de- velopment would probably be necessary. Also, it can be anticipated that highly precise process control would be required in practice. Absorption or adsorption processes other than ion exchange in general have the same disadvantages as inor- ganic ion exchange. The selectivi- ties of such absorbents or adsorbents are sometimes high, but their capaci- ties are generally low, and they tend to be better suited to the removal of trace constituents from low ionic strength solutions than to partition- ing of macro components from strongly electrolytic solutions (which is the situation for partitioning). A pos- sible application might be the use of a solid absorbent such as activated carbon to remove iodine from either a solution or a gas stream after vola- tilization of iodine. There are undoubtedly a rather large number of precipitation methods that could be devised for partition- ing of radioactive wastes. Consider- able research during the Manhattan Project and for some time later was devoted to precipitation methods, both direct and carrier, for process- scale as well as analytical and research-scale separations. To a very large extent these methods have been replaced on the process scale by solvent extraction and ion-exchange methods. The reasons why this is so for large-scale processing, but not necessarily so for analytical and re- search applications is that recovery of the desired product is improved and waste volumes generally are smaller with extraction. Virtually all known separations methods have been proposed and exam- ined for the separation and/or puri- fication of nuclear materials. Most have been considered for radioactive wastes. The methods include tech- niques such as: volatility, biolog- ical, flotation and el ectrof 1 otati on , molecular sieve filtration of metal complexes, dialysis, electrophoresis, and other electrochemical and pyro- chemical methods. All of these methods are deficient in one or more ways when considered for the present problem, although it is possible that some may have appli- cation to final separation of spe- cific nuclides. Deficiencies for these methods (in comparison with solvent extraction and ion exchange) were not evaluated in detail during the course of this study. It was concluded that achieving actinide element partitioning en- tirely by modifications to and 7.10 BNWL-1900 improved operating conditions in cur- rent and near-future Purex-type fuel reprocessing plants is not practical. Current flowsheets deliberately route the trivalent actinide elements to the first- cycle waste along with the lanthanides and other fission prod- ucts. Extensive modifications would be required to separate these acti- nides from the total waste in a Purex proces s . Some potential exists for improv- ing U and Pu recovery in present Purex processes through modification of operating conditions and/or adding new columns, particularly if very high add i tonal decontamination fac- tors for these elements are not re- quired. Coupling this approach with an adjunct facility for removal of trivalent actinides should be consid- ered. The solvent extraction flow- sheet studied at Karlsruhe^ ; for re- covering Am and Cm from high-level waste should be considered a candi- date for this. 7.3 ANALYTICAL REQUIREMENTS In practice, analytical measure- ments will be needed to confirm that decontamination goals required of the partitioning or fractionation process have actually been achieved. Low- level concentrations of a-emitting isotopes will have to be detected and measured with high accuracy in a strong B-y field. Although not a re- quirement, these measurements should be made on-line (during partitioning operations) to avoid possible need for post-operations holdup (to make measurements) and recycle of out-of- specifications waste. A detailed review of the analyti- cal problems associated with a waste partitioning program was made both with- respect to the general problems of sampling in heterogeneous systems and with respect to detection and measurement of specific elements of concern. This review indicated that the state-of-the-art for analyt-ical measurements may not be equal to the requirements for partitioning, de- pending on the decontamination fac- tors required. Developmental work may be required to provide the ana- lytical measurements needed to assure control of partitioning processes and to verify whether the required sepa- rations have been accomplished. Certainly, further analytical devel- opment will be needed if high decon- tamination factors are required. The presence of solids in the waste and the potential for other solids being generated during the partitioning processes present major analytical problems both from the standpoint of being able to obtain representative samples and of assur- ing that the analytical procedures properly account for the constituents of the solids. In many cases, separation of the measured constituent from the bulk sample will be required. Developing required separations procedures and adapting them for on-line control will be major problems. 7.4 PARTITIONING COSTS The costs to separate high-level waste into a long-lived and a short- lived fraction will, of course, de- pend on the degree of separation re- quired, the number of elements which 7.11 BNWL-1900 must be removed from the short-lived fraction and the required purity of the long-lived fraction. A Delphi- type procedure was used to obtain an estimate of the cost to produce a short-lived fraction which will decay to negligible radioactive toxicity in about 1,000 years, i.e., removing the long-lived elements to very high de- contamination factors. Several peo- ple knowledgeable in the costs of fuel reprocessing were asked to esti- mate the factor by which fuel repro- cessing costs would be increased by the addition of facilities to accom- plish this separation. The estimates ranged from a factor 1.25 to a factor of 4, with a mean factor of 2 corre- sponding to about $35,000 per metric ton of fuel for partitioning. Another cost estimate was made on separating 99% (DF= 1 00 ) of the acti- nide elements only from high-level waste. This degree of separation is similar to that achieved in a process developed at Karlsruhe for removinq (2 ) Am and Cm from waste, and that pro- cess was considered in making the es- timate. The volume of concentrated high-level waste to be treated per 1,000 kilograms of irradiated fuel processed is less by a factor of about seven than the volume of feed to the reprocessing plant. The basic reprocessing cost, c_a. $35,000 per 1,000 kg was reduced accordingly by using a (x) 1 / 2 scaling factor for the reduction in plant size. The result- ing cost was further modified by com- paring the number of process cycles for partitioning to the number re- quired for fuel reprocessing. The following total cost estimates for partitioning were derived. Acti ni des plus 1 % of the Fission Products Actinides less U + 1% of the Fission Products Actinides less U + 0.1% of the Fission Products Cost/MT Fuel , $ 10,000 15,000 20,000 The first case is believed to ap- proach the need for transmutation processing (see Section 9). The lat- ter two cases are for the major ex- traterrestrial disposal cases in this study (see Section 8). 7.5 RESEARCH AND DEVELOPMENT REQUIREMENTS Data are needed on the accomplish- ments possible toward long-lived iso- tope removal and the cost involved. These will permit assessment of the technical and economic feasibility of the various approaches to managing the long-lived fraction as well as the overall economic impact on the nuclear fuel cycle. Hence, Research and Development work on partitioning (and fractionation) should not be de- layed by awaiting completion of the alternative disposal studies. The following summary of waste separation Research and Development requirements is based on the assump- tion that waste fractionation will not be required and that removal of actinide elements (partitioning) from the short-lived fraction to less than 1% ( D F> 1 00 ) of their concentration in fuel reprocessing plant waste will be required. The overall Research and 7.12 BNWL-1 900 Development program for this basis is estimated to cost $3 to $5 million and to require about 5 years to com- plete.^ If further studies, as noted earlier, indicate waste frac- tionation or higher actinide decon- tamination factors are required, the Research and Development program must be modified accordingly. It has been emphasized during these studies that solids in waste constitute a major problem in waste partitioning and that little is known about the amount and composition of such solids. A program to obtain and characterize waste comparable to that expected from fuel reprocessing plants is needed. Resolution of problems of adequate solid-liquid separation, determination of whether treatment of the solids to remove ac- tinides is required and definition of processes for treating the solids would be part of this program. Laboratory-scale studies are re- quired to test conceptual flowsheets for attaining the required actinide element partitioning. Ideally these should be done with processing plant waste or process streams. However, because these are not readily avail- able, work with simulated materials should be done initially to define basic separations parameters; prob- lems posed by plant-derived waste would be resolved as such wastes be- come available. As conceptual flowsheets are tested and processes are developed, capital and operating cost estimates pertinent to the processes are needed both to guide the selection of a pro- cess or processes for further study and to provide cost input to other facets of waste management studies. Pilot plant scale testing of the most promising flowsheets will be needed. This will entail facility design, construction or modification of a facility and operation of the facility to demonstrate the flow- sheets, resolve problems posed by plant scale operation and provide more reliable cost data. Development of analytical tech- niques should parallel flowsheet de- velopment studies. Specific analyti- cal requirements will be defined in part by the flowsheets devised. These developments should be sched- uled to permit testing of in-line analytical techniques during pilot plant testing of flowsheets. An approximate schedule for the needed Research and Development pro- gram is shown in Figure 7.3. a. It is assumed that the facility exists for pilot-plant testing, and costs for such a facility are not included in this estimate. 7.13 BNWL-1900 YEAR 1. OBTAIN AND CHARACTERIZE WASTE 2. EXPLORATORY DEVELOPMENT OF CANDIDATE FLOWSHEETS 3. DETAILED DEVELOPMENT OF SELECTED FLOWSHEETS 4. COST ESTIMATES ON SELECTED FLOWSHEETS 5. DESIGN AND CONSTRUCTION OR MODIFICATION OF PILOT PLANT FACILITY 6. START-UP AND OPERATION OF PILOT PLANT FACILITY 7. ANALYTICAL PROCEDURES DEVELOPMENT AND TESTING 1 2 3 4 5 FIGURE 7.3 . Schedule for Research and Development of Partitioning 7.14 BNWL-1900 REFERENCES FOR SECTION 7 1. Bartlett, J. W., Coordinator, Fea- sibil i t y Evaluation and R&D Pro- gram Plan for Transuranic Parti- tioning of High-Level Fuel Reprocessing Waste , USAEC Report BNWL-1776, Battelle, Pacific Northwest Laboratory, Richland, WA, November 1973. 2. Koch, G. et al., Recovery of Transpl uton ium Elements from Fuel Reprocessing High-Level Waste Solutions , Report No. KFK-1651, Karlsruhe, Germany, November 1972. Other references representative of the literature surveyed in making this feasibility study are as follows. Brooksbank, R. E. and W. T. McDuffee, Recovery of Plutonium and Other Trans- uranium Elements from Irradiated Plu- tonium Aluminum Alloy by Ion Exchange Methods , ORNL-3566, Oak Ridge National Laboratory, Oak Ridge, TN, 1964. Burger, L. L., "Neutral Organophos- phorus Compounds as Ex trac tants , " Nucl . Sci . Eng . , vol. 16, p. 428, 1963. Cleveland, J. M., The Chemistry of Plutonium , Gordon and Breach Science Publ ishers , New York , 1970. Leuze, R. E., R. D. Baybarz and B. Weaver, "Application of Amine and Phosphonate Extractants to Transplu- tonium Element Production," Nucl. Sci Eng . , vol . 17, p. 252, 1963. Orth, D. A. et al., "Isolation of Transpl utonium Elements," Proceedings International Solvent Extraction Con- f erence , Ind . , p . The Hague, 534, 1971 . Soc. of Chem. Peppard, D. F tion of Metal , "L i qu i d-L i qui d Ex trac Ions , " Advances in Inorganic Chemistry and Radiochemis - try , vol. 7, H. J. Emeleus and A. G. Sharp, Editors, Academic Press 1965. Reactor Handbook, Vol. II, Fuel Re - processing - Section B, Aqueous Sepa ration R. New Processes , S . M Editors > . Ri chards York, 1961 Stoller Interscience, Rhode, K. L. and J. A of Irradiation on the Buckham, Effect Dissolution Characteristics of Important Fuel and Cladding Materials , IAEA Panel on Re- processing of Highly Irradiated Fuels Vienna, Austria, May 27-30, 1969. Siddall, T. H., "Organophosphorus Com- pounds Other Than TBP for Processing Irradiated Fuels and By-Products , " Symposium on Aqueous Reprocessing Chemistry for Irradiated Fuels , Brussel s , 1 963. Stevenson, C. E. and D. M. Paige, "Aqueous Processing," Reactor and Fuel Processing Technology , vol. 11, no. 2, Spring 1968. Weaver, B. and R. R. Shoun, "Compari- son of Some Monoacidic Organophos- phorus Esters as Lanthanide-Actinide Extractors and Separators," J . Inorg . Nucl . Chem . , vol. 33, p. 1909, 1971. Wheelright, E. J Generic Nuclear Industry Ion Exchange: A Process for the Recovery and Final Purification of Am, Cm, Pm , Sr, Pu, Np, Cs, Tc , Rh , and Pd , BNWL-SA-1 945 , October 1968. BNWL-1900 SECTION 8: EXTRATERRESTRIAL DISPOSAL Section 8 Contributors K. Drumheller, Study Leader C. L. Brown B. Griggs R. E. Hyland et al., NASA-Lewis J. S. MacKay, NASA-Ames D. R. O'Keefe, Gulf Energy & Environmental Systems Company 8.iii BNWL-1900 CONTENTS SYSTEM ori ty LIST OF FIGURES LIST OF TABLES. .... 8.0 EXTRATERRESTRIAL DISPOSAL. 8.1 EXTRATERRESTRIAL WASTE MANAGEMENT 8.1.1 Concepts Description 8.1.1.1 Waste Content and Disposal Pr 8.1.1.2 Waste Capsule. 3 SpaceFlight. Systems Requirements 1 Waste Treatment and Handling 2 Site Preparation . 3 Transportation to the Site Operation of the Site. Final Monitoring . TECHNICAL FEASIBILITY. 8.2.1 Capsule Design Criteria. 8.2.2 Capsule Design Details . 8.2.3 Encapsulation Processes. 8.2.4 Ground Transportation 8.2.5 Space Fl ight .... 8.2.5.1 Potential Destinations 8.2.5.2 Potential Space Transportation Vehicle Performance and Cost . 8.1.1 8.1.2 8.1 8.1 8.1 8.1 .2.4 8.1.2.5 8.2 8.2.6 8.2.7 8.2.7, 8.2.7, 1 2 8.2.7.3 8.2.7.4 8.2.7.5 8.2.7.6 8.3.2 8.3.3 8.3.4 8.3.5 Energy Balance .... Safety . . .... Nuclear Safety Requirements Accident Model Considerations Analytical Results of Accident Model Recovery of Nuclear Waste Package. Summary of Safety Analyses by NASA Space Contamination .... 8.3 RESEARCH AND DEVELOPMENT REQUIREMENTS. 8.3.1 Waste Content Disposal Priority Partitioning Capsul e Desi gn Encapsulation Process Development 8.3.6 Ground Transportation .... 8. 3. 7 Space Fl ight 8.3.8 Economic Analysis 8.3.9 Safety Analysis 8.3.10 Public Response v vi 8.1 8.1 8.1 8.2 8.5 8.6 8.8 8.8 8.9 8.9 8.11 8.1 1 8.11 8.11 8.13 8.23 8.26 8.26 8.29 8.38 8.44 8.46 8.46 8.47 8.48 8.51 8.51 8.51 8.53 8.54 8.54 8.55 8.55 8.56 8.57 8.57 8.58 8.58 8.58 8.iv BNWL-1900 CONSIDERATIONS 8.4 TIME REQUIREMENTS FOR COMMERCIAL OPERATION 8.5 CAPITAL AND OPERATING COSTS . 8.5.1 Partitioning 8.5.2 Encapsulation Costs 8.5.3 Costs for Space Transportation . 8.5.3.1 Launch Costs 8.5.3.2 Ground Facilities Costs 8.5.3.3 Total Space Transportation Costs . 8.5.4 Cost Summary 8.6 PUBLIC RESPONSE, POLICY AND ENVIRONMENTAL 8.6.1 PublicResponse 8.6. 2 Pol icy Confl icts 8.6.3 Environmental Considerations REFERENCES APPENDIX 8. A NASA Executive Summary - Feasibility of Space Disposal of Radioactive Nuclear Waste APPENDIX 8.B Study of Extraterrestrial Disposal of Radioactive Wastes-Part II .... . APPENDIX 8.C An Evaluation of Some Special Techniques for Nuclear Waste Disposal in Space . APPENDIX 8.D Feasibility of Using an Orbiting Accelerator to Eject Radioactive Waste Products into Space . APPENDIX 8.E Terrestrial and Stellar Contamination from Space Disposal of Nuclear Wastes ..... APPENDIX 8.F Cost Estimate Data 8.58 8.59 8.59 8.59 8.60 8.61 8.61 8.62 8.62 8.62 8.62 8.64 8.66 8.72 8.A.1 8.B.1 8.C.1 8.D.1 8.E.1 8.F.1 8.v BNWL-1900 LIST OF FIGURES 8.1 Space Disposal System 8.2 Total Waste Capsule for Space Disposal . 8.3 Transuranic Waste Capsule for Space Disposal 8.4 Re-entry Shield and Transuranic Disposal Package for Solar Escape Destination 8.5 Launch Systems for Solar Orbit, High Earth Orbits and Solar Escape .... 8.6 System Requirements for Managing High-Level by Extraterrestrial Disposal 8.7 Typical Front Re-entry Shield Composite for Radioactive Waste Package .... 8.8 Total Package Weight Ratio for Different Exterior Dose Rates 8.9 Steps in Transuranic Oxide Encapsulation Process for Space Disposal .... 8.10 Illustration of Solar Sail Vehicle 8.11 Possible Launch Vehicles for Space Disposa of Radioactive Waste 8.12 Research and Development Program - Space Disposal of Transuranics . 8.2 . 8.6 1 8.7 8.7 • • • • . 8.9 Radioactive Waste 8.10 8.17 8.20 • • • • • 8.25 8.28 8.39 8.54 8.vi BNWL-1 900 LIST OF TABLES Quantity of Actinides in Waste from Reprocessing Various Reactor Fuels 10-Year Decay Projected Quantities of Uranium, Plutonium and Thorium in the Spent Fuel Shipped to the Reprocessing Plant Example Waste Transuranic Quantities and Mission Requirements for Space Disposal Summary of Potential Space Destinations Summary Data on Package Configuration for Transuranic Disposal to Space .... Thermal Power and Radioactivity of Transuranics in 10-Year-01d Waste Composition of Transuranic Waste in Space Disposal Capsule for Criticality Analysis 8.8 Materials and Dimensions of Space Disposal Capsule for Criticality Analysis 8.9 Estimated Fraction of a Minimum Critical Mass Represented by Each Transuranic Nuclide in One Sphere 8.10 Summary of Potential Space Destinations 8.11 Space Launch Vehicle Performance and Cost for High Earth Orbits and Solar Orbits 8.12 Space Launch Vehicle Performance/Cost for the Direct Solar Escape Mission 8.13 Space Transportation Cost for Disposal of Transuranic Waste 8.14 Estimated Space Transportation Cost for Disposal of Transuranic Waste 8.15 Cost for the Disposal of Transuranium Elements in Space . . ." 8.16 Combustion Products of Concern Emitted by the Space Shuttle into Selected Atmospheric Layers 8.4 8.4 8.5 8.8 8.19 8.20 8.22 8.22 8.23 8.37 8.40 8.45 8.63 8.63 8.64 8.69 8.1 BNWL-1900 8.0 EXTRATERRESTRIAL DISPOSAL This study was conducted concur- rently and cooperatively with a study by the National Aeronautics and Space Administration. NASA Lewis Research Center, Cleveland, Ohio, was the lead organization for the NASA study. Many NASA organizations as well as the Space Nuclear Systems Division of the Atomic Energy Commission contrib- uted to their study. NASA has pub- lished several reports relating to the study. These are referenced or included as appendices. This study includes several direct extractions from NASA documents. The approach to the study was to analyze space disposal of nuclear waste using existing technology as the base case and to consider advanced systems as potential improve- ments on the base case. 8.1 EXTRATERRESTRIAL WASTE MANAGEMENT SYSTEM A preliminary conclusion of the study was that partitioning of the waste and shipment of only the trans- uranium elements to space is more likely to be practical than shipment of the total waste. From that initial observation, the study empha- sis was on space disposal of selected, long-lived portions of the waste. However, with improvements in tech- nology or with significantly lower disposal costs, space disposal of the total waste could become more practical. 3 ' 8.1.1 Concepts Description The basic concept of extraterres- trial disposal includes packaging waste materials in a safe manner and transporting the material by rocket or other means to a location off the earth. Several different trajec- tories have been considered. These ■ i a (b) include: 1 . A high earth orbit on the order of 150,000 kilometers. 2. Transport to the sun. 3. A solar orbit other than that of the earth and planets. 4. Solar system escape. Vehicles used in the analysis as current technology include existing space vehicles and the planned space shuttle and tugs. A shuttle-tug combination is illustrated in Figure 8.1. The manned space shuttle launch vehicle with an expendable external propellant tank is placed into a low-earth orbit, with lift-off assistance by two solid-fueled rocket motors. After firing, the solid- fueled motors are separated and dropped into the ocean where they are recovered. After use, the expendable external propellant tank is separated from the orbiting shuttle and then de-orbited by a small retro-rocket. The tug, with the waste payload, is deployed from the payload bay of the orb iter. For some space destinations (e.g., escape from the solar system), a second tug is required to provide the final stage propulsion; thus two a. This point is not a NASA conclusion. b. Placement of the waste material on an extraterrestrial body such as the moon, asteroids, and other planets was considered only briefly by NASA for reasons discussed later in this report. 8.2 BNWL-1900 TO FINAL SPACE DESTINATION TUG WITH WASTE SHUTTLE TO EARTH WASTE CONTAINERS SHUTTLE IN LOW EARTH A-» ■ EXPENDABLE EXTERNAL PROPELLENT TANK SOLID FUEL ROCKET MOTORS LAUNCH HUTTLEWITH TUG AND PAYLOAD FIGURE 8. 1 . Space Disposal System space shuttle launches must be made for one waste package. The orbiting shuttle returns to earth at a pre- scribed landing site and would be ready for a repeat flight in about two weeks . 8.1.1.1 Waste Content and Disposal Priori ty Space disposal is unique in that the cost of transportation is likely to be over $2,000/kg of payload. Thus the weight of shielding creates an unusually large economi c v penal ty . There are also unique requirements for capsule integrity to provide a reasonable degree of assurance of sur- vival in the case of abort. The means of cooling are limited. Hence the heat content of the waste capsule is significant. The weight of the waste itself is also significant. Because of the unusual require- ments for space disposal, the deter- mination of waste content and pri- ority for disposal is particularly significant. A primary consideration in estab- lishing a priority for disposal is the potential long-term toxicity to man and the environment if the iso- tope is released. Many studies have been made of the relative harmfulness ( 2-5 ) of different isotopes. ' A review of long-term hazards relative to these waste disposal options is in- cluded in Volume 1 of this report series. Most investigators generally conclude that the actinides are the most toxic in the long term. For this reason, this study suggests that transuranium elements (actinides minus uranium) receive first priority for extraterrestrial disposal. 8.3 BNWL-1900 I o d i n e - 1 2 9 and technetium-99 are examples of fission product elements which may have long-lived toxicity comparable to that of the actinides. These fission products are considered by some to be almost nonradioactive, but there are others who feel that they are very toxic. Several early investigators of space disposal or other permanent disposal methods considered primarily shipping either the fission products with high radiation levels or the bulk high-level wastes without parti- tioning. " ' However, it seems reasonable to store fission products such as cesium and strontium, which have relatively short half-lives in the context of this study, in care- fully designed manmade facilities which can assure no release to man's immediate environment during their lifetime. Radioactive cesium-137 and strontium-90 would decay by factors of several billion in one thousand years. Actinides as a group will decay by a factor less than 10 in one thousand years, and will be toxic for thousands of years. Three different cases of waste con- tent for space disposal were consid- ered in this study: Case 1 . Package the bulk waste as it comes from the reprocessing plant in solid form. The package can in- clude incorporation of the waste in forms such as glasses, ceramics or cermets. For this study the waste is assumed to be incorporated into boro- silicate glass. Case 2a . Remove nearly all fis- sion products and uranium and thorium from waste, leaving only the trans- uranium elements for disposal. Only 1% of the fission products and 0.5% of the uranium are included with the transuran i cs . Transuranium elements contaminated with fission products will be in an oxide form. Case 2b . Case 2a with 0.1% of fission products remaining with the transuranics for space disposal. Case 3 . Same as Case 2b with 99% of curium also removed from the transuranics to be disposed of into space . A preliminary analysis of Case 1 indicated that it would have very high costs, high package integrity would be difficult to achieve, and the number of space flights would be excessive (in the thousands of flights per year in the year 2000). Therefore, this case received less detailed consideration than Case 2. Cases 2a, 2b, and 3 would require chemical separation (partitioning) of the constituents disposed of extraterrestr i al ly from the remaining constituents, which would require separate management. As the study progressed, Case 2a and 2b appeared to be the closest to the most prac- tical from the standpoint of dis- posal into space, and these received the most emphasis for this study. In practice, the cost of additional par- titioning and the additional weight which must be shipped would be bal- anced to achieve an optimum safety and cost. Case 3 was considered only briefly Curium-244 has a short half-life and provides up to 87% of the radioactiv- ity and 93% of the heat in the LWR actinide waste. Thus removal of the curium could reduce the space package weight significantly. If the curium 8.4 BNWL-1900 were removed, the curium would prob- ably be stored on earth for around 100 years until it decays to pluto- nium and other daughters. These longer-lived daughters with much re- duced heat output and radioactivity could then be shipped to space sepa- rately at significantly lower cost. For space disposal purposes the waste is assumed to be held for 10 years after reprocessing. This is primarily to allow decay of the hot- test isotopes. In practice an opti- mum time would be arrived at based on separations costs, encapsulation re- quirements, and space flight costs. The weights of actinides and fis- sion products pertinent to space dis- posal are shown for several reactor types in Table 8.1. The quantity of irradiated fuel sent to the repro- cessing plants is summarized in Table 8.2. Based upon these values, quantities of transuranics for dis- posal and the space flight require- ments are shown in Table 8.3 for several cases. TABLE 8.1 . Quantity of Actinides in Waste from Repro- cessing Various Reactor Fuels 10-Year Decay Reactor Type Total Actinides Gms/MT Fuel Charged 5,480 U rani urn + Gms/MT Fuel Charged Th Actinides Less U+Th Gms/MT Fuel Charged 706 Fu Curium Gms/MT el Charged 23 Fission Products, Gms/MT Fuel Charqed PWR-U 4,774 34,400 PWR-Pu 6,980 4,685 2,295 472 35,100 HTGR, with Recycle Make up 15,700 11,740 3,960 13 99,000 LMFBR-AI 6,360 4,327 2,033 12 38,100 LMFBR-GE 6,030 4,407 1,623 19 42,600 TABLE 8.2. [ 3 roj ected Q uanti ties of Uranium. i Plutonium and Thorium in the Spent Fuel Shipped to the Reprocessing PlantCO) Metric Tons/Year Year LWR-U LWR-Pu HTGR AIF0 GEF0 Total 1980 2,388.6 268.9 5.8 0.0 0.0 2,663.3 1990 6,959.2 278.5 874.5 285.1 11.9 8,409.1 2000 4,845.2 0.0 2 ,754.2 446.9 6 ,804.5 14,850.8 2005 3,842.9 0.0 3 ,604.6 351.4 13 ,555.3 21,354.2 8.5 BNWL-1900 TABLE 8.3 . Example Waste Transuranic Quantities and Mission Requirements for Space Disposal Basis: Waste held 10 years before space disposal Year of Spare Disposal 1990 2000 2010 Disposal Case Transuranics Shipped Per Year, Kq No. of Space Flights Per Yr No. of Shuttle Launches Per Yr Transuranics Shipped Per Year, Kq No. of Space Flights Per Yr No. of Shuttle Launches Per Yr Transuranics Shipped Per Year, Kg No. of Space Flights Per Yr No. of Shuttle Launches Per Yr Earth or Solar Orbit Case 1. Total Waste' a) -- 700 700 -- 2,500 2,500 -- 6,000 6,000 Case 2a (288 Kg Transuranics/mission) 2,315 8 8 9,600 33 33 26,200 91 91 Case 2b (447 Kg Trans uranic/miss ion) 2,315 5 5 9,600 21 21 26,200 59 59 Solar Escape Case 1 Total Waste K ' — 1,800 3,600 — 6,500 13,000 -- 15,600 31,200 Case 2a (113 Kg 2,315 24 48 9,600 85 170 26,200 232 464 Transuranics/mission) Case 2b (191 Kg 2,315 12 24 9,600 50 100 26,200 137 Transuranics/mission) 274 a. Based on NASA TMX-2911, Feasibility of Space Disposal of Radioactive Nuclear Waste , December 1973. This is a different base than that used in the rest of the table. These numbers are used here only to indicate the order of difference with total waste and transuranics. The actual quantities which might be considered for space disposal will vary with developments in the fuel cycle. Since parts of the waste stream are elements which have not been recovered because it was not eco- nomical to do so, the development of improved partitioning processes or the development of new uses for radio- isotopes could reduce the quantities which are presently considered to be waste. Quantities discussed herein should therefore be considered only as a very preliminary example. 8.1.1.2 Waste Capsule Encapsulation of the waste con- stituents is needed to prevent release of radioactive materials in event of a flight abort and to pre- vent harm to life. Containment is needed during surface transportation, during any phase of flight or abort, and during some lifetime in flight. Consideration must also be given for removal of decay heat, criticality control (for cases with actinides only), and for minimizing weight. Based upon these considerations, the design of the capsule for total waste is shown in Figure 8.2 and for 8.6 BNWL-1 900 TOTAL WASTE AS BOROSILICATE GLASS OUTER IMPACT SHELL (STAINLESS STEEL) HIELDING (DEPLETED URANIUM) NNER STORAGE CONTAINER (STAINLESS STEEL) A -A FIGURE 8.2 . Total Waste Capsule for Space Disposal actinides only in Figure 8.3. For total waste, the borosilicate glass containing the bulk waste oxides is encapsulated in an inner storage con- tainer of stainless steel. An outer container of stainless steel is pro- vided for impact resistance. De- pleted uranium was used as shielding material between the two containers. (However, it has been noted that the depleted uranium may be unsatisfac- tory at the high temperature i nvol ved. ) The conceptual capsule design for shipment of the transuranics (Figure 8.3) includes transuranic oxide particles (with some fission product contamination) in the form of small spheres approximately 3.3 milli- meters in diameter. These oxide spheres are encapsulated v in tungsten to provide a capsule with high temperature capability. A void space is left around the sphere to allow for helium buildup as the radioactive material decays. The tungsten is coated with an oxidation-resistant coating such as aluminum oxide so that the small capsule itself can withstand exposure to a hi gh- tempera- ture, oxygen-containing atmosphere. These micro-capsules are mixed in a matrix of lithium hydride for shielding and aluminum or copper for thermal conductivity. This matrix material is compacted in the form of a sphere. The sphere, which consists of small encapsulated particles in a lithium hydride-aluminum matrix, is surrounded with additional lithium hydride and tungsten for shielding and is then encapsulated in stainless steel for impact resistance. For protection in event of re- entry, an additional shell is added to the exterior of the waste capsule. A conceptual reentry shield, which surrounds the waste capsule, was de- signed to ensure stability and mini- mize the weight penalty and is shown in Figure 8.4. 8.1.1.3 Space Fl ight The potential space destinations considered are listed in Table 8.4, along with the incremental velocity (delta V) beyond low-earth-orbit vel oc i ty . The conceptual launch systems for several destinations are shown in Figure 8.5. The first step is propul si on of the launch vehicle (a manned space shuttle) into a low circular parking orbit of about 370 kilometers From this orbit, the upper stage or stages (Tugs) inject the waste pack- age into its final destination. For solar or high-earth orbits, a single 8.7 BNWL-1 900 TUNGSTEN CAPSULE FOR HIGH TEMP STABILITY VOID VOLUME FOR HELIUM BUI TRANSURANICS AND LiH 50 VOL% A TUNGSTEN SHIELDING STAINLESS STEEL ALUMINUM OXIDE COATING FOR OXIDATION RESISTANCE LITHIUM HYDRIDE- ALUMINUM MATRIX BORON PARTICLES 3.3 mm FIGURE 8. 3 Transuranic Waste Capsule for Space Disposal SS IMPACT SHELL LiH NEUTRON SHIELD TUNGSTEN GAMMA SHIELD REENTRY SHIELD TOTAL PACKAGE WEIGHT: 3270 GK FIGURE 8.4 . Re-entry Shield and Transuranic Disposal Package for Solar Escape Destination 8.8 BNWL-1900 TABLE 8.4 . Summary of Potential Space Destinations Delta V, km/ sec High-Earth Orbit 4.11 Solar Orbits Via: Single Burn Beyond Earth Escape 3.65 Circular Solar Orbit 4.11 Venus or Mars Swingby 4.11 Solar System Escape: Direct 8.75 Via Jupiter Swingby 7.01 Solar Impact: Direct 24.08 Via, Jupiter Swingby 7.62 (a) a. Delta V is the sum of the velocity increments required beyond that of reaching low earth orbit. tug which requires two fuel burns is used for each mission. For solar es- cape, two Tugs are required for each mission. Solar impact is not shown in Figure 8.5. Direct solar impact is not possible with these vehicles. However, solar impact using a swingby of Jupiter could be done using a sin- gle propulsion phase from the low- earth orbit. Similarly, swingbys of planets could be used for solar or- bits or solar system escape. How- ever, proper guidance and controls for all swingby missions would be most difficult to assure. The fact that missions utilizing swingby of other planets can be accomplished only at intervals of many months also makes utilization of such systems difficult. The NASA studies indicate that transport of waste constituents to the sun or other extraterrestrial bodies (e.g., the moon or other planets) are technically feasible. However, these destinations were con- sidered only briefly. (See Section 8.2.4.1 ) . 8.1.2 Systems Requirements System requirements for the space disposal concept are summarized in Figure 8.6. As illustrated in this figure, high-level liquid wastes would be conditioned, specifically by dens ificat ion and encapsulation as solids, transported to the launch site and flown to a final trajectory. 8.1.2.1 Waste Treatment and Handling The general system flow diagram shown in Figure 8.6 starts with high-level radioactive liquid waste from the reprocessing plant. The waste may be conditioned and stored retrievably for some period of time. Chemical adjustment (such as neu- tralization, which would not be expected in this case) and storage as liquid in tanks for a 5 to 10 year cooling time or encapsulation and interim storage are potential exam- ples of these steps. The waste may then be chemically separated into an actinide or transuranium stream and one or more fission product streams. For space disposal the separation into transuranic streams is most likely. A further separation of the curium from the transuranic stream could also be practiced to reduce the heat and radioactivity and weight of the package being disposed of in space. In addition, it is possible that long-lived fission products could be separated from the waste stream and included for space dispo- sal. The waste constituents to be transported to space would then be encapsulated into high integrity pack- ages in preparation for transporta- tion to the launch site. 8.9 BNWL-1900 SOLAR ORBIT 0.9 AU, AV~ 4.1 KM/SEC EARTH ORBIT HIGH EARTH ORBITS AV-4.1 KM/SEC FROM LOW EARTH ORBIT * I (EARTH l« \ SINGLE SHUTTLE LAUNCH TO 370 KM ORBIT 2 BURNS TO CIRCULAR SOLAR ORBIT (0.9 OR 1.1 AU) TIME BETWEEN BURNS ~6 MONTHS NOTE: AU IS THE MEAN DISTANCE BETWEEN THE EARTH AND THE SUN SINGLE SHUTTLE LAUNCH TO 370 KM ORBIT 2 BURNS TO -90, 000 KM CIRCULAR ORBIT TIME BETWEEN BURNS »20HR ALTITUDE ABOVE SYNCHRONOUS ORBIT SOLAR ESCAPE AV~ 8.75 KM/SEC /" x / \ \ 1 'Vv ! TWO SHUTTLE LAUNCHES TO 370 KM ORBIT 1 SHUTTLE CARRIES PAYLOAD AND EXPENDABLE TUG THE OTHER CARR I ES REUSABLE TUG 2 BURNS AT PERIGEE TIME BETWEEN BURNS «8HR FIGURE 8.5 . Launch Systems for Solar Orbit, High Earth Orbits and Solar Escape 8.1.2.2 Site Preparation It is assumed that existing Cape Kennedy facilities have the capacity and could be utilized for launch until about the year 2000. Some ad- ditions to the facilities would be re- quired in the way of waste package handling, launch pads, and other buildings. Shortly beyond the year 2000, major additional launch facili- ties would likely be required to accommodate the higher capacities. 8.1.2.3 Transportation to the Site Transportation from the repro- cessing plant to the launch site will be in special railroad cars, trucks, 8.10 BNWL-1 900 5o £5 V c >> •i- _Q CD (O OJ c +-> (O (/) 3 (O S- (/) OHO <4- > Q- •r- +->•(- +J (J Q C ro E -r- (O CU "O T- s. n u •i- CE -t-> 3 10 CJ-i— s- oj oi E _i +j .C S- in cr>4-> >»•<- X WIUI 00 LlJ a: => IS 8.11 BNWL-1 900 or possibly barges. Because of the stringent packaging requirements for space transportation, additional packaging requirements for the land transportation from the reprocessing and encapsulation site to the launch site are expected to be relatively minor. Some minor additional shield- ing and special transportation restraints will be needed. 8.1.2.4 Operation of the Site Operations at the site will con- sist primarily of unloading encapsu- lated material from land transporta- tion vehicles in modest hot cell type of facilities, and assembling it in the space vehicles. Because the m a t e ■ rial will be encapsulated to survive launch pad accidents or re-entry from outer space and still maintain a radiation exposure of less than 1 rem at 1 meter from the surface, remote handling facilities will be minimal. Prior assembly, positioning, and fueling of the rockets will have been done. Final countdown and typical launch preparation procedures will be required. The manned shuttle and its appurtenances will be launched into a low-earth orbit, the payload bay will be opened and the tug with the waste capsule will be launched and pro- grammed toward the final destination. The shuttle will then be returned to earth for re-use. 8.1.2.5 Final Monitoring Final monitoring would involve position monitoring of each tug mis- sion to assure proper final destina- tion. This would require up to a few years for each mission. In addition, continuous monitoring would be required on the contamination of space from failed capsules. Two tra- jectories seem most likely to assure true final disposal: impact with the sun (which is not likely with present vehicles) and escape from the solar system. Should there ever be any return of waste constituents from these two trajectories, the dilution of the waste constituents which might intercept the earth is expected to be so great that it would be undetect- able by present day techniques. How- ever, even with these disposal loca- tions, continuous monitoring of the high-earth atmosphere for the next few hundred years may be desirable. 8.2 TECHNICAL FEASIBILITY Key considerations in the tech- nical feasibility of extraterrestrial disposal are: providing a reliable space flight system to assure that the waste will reach and remain at its proper destination; providing a high-integrity capsule to assure non failure during required residence times in space and to assure minimum spread of radionuclides in event of an abort; providing adequate provi- sions for cooling, shielding, and critical ity control; and minimizing weight to keep the costs practical. Most of these considerations have been discussed briefly in the pre- ceding descriptions but are covered in more detail in this section as they relate to technical feasibility. 8.2.1 Capsule Design Criteria The basic design criteria used in the study for the above-described design factors are described in the f ol 1 owi ng secti ons . BNWL-1900 Radiation Shielding A general criterion for radiation shielding was established that the capsule at all times must have radia- tion levels less than 1 rem per hr at 1 meter from the surface. Higher dose rates than the 1 rem at 1 meter were considered in some cases to de- termine the effect of higher dose rates on costs. For shielding calculations used, it was assumed that the capsule shape is retained after impact following a postulated abort. For example, if a spherical capsule containing actinide waste impacts the earth, it is as- sumed that the capsule will retain its spherical shape and the resul- tant radiation level at 1 meter would not increase. In fact, it is likely that under some circumstances the capsule will be grossly deformed and actual dose rates could be some- what different. Shielding calcula- tions on distorted capsules are be- yond the scope of this study but should be examined before space dis- posal is used. Temperature Temperatures within the capsule, from the center-line to the edge, should be sufficiently low that mate- rials in the capsule will not decom- pose or react excessively with each other to result in no adverse effects on the capsule integrity. This sta- bility is necessary in outer space, in re-entry circumstances, in a vacuum in the shuttle bay, and for accidental burial in earth or resi- dence in a fire. Consideration was given to estab- lishing conditions such that the cap- sule be sufficiently cool that it can be touched if it lands accidentally in an unknown location. However, it was generally agreed that the poten- tial for severe harm to people touch- ing a thermally hot capsule would be very low. Hence, this criterion was not used . Containment • Ground Transportation A typical accident during ground transportation should not result in a serious hazard because a capsule designed to withstand flight abort and re-entry from outer space should withstand virtually any conceivable conditions during an earth accident. While the criterion of 1 rem per hour at 1 meter from the surface is consis- tent with the Department of Transpor- tation regulation^ ' for exclusive use vehicles of 1000 millirem per hour at 3 feet from the external pack- age surface, some minor vehicle design features may be necessary to meet the requirements for radiation levels external to the vehicle and in positions of the vehicle which are occupied by persons. • Flight Aborts Launch Pad . The capsule should be able to successfully resist overpres- sure, high velocity fragments, fire- ball heat, impact, and residual fire of a launch pad accident. Pre-orbi ta 1 Abort . The capsule should be able to survive intact an accident during launch and prior to achieving orbital velocity. In gen- eral, a capsule which will withstand abort from orbit will withstand any condition that can occur between launch and orbit. Abort from Orbit . The capsule should be able to survive re-entry 8.13 BNWL-1900 at a velocity of 11 km/sec with sub- (12) sequent impact on earth. ' The impact conditions assume a hard gran- ite surface at terminal velocity of the order of 300 meters per second. Elevated temperature must be assumed as a result of re-entry. Obviously, rough impact surfaces may be encountered . Following impact, the capsule should be able to withstand burial in the deepest ocean trenches or burial in a variety of soils for reasonable time periods. A capsule landing in relatively soft sand is likely to be buried to depths beyond 6 meters. Hence, it may be very well insulated. If a capsule lands in the ocean, it is generally assumed that it will be recovered some day. However, there are real probabilities that it can never be found. Hence the ability to withstand seawater and the seabed environment for long time periods is highly desirable. L ong-Time Residence in Space The necessary lifetime of the capsule during flight depends largely on the destination. If the destina- tion is direct solar impact, then a containment lifetime of only a few months and a shielding lifetime only long enough to be removed from poten- tial exposure to man may be all that is necessary. If the destination is a storage type of orbit, then a cap- sule lifetime of hundreds or thou- sands of years may be needed. If the destination is solar escape, a life- time of millions of years could be desired if no contamination of the universe is allowed, or tens of years if contamination of the universe is of no major concern. Thus capsule lifetime needs can vary greatly depending on the circumstances. At the present time, capsule design assuming a lifetime of a few hundred years may be the best that can be achi eved . Cri tical i ty Criticality should not be reached under any foreseeable circumstances. These include loss of neutron-absorb- ing materials and dissolution of the capsule in water. Such considera- tions must take into account the fact that many of the actinides have very low critical masses. We i g ht Limitations of current rocketry in carrying payloads to certain desti- nations were used as the basis for sizing maximum capsule weights. 8.2.2 Capsule Design Details Conceptual capsule designs for cases involving space disposal of total waste and transuranics only are presented below. Capsule Design - Total Waste - Ca_se_li a) A conceptual design was developed (13) by NASA Lewis for shipment of a. Shielding calculations were based on fission products only. 8.14 BNWL-1900 total waste. The design was based on the incorporation of total waste oxides in mixtures of 30 mole% in boros i 1 i ca te glass. Consideration was given early in the analysis for all cases to reduc- ing the shielding weight required. Possible methods considered were shadow shielding and reusable shielding. With shadow shielding, areas in line with personnel or radiation sensitive components are more heavily shielded than those areas in which radiation would not have harmful effects. As an example, a waste con- tainer might have heavy shielding on the side next to the space vehicle and light shielding on the side exposed to space. The capsule would be transported on earth in shielded containers and installed in the space vehicle with special equipment and procedures. Such a system would be useful if successful flight could be assured. However, the potential haz- ards of such a container in event of unsuccessful flight indicated the need for extensive analysis beyond the scope of this study and was not considered further. With disposable shielding, the package would be heavily shielded until some point early in the flight path where the heavy shielding would be separated. Only lighter shielding would be accelerated to the velocity necessary for disposal. For example, full shielding might be carried by a space shuttle to low earth orbit. The tug or tugs which accelerate the container to a subsequent trajectory would carry only a lightly shielded package. The heavy shielding would remain with the shuttle and be returned to earth. The heavy shield- ing could be reusable. This system could require much less total energy for disposal. However, this concept was not considered further for the same reasons given for shadow shielding. Calculations were performed by the NASA staff on containment vessels and shielding for total waste which would provide dose rates of 1 rem/hr, 10 rem/hr, and 500 rem/hr at 3 meters from the container centerline. In these calculations it was assumed that only the fission products con- tributed to the radiation dose. Although some differences would exist with ac tin ides included, the differ- ences should be unimportant. The NASA analysis of this capsule is included as Appendix B. The container design was illus- trated in Figure 8.2. The ability of this container to provide the impact resistance necessary for containment during any strenuous abort conditions is untested. The NASA analysis concluded that the space transportation cost alone for disposal of the above containers shielded to 1 rem per hour at 3 meters from the surface would add to the cost of electricity at the bus bar approximately 4 mills per kW-hr for earth escape and 28 mills per kW-hr for solar escape. Increasing the dose rate at 3 meters from the outer surface of the package from 1 to 500 rem/hr would result in a factor of 3 reduction in cost. 8.15 BNWL-1 900 The above costs were high and space flight requirements were extreme. The cost of providing greater package integrity than the simple cylinder would add still fur- ther to the complexity and cost. For these reasons the principal effort was applied to Case 2, transuranics in waste only. Capsule Design - Transuranics only, with Fission Product Contamination - Case 2 Extensive developmental work on the use of isotopes in space and on aerospace nuclear safety is appli- cable to this case. A prel imini nary capsule design for shipment of the transuranics in waste was shown in Figure 8.3. Analysis and testing of containers of this type have been done as a part of pro- grams for the utilization of radioiso- topes in space and aircraft reactor (14-31 ) programs. ' An analysis of a capsule design was made by the staff at NASA Lewis. The following (up to "Critical ity Considerations") is ex- tracted completely from their re- (12) port. ' Slight modifications have been made to conform to the format of this report. Encapsulation of the transuranics with inclusion of small percentages (0.1 to 1%) of the fission products was studied in some detail to pro- vide a conceptual design for deter- mining the feasibility of the approach . There are some differences in waste content as defined in the NASA studies and in the balance of this report. However, the differences are minor and have no effect on the deter- mination of feasibility of the process . The amount of transuranic waste relative to the matrix was varied parametr i ca 1 1 y to obtain the maximum amount of transuranic waste products in a payload without exceeding the limiting temperature in the (32) matrix. The optimum transuranic waste content was approximately 8 to 10 volume percent of the matrix to maintain the temperature of the ma- trix below the prescribed temperature limit of 860°K. The temperature limit was established primarily by the elevated temperature character- istics of lithium hydride. • Radiation Shielding Shielding for the transuranic waste is required to reduce the exter- nal dose rate to acceptable levels. These levels are based upon accept- able levels for handling, acceptable levels for transporting in a manned shuttle, or accidental exposure levels to the general public. The value of 1 rem/hr at 1 meter from the external surface of the package was chosen as the base point (based on 10 CFR 71^ 33 ^ for transporting of radioactive waste) for accidental exposure to the general public. Ef- fects of extending the dose rate to 100 rem/hr were examined for this study. The shielding for gamma, neutron, beta and alpha radiation was examined (32) for a weight optimization study. ' Based on these calculations, a single layer of tungsten and a layer of LiH were used for capsule shielding. The 8.16 BNWL-1 900 calculations showed that the main source to be shielded was the gamma radiation from the fission products and not the neutron or the alpha ra- diation from the actinides. To minimize the shield weight, which is the major portion of the total payload weight, a spherical geo- metry was chosen for the waste cap- sule. The layer adjacent to the matrix material containing the acti- nides is composed of high-density gamma shielding material (tungsten). For safety purposes, in the event of a break in the outer vessels, a layer of stainless steel was added to the outside of the tungsten to prevent oxidation of the tungsten. The next and last shielding layer is the neu- tron shielding, or lithium hydride as selected for the study. The layers of material external to these were not considered as part of the shield- i ng analyses . • Impact Protection The primary impact protection for the transuranic waste package is a spherical shell on the outside of the lithium hydride. The spherical shell selected, based on experiments and analyses, was 1.58 centimeters (5/8 inches) of 304 stainless steel. This was backed up by an additional 0.95 centimeter shell of 304 stain- less steel between the lithium hy- dride and the tungsten shields. To prevent any free hydrogen that could be present in the lithium hy- dride from reaching the outer stain- less steel containment shell, a 0.127-millimeter (0.005 inch) layer of tungsten was assumed to be deposi- ted on the inside of the stainless steel shell. This should prevent hy- drogen from diffusing into the stain- less steel shell. • Re-entry Shell For re-entry protection, a re- entry shell must be added to the exterior of the transuranic waste capsule. The stable configuration selected to minimize the weight penalty was shown in Figure 8.4. In re-entry, the heating rates on the package vary depending on velo- city, angle and atmospheric density. At low heating rates the convective heat transfer away from the shell is most important, for which materials such as graphite function well. Appollo re-entries are an example of convective heating rates. For high velocities such as can be encountered in planetary entry following earth- escape velocities at steep angles, the radiative heating rate dominates and a reflective type of re-entry shell material is required. One such material that has high reflective capability with multiple reflective sites for scattering and reflecting heat is a composite made of quartz fibers woven into a mat, similar to fiberglass, with a silica binder added. This material proposed by the staff at NASA/Ames Research Center results in a very good reflec- tive barrier for the re-entry shell. Some of the thermophysical proper- ties are: 3 Density 2.5 g/ cm Specific Heat 1.15 Joule/g-°K Thermal Conductivity 2.5 x 10" 2 Joule/sec-cm-°K Thermal Expansion Coefficient 5.6 x 10" 7 /°K 8.17 BNWL-1 900 This layer of composite silica fibers must be backed with a thin silver film followed by graphite to act as the re-entry heat shield in the slower re-entry mode. A pictorial description of these layers is pre- sented in Figure 8.7. The insulation is added primarily to protect the stainless steel containment vessel from heating during the re-entry or in the event of a launch pad fire. On the back side of the re-entry shield, the thicknesses are reduced and the insulation has been removed to allow for the waste heat from the actinides and fission products to be conducted and radiated away. The re-entry shell weights represent approximately 13% of the total pack- age weight. • Single and Multiple Re-entry Packages While the nuclear waste package is in the shuttle bay, the total package after having been precooled before launch, slowly increases in temperature. After separation from the low-earth orbiting shuttle, the waste package will come to an equili- brium temperature, based on the in- ternal heat source (transuranics and fission products), the thermal con- ductivity through the layers of mate- rial and the heat sink temperature in space. Steady-state temperature calcula- tions were conducted for various sizes of waste packages. Using the temperature limit for the matrix material, the minimum number of pack- ages per total payload was deter- mined. MATERIAL STEEL SILICA FIBER COMPOSITE — SILVER GRAPHITE FIBER COMPOSITE THICKNESS 0.1 CM 4.0 CM 1.0 CM NSULATION 2.0 CM FUNCTION ATTACHMENT REFLECT RADIATIVE HEAT DURING REENTRY 0.01 CM MIRROR-REFLECT ABSORB CONVECTIVE HEAT DURING REENTRY INSULATE FROM REENTRY HEAT FIGURE 8.7 . Typical Front Re-entry Shield Composite for Radioactive Waste Package 8.18 BNWL-1 900 The temperature limit on the ma- trix would be exceeded for the total payload in a single sphere for the high earth orbit package with 0.1% fission products. Therefore this pay- load would have to be divided into two or more packages. For other Case 2 waste conditions, the design can use a single package per payload unless a safety assessment indicates multiple packaging is an improvement. • Overall Package Configuration An overall configuration was established using the previously-de- rived re-entry shield, impact protec- tion and shielding against radio- activity. For high earth orbits the payload with transuranics and 0.1% fission products was divided into three equal packages, each with its own impact and re-entry protection. The largest single package has a diameter of 2.8 meters, with a 1.37-meter diameter stainless steel sphere containing the nuclear shield- ing and the waste-containing matrix. This package weighing 8400 kilograms contains 384 kilograms of actinides plus 134 kilograms of fission prod- ucts. Other data for all the. acti- nide packages considered are found in Table 8. 5. • We i ght The payload capabilities, assuming a shuttle as the basic launch vehicle, used for the basis of the package de- signs were 8480 kilograms for high earth orbit or solar orbit missions and 3270 kilograms for the solar sys- tem escape missions. These payloads may be designed as single waste pack- ages or as multiple waste packages, each with its own re-entry shell. • Total Packaging Weight Ratio Versus External Dose Rate The above discussion and design has been primarily for an external dose rate of 1 rem/hr at 1 meter from the surface of the impact shell. If this dose rate could be increased (subject to safety assessment and acceptance), the amount of waste con- stituents could be increased per launch. The effect of allowable dose rate on the packaging weight ratio (total weight: waste material weight) is shown in Figure 8.8. The choice of destination does not substantially change the pack- age weight ratio. The greatest effect on the weight ratio is the reduced shielding for the higher dose rates. The predominant gain is obtained by raising the allowable dose rate to 10 rem/hr. Further in- creases to higher allowable dose rates are less effective in increas- ing the package ratio. Capsule Design - Transuranics Only, with Fission Product Con - tamination, with Most of Curium Removed - Case 3 The amount of curium in the trans- uranic waste from several different reactor types is summarized in Table 8.6. Up to 93% of the heat and 87% of the radioactivity of the acti- nides is associated with the rela- tively short-lived curium from LWR's. However, curium content in LMFBR waste transuranics is much less than in LWR waste, and the curium content in HTGR waste is less than 6% of the total . The removal of curium neutron ra- diation from the package permits a reduction in the amount of lithium 8.19 BNWL-1 900 o OO OH i — i — CO «=fr Lf) CM ooi— oooomo criinr-.cn i — 0«3"i — Or — O CM OCSICO^ - to i— i— co +-> rO j* i— <_> rO ro to a. o o. c ui o ■■- i=i ra +-> o rO -r- o c >s- S- 3 rO 00 E c E *> 3 S- co a; o cc CO err o o o i — CO r--.i — oooomo r-O'Jr-Or-ON r--i— ooooino r-O^r-Or-CN CM <" CO «3" CT) in CM 00 O CM CO «d" in CO o CO o s* oi in co to r~. 00 O cm oi co to en r— cOr-ooooLno CM O «3" i — Or— OCM i — CM i— Lf> r-. «* >3- >s- to CTl CO >3- in oo "d" in to 00 COr— OOOOLDO CM O «3" i — O ' — OCM r — CMCMOO CO to r- to onomNi/ioin r- . — «d-i — tococoo cm i — <3- in i — **■ in CO r— Oi — totnooooun Mji i— ^r r-» en cm CM i — «3" tO i — i — tO O o oo Oto^-O">rcocnin OOlfONO. — >=3- r- oo co in r— cm >3- CM i— oeninoinoocMCO o "=i- r— to to co en oo i — co in r — o ^r o^-"^-cocMcr>om OOOCOCOCMCMCOin ■^•coi — oi — ^-into CO l— I— CO i— o^-CMCOininoo OCOCMCOCTlinCOCO <* tO OW*NO 00 i — i — CM CM to to to O O CM r— X X CTl to t-«. 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S- ro 1 — 1 00 aj c 3 -c j^ x: ^ _C ^Z -£Z • jz x: JC 4- cu CU CU CU cu CU cu cu CU E CU QJ E CO I— •1— -Q OJ O t— r— I— 1— 1 — 1 — 1 — Ohr- I— O 0£ 3= 3 3 3 3 3 3 3 CU O X) XI <- OO Ll_ E E to s_ E E cu UJ 3 O 4- 3 3 x: o *« Z a o Z S 1— ro 8.20 BNWL-1 900 30 < O = 20 1- =5 o LiJ 0.7 Pu-240 0.9 96.4 <0.1 Pu-241 0.14 0.26 0.5 Am-241 2.5 71.4 <0.1 Cm- 244 4.2 14.2 0.3 a. Each estimate is based on most severe conditions. b. Based on 30% Pu-240 With the lithium-aluminum treated as voids, the neutron multiplication factor, koo for the material described in Table 8.8 is 0.3. Since lithium is a neutron absorber, koo for the ma- terial would actually be less than 0.3. Consequently, an infinite amount of this material would be well subcri ti cal . The nuclear criticality safety analysis has been limited to the acti- nide waste package as designed. Pos- sible accident conditions that could affect the nuclear criticality of the package during the fabrication, launch, or post launch were not con- sidered for this study. 8.2.3 Encapsulation Processes Proper encapsulation of the waste for space disposal is highly impor- tant and complex. Processes for achieving the encapsulation were re- viewed, and summarized in this secti on . The encapsulation process for con- tainers for total waste, Case 1, con- sists primarily of welding existing containers inside secondary con- tainers for shielding and impact re- sistance. For reasons given earlier, this case was not considered in detail; hence, a conceptual encapsu- lation process was not developed for this container. 8.24 BNWL-1900 The encapsulation process contem- plated for transuranic waste is rela- tively complex to minimize weight and to provide a high degree of contain- ment reliability. Encapsulation pro- cesses have been developed exten- sively for radioactive isotopic heat sources. " Steps in the trans- uranic oxide encapsulation process are shown in Figure 8.9. The process steps illustrated are "state of the art" and have been used for the prepa- ration of cermets of plutonium oxide, promethium oxide, and other radio- active material s . The steps of the process are de- scribed briefly below: 1 . Purification or Treatment . This step might be the last one in the separation process or the first step in the encapsulation process. It consists of the processing neces- sary to ensure stability of the final product and container. It includes such activities as vacuum outgassing of the powdered oxide at high tempera- ture to remove impurities such as moisture, trace chemicals, etc. 2. Powder consolidation . Trans- uranic oxide powders are pressed to small billets. This step provides an initial densifi cation and puts the material into a form from which it can be crushed for sizing before final form making. 3. Crushing . The powder billets from Step 2 are crushed in much the same way that large rocks are broken up into gravel. This step further refines particles for sizing into the desired range. 4. Sizing . The particles from Step 3 are sized by passing them through different sizes of sieves. 5. Si nteri ng . The particles from Step 4 are sintered to convert them to the final solid dense material. If desired, they may be passed through a plasma arc to make them spherical and further increase their densi ty . 6. Recycl e . Particles which have not attained proper size in the ini- tial sizing are recycled back through the consolidation operation. 7 . Coating Material Treatment . Materials to be used for coating are purified as necessary. 8. Sacrificial Coating of Magne- sium . A thin coating of magnesium is applied to the actinide oxide parti- cles by sputtering. This coating, when later removed, provides for a void in each final coated particle. Particles are placed on a tray and exposed to sputtered materials. 9 . First Phase Tungsten Coating by Thermal Decomposition . Particles are thinly coated by thermal decompo- sition of tungsten hexafluoride in a fluidized bed. This coating provides a base for the subsequent coating after removal of the sacrificial magnesi urn. 1 0. Sacrificial Material Removal . The magnesium coating is removed by placing the particles in a high tem- perature vacuum furnace to evaporate the magnesium. This process leaves a void inside of the initial tungsten shell . 11. Decontami nation . Coated par- ticles are decontaminated by chemical treatment which may consist of etch- ing in nitric acid, neutralizing in a stop bath, rinsing and subsequently drying in vacuum. Decontamination is required to effect final tungsten coating. 8.25 BNWL-1 900 1 r-dXIXD 2 [-0 3 4 Ki>0 1 5 KD0 TRANSURANIC OXIDE POWDER POWDER CONSOLIDATION CRUSHING — » . SIZING SINTERING ' t 6 - RECYCLE KD |-Ki>®0 , -000 KD 7 ' , 8 FIRST PHASE TUNGSTEN COATING THERMAL 10 n COATING MATERIAL TREATMENT SACRIFICIAL MAGNESIUM COATING SACRIFICIAL MATERIAL REMOVAL DECONTAMINATION DECOMPOSITION DICONTAMINATION 12 r0K^>0 13 SECOND PHASE TUNGSTEN COATING r-0 ALUMINUM OXIDE COATING KD LiH TREATMENT TREATMENT If, r-00 r00>0 LiH-AI ACT. OXIDE MIXING COMPACTING r0000 r000 cuniNG FORMING * MACHINING NOTE: • RECTANGULAR BOXES INDICATE PROCESS STEPS. CIRCULAR BALLOONS BETWEEN BOXES SHOW INSPECTION STEPS WHICH MIGHT BE CARRIED OUT DURING PROCESSING. L* CLEANING p00<^000000 NG -* ASSEMBLY 23 M MAl^i WELDING SHIPMENT INSPECTION OPERATION DESCRIPTION 1. DIMENSION MEASUREMENTS 2. DYE PENETRANT 3. IMPURITY ANALYSIS 4. DENSITY DETERMINATION 5. He LEAK CHECK 6. METALLOGRAPHY 7. I SOTOP I C ANALYSIS 8. X-RAY DIFFRACTION 9. RADIOGRAPHY 10. ULTRASONIC TESTING 11. DECON. SMEAR CHECK 12. RADIATION MAPPING 13. GAMMA SPECTROMETRY 14. CALORIMETRY FIGURE 8. 9 Steps in Transuranic Oxide Encapsulation Process for Space Disposal 1 2 . Second Phase Tungsten Coating . The tungsten shell is finished by pro- viding a heavy coating of tungsten by thermal decomposition of tungsten hexafluoride. 13. Aluminum Oxide Coating . A pro- tective coating of aluminum oxide is provided by flame spraying. 1 4 . Lithium Hydride Treatment . Lithium hydride is treated by chem- ical preparation and particle sizing. 15. Aluminum Treatment . Aluminum powders are treated by high tempera- ture vacuum outgassing to remove impurities. 1 6 . Lithium Hydr i de-Al umi num- Actinide Oxide Particle Mixing . The coated actinide particles are mixed with particles of lithium hydride and aluminum powder to provide a uni- form dispersion. 17. Compacting . The powders are compacted into a solid mass by press- ing at high pressure and subsequently sintering. 18. Cutting . The outer capsule for shielding and containment is formed by conventional means. The first step is cutting to size. 8.26 BNWL-1 900 19. Formi ng . Forming of capsule halves is accomplished by conven- tional pressing. 20. Machining . The capsule is machined to the fine tolerances necessary for adequate welding. 21. Cleaning . The components of the capsule are cleaned chemically. 22. Assembly . The lithium hydride- aluminum-transuranium oxide is assem- bled in the capsule halves. 23. Wei di ng . Welding together of the two capsule halves is accom- plished by tungsten inert gas welding, 24. Shi pment . The completed con- tainer is packaged for shipment. 8.2.4 Ground Transportation Assuming that partitioning and en- capsulation will take place at the reprocessing plant site, transporta- tion of waste to the space transpor- tation site should not present unu- sual problems. A capsule which is encapsulated for space flight should be suitable for earthbound transpor- tation. If shadow shielding were used for space flight, additional shielding might be required for earth transportation . For more unusual space flight sys- tems such as particle acceleration, additional analysis of earth transpor- tation requirements may be required. 8.2.5 Space Flight Consideration was given to several different types of space flight. Con- cepts considered, some only briefly, include: • Utilization of the energy in the waste for propulsion. • Solar sails. • Nuclear and ion propulsion. • Acceleration of waste particles electrically from an orbiting pi atf orm. • State-of-the art vehicles (this in- cludes the space shuttle and space tug, which are advanced vehicles but will use existing technology). The latter four variations appear to have potential for increasing the payload or providing increased pro- pulsion velocity capability. There- fore, the differences in their use versus state-of-the-art vehicles are primarily in degree, as opposed to changing the overall conclusions from the study. (However, it is possible that other trajectories for final des- tinations-^. g., disposal into the sun--could become more practical with development of such concepts.) The non-state-of-the-art systems are much less developed and would require stu- dies beyond the scope of this study to evaluate them fairly and uniformly, Therefore, while some warrant further study, it was concluded that the analysis using state-of-the-art vehi- cles provides a base point for evalua- tion of f easi bi 1 i ty . • Utilization of the Energy in Waste for Propulsion NASA/Ames has analyzed the poten- tial for use of waste-generated heat for propulsion. The analysis is in- cluded in Appendix C. This approach requires high heat density in the waste materials and appears promising only if short-lived isotopes such as stronti um-90 are included in the waste package. The extra energy in the waste would, in turn, be compen- sated for by the increased shielding (and payload) requirements. The con- clusion of that preliminary study was 8.27 BNWL-1 900 that the approach warrants further considerations . • Solar Sail A solar sail vehicle utilizes a relatively large surface that is pushed upon by the sun's photon ra- diation, Figure 8.10. It can be utilized to propel a vehicle into the sun as well as away from the sun. To propel a vehicle into the sun, the sail must be placed at an angle such that the force of the sail tends to reduce the vehicle's solar orbital velocity. The solar force on the sail at the earth's distance from the -4 2 sun is on the order of 10 dynes/cm . At speeds close to the earth's or- bital speed, which is the maximum the vehicle would have, the drag is about -8 2 10" dynes/cm . This drag assumes a 3 3 particle density of 10 particles/cm . Solar sail propulsion was studied by a group of students at Massachu- setts Institute of Technology^ ' as a possible means of reducing the cost of space disposal. The conclusion was that: "The solar sail-powered mission offers such advantages over present propulsion schemes that it may be termed the only way to sunfall. It is basically unsophisticated, and no breakthrough in the state-of-the- art is necessary. It offers the in- herent advantages of the direct sun- fall mission: open launch window, low navigational requirements, defi- nitive payload rate, and low time of flight. There are some uncertain- ties, however. Perhaps the main one is the final phase of flight, where solar radiation, even if simply re- duced to infrared band, is likely to melt down what is left of the space- craft. Obviously, the velocity con- ditions at the moment must be such that free fall will result in sun impact. " The solar sail vehicle could pos- sibly provide a realistic means of achieving direct solar impact. While there are a number of unanswered tech- nical questions (such as the life of the sail), it appears to be a techni- cal possi bi 1 i ty . • Nuclear and Ion Propulsion High specific impulse vehicles such as nuclear rockets and ion thrus- ters have been mentioned in some stu- dies, and the possibility of using such vehicles was considered briefly here. Use of such vehicles could ul- timately make significant improve- ments in the economics of space di sposal . • Acceleration of Waste Particles from an Orbiting Platform The brief study on waste particle acceleration is included in Appen- dix 8.D. The particle acceleration method contemplates transporting waste from earth to a permanently or- biting platform. On the orbiting platform the waste would be con- verted to the form of small charged particles. These particles would then be electrostatically accelerated to the desired velocity. One of the principal advantages of this system is the savings in energy requirements made possible by elimination of a conta i ner . The remainder of this section will treat the technical feasibility of state-of-the-art space flight, which was the primary emphasis for this study . 8.28 BNWL-1900 SOLAR SAIL, APPROXIMATELY 1 KILOMETER IN DIAMETER RING STRUCTURE ACTIVE STABILIZATION PACKAGE TUG WITH WASTE "RIP" -PROPAGATION ARRESTING REINFORCEMENTS ORBIT RADIUS, r SUN I .-'SHIP LIGHT iV^ SHIP -+-Zk X SAIL SHROUD SAIL LINE The sail angle, 9, can be adjusted so the light force, F s , on the sail and the sun's gravity force, Fg , inter- act to either increase or decrease the orbital velocity of the ship. A decrease in orbital velocity will cause fall into the sun. FIGURE 8.10 Illustration of Solar Sail Vehicle • State-of-the-Art Vehicles State-of-the-art space flight includes vehicles such as the space shuttle and the space tug. These vehicles are not yet built, but they can be built with current engineering knowledge. The space shuttle is cur- rently in the development stage. The analysis below (through the section "Potential Space Transporta- tion Vehicle Performance and Cost") was performed by the staff at NASA (12) Lewis v ' and is included here nearly verbatim. Minor modifications were made to conform to the format of this report . 8.29 BNWL-1900 The space destinations considered in this study include Earth orbits, solar orbits, solar system escape and solar impact. The space destinations are discussed in the order of increas- ing energy requirement. All launches are assumed to occur from the Eastern Test Range (ETR) in an easterly di recti on . It is assumed that the launch vehicle will first launch into a low circular Earth parking orbit, al- though this is not always necessary nor advantageous. After parking in this orbit, the launch vehicle upper stage or stages will inject the waste package to its final destination. Mission energy is characterized by the mission Delta-V requirement which is defined as the sum of all the velo- city increments that the launch vehi- cle has to provide after reaching low Earth orbit. In some cases the launch system alone can place or in- ject the waste package to its final destination. In other cases the waste package, after separation from the launch vehicle, will require sub- sequent trajectory (midcourse) cor- rections or propulsion upon reaching its destination. In these cases the waste package is part of an active spacecraft requiring propulsion sys- tems and the addition of guidance, control and communications systems. 8.2.5.1 Potential Destinations • High Earth Orbit To achieve high circular final Earth orbits, starting from a low circular parking orbit, two propul- sion maneuvers are required. The first is made in the parking orbit and places the payload on an ellip- tical transfer orbit. After coasting along the transfer orbit to the de- sired final altitude, the second ma- neuver is made to circularize the final orbit. Both of these maneuvers are expected to be performed by the launch vehicle upper stage. For high Earth orbits, in the event of a propulsion failure after reaching the parking orbit but prior to final placement, corrective action can be taken. The resulting orbit would have an adequate lifetime (sev- eral months) so that a second launch could be made to rendezvous with the waste package. The waste package would then either be sent to its final orbit or be retrieved. This discussion also applies to other des- tinations if the propulsion failure occurs prior to reaching Earth escape vel oc i ty . For the disposal of nuclear waste it is not clear as yet which orbit altitudes are acceptable. Orbit life- time is a primary factor. Orbit life- times of a million years or longer may be required if long-lived wastes are to be disposed of in space. At reasonably high orbit altitudes, above several thousand kilometers, at- mospheric drag is negligible, but other perturbations such as solar pressure and solar, lunar, and plane- tary gravitational perturbations must be considered. Orbits near the Moon must be avoided to minimize lunar per- turbations. High traffic regions or orbits important from a science or applications point of view (such as synchronous orbit altitude and some lower altitudes) should not be chosen. 8.30 BNWL-1 900 Therefore, probably the best choice for Earth orbits would be those or- bits lying between synchronous orbit altitude and the Moon. However, such orbits have the highest Delta-V re- quirement of the high Earth orbits, on the order of 4.1 km/sec. Advantages of High-Earth Orbit • The Delta-V required is relatively low in comparison to some of the other destinations. • The waste package could conceiv- ably be retrieved at a later date either to recover the waste material or to remedy some unforeseen problem. • There is a launch opportunity any day. • The waste package could be passive, requiring no guidance or propulsion capability since the second propul- sive burn is performed by the launch vehicle upper stage. Disadvantages of High-Earth Orbit • The stability of high Earth orbits and hence orbit lifetime over a long period of time (on the order of a mil- lion years) is not well understood. To date the complexity of the multi- perturbation problem precludes rigor- ously verifying the stability of these orbits over these long-time periods . • There is no assurance of the integ- rity of the relatively hot waste pack- age when exposed to the space environ- ment over these long periods of time. • Eventually, the waste packages will be randomly located within a belt around the Earth. Gravitational perturbations cause orbits of the waste packages to vary slightly, with time, thus producing variations in orbits in this region. This region would be regularly penetrated by fu- ture planetary spacecraft. However, because of the wide spacing between waste packages at such high placement altitudes, the probability of a colli- sion would be extremely remote. Since neither orbit stability nor waste package integrity problems are well understood, high Earth orbits cannot be considered a permanent dis- posal site for times on the order of a million years. Unless further stu- dies can resolve these problems, Earth orbits should only be consid- ered for hundreds or a few thousand years, with potential need for re- quiring further action at a later date . • Solar Orbits The solar orbits considered in this study are those achievable with relatively low Delta-Vs. These in- clude (1) solar orbits achievable by injecting the waste package to Earth escape velocity or slightly beyond, (2) circular solar orbits slightly inside or outside the Earth's orbit about the Sun achieved by additional propulsion after escaping the Earth and (3) solar orbits achievable by swinging by Mars or Venus. Earth Escape . The simplest method for achieving a solar orbit is to have the launch system inject the waste package to Earth escape energy. This can be done with a single propul' sive burn from Earth orbit with a Delta-V of approximately 3.23 km/sec. The waste package would then be sepa- rated from the launch vehicle and after escaping the Earth's gravita- tional field would be in an orbit about the Sun. The waste package 8.31 BNWL-1900 would be in essentially the Earth's orbit about the Sun but in a differ- ent angular position. With the waste package in this orbit, there is a high probability of the waste package re-encountering the Earth at some future time. Due to in- herent limitations on injection accu- racy and long-term gravitational per- turbation ef f ects--pr i nci pal ly from the Earth--the waste package cannot be maintained at a fixed position from the Earth. As a result of these effects the waste package tends to drift with respect to the Earth, and preliminary calculations indicate a high probability of re-encountering the Earth within a few thousand years or less. A better approach would be to pro- vide somewhat more Delta-V than re- quired for Earth escape (an addi- tional Delta-V on the order of 0.42 km/sec for a total incremental Delta-V of 3.65 km/sec), so that the waste package would be in a slightly elliptic solar orbit with a small in- clination to the ecliptic plane (plane of the Earth's orbit about the Sun). Initially, the orbit of the waste package would intersect the Earth's orbit at only one point. Fur- thermore, planetary gravitational effects tend to precess or vary the orbit of the waste package with res- pect to the Earth's orbit making an encounter even less likely. Prelimi- nary calculations indicate that such is the case at least for a few thou- sand years. Advantages of Elliptical Solar Orbits • Of all the mission destinations or orbits considered except for some Earth orbits, this requires the lowest Delta-V. The Delta-V required is approximately 3.65 km/sec which is slightly more than required to reach Earth escape velocity. • Only a single propulsive phase from low Earth orbit is needed. • There is a launch opportunity any day. The waste package could be passive, requiring no active spacecraft systems . Disadvantages of Elliptical Solar Orbit • There is no assurance that the waste package will not re-encounter the Earth for periods measured in thousands of years. • There is an abort gap past Earth escape velocity which would preclude recovery. If the launch vehicle should fail after reaching Earth es- cape velocity, the waste package would be left in an unplanned solar orbit with subsequent Earth encounter possibilities. With the current state-of-the-art launch vehicle tech- nology, it would be impractical to re- cover the waste package from these orbi ts . There is no assurance that trajec- tories can be developed (and demon- strated analytically) which eliminate the possibility of reencounter with Earth for times on the order of a mil- lion years. Because of this uncer- tainty, Earth escape type of solar or- bits cannot be established as a proven, acceptable destination for time periods greater than thousands of years at this time. Circular Solar Orbits In order to provide a positive separation between the orbit of the 8.32 BNWL-1 900 waste package and the orbit of the Earth, the waste packages could be placed in circular solar orbits, ei- ther inside or outside Earth's orbit about the Sun. These circular orbits should be either inside 0.983 AU (as- tronomical unit, which is the mean distance between the Earth and the Sun, 149,000,000 kilometers) or out- side 1.071 AU (the perihelion [the point nearest the Sun] and aphelion [the point furthest from the Sun] distances of the Earth's elliptical orbit around the Sun) to ensure that the waste package does not collide with the Earth. There is an incen- tive, however, for going no further than necessary since the required Delta-V increases with increasing dis- tance from the Earth's orbit. For comparison purposes, a final orbit ra- dius of 0.90 AU , which is inside the Earth's orbit, is used in this study. Starting from Earth orbit, two pro- pulsive burns are required to reach the desired 0.90 AU circular solar or- bit. The first burn requires 3.26 km/sec Delta-V from the launch system to inject the payload to slightly past Earth escape energy. After escaping from the Earth, the waste package is in an elliptical solar transfer orbit having the de- sired perihelion but with an aphelion still at the Earth's orbit from the Sun. The second burn adds 0.81 km/sec (for a total incremental Delta-V of 4.07 km/sec) and circu- larizes the orbit. This burn is per- formed by another propulsion stage, upon reaching perihelion after approx- mately a six-month coast. Advantages of Circular Solar Orbits • The Delta-V required is low in com- parison to some of the other destina- tions. For the 0.90 AU circular solar orbit, a total Delta-V of 4.07 km/sec is required. • There is a launch opportunity any day. Disadvantages of Circular Solar Orbits • The problem of assuring the sta- bility of solar orbits for times on the order of a million years is unre- solved, generally for the same rea- sons given for Earth escape and high Earth orbits. Presumably, the final orbit could be placed sufficiently far from the Earth's orbit to pre- clude a subsequent collision with the Earth over the times required. • There is an abort gap past Earth escape velocity. • In addition to the launch system, another propulsion system along with guidance, control and communications is required to perform the second burn. It is impractical to accom- plish this burn with the launch sys- tem due to the long coast phase (about 6 months). This disadvantage could be diminished by performing the second burn for ci rculari zation with a simple spin stabilized, solid rocket motor. • If the ci rcul ari zation burn should fail, the waste package would be left in an elliptic solar orbit, intersect- ing the Earth's orbit near aphelion. For these cases there is a high proba- bility that the payload will even- tually re-encounter the Earth. This 8.33 BNWL-1 900 probability can be reduced by using departure trajectories similar to those suggested earlier for the Earth escape case. Integrity of the waste package is an important consideration for this mission because its possible disinte- gration over long periods of time can influence the choice of an interior or exterior orbit. If the waste pack- age should disintegrate, the Poynting- Robertson effect will tend to draw the smaller fragments into the Sun. If part of the waste package should vaporize, the solar wind could tend to move some of the material out from the Sun. If the integrity of the waste package cannot be assured over the long time period, these and other effects will have to be evaluated, not only in making the selection of orbit location, but also to establish the ultimate destinations of the waste mater i a 1 . If the integrity of the package and the stability of the circular solar orbits (near Earth) can be as- sured for for the time period of con- cern, circular solar orbits can be considered as a possible disposal destination. In addition, further study is required to evaluate the consequences of possible failure si tuations . Solar Orbit via Venus and Mars Another method for achieving solar orbits that do not cross the Earth's orbit is to swingby another planet, using the gravitational attraction of that planet to change the initial swingby trajectory. The resulting post-swingby trajectory does not cross the Earth's orbit; however, it will periodically cross the swingby planet's orbit. Both Mars and Venus swingbys can be achieved with Delta-Vs only slightly higher than Earth escape. The total Delta-V con- sists of two Delta-Vs. The first Delta-V performed by the Launch vehi- cle injects the payload onto a target planet swingby trajectory, the second Delta-V, performed by another propul- sion system after swinging by the target planet, places the waste pack- age in the desired solar orbit. The above maneuver is performed to pre- vent a subsequent encounter with the swingby planet. The total Delta-V for either a Venus or Mars swingby missions is approximately 4.11 km/sec Advantages of Solar Orbit via Venus or Mars • The total Delta-V required for either a Venus or Mars swingby mis- sion is relatively low in comparison to some of the other destinations. • With a properly oriented swingby the trajectory can be altered so that the post-swingby orbit will no longer cross the Earth's orbit. Disadvantages of Solar Orbit via Venus or Mars • For swingby missions the launch opportunity is limited. A launch opportunity to Venus occurs only once every 19 months and to Mars about once every 26 months. The duration or "width" of each of these launch opportunities can be about three to four months long without major in- creases in injection Delta-V (the wider the launch opportunity the higher the required injection Delta-V). 8.34 BNWL-1 900 • The waste package will require a midcourse trajectory correction system (with currently achievable in- jection accuracies) to insure achiev- ing a proper swing by position at the swi ngby planet. • An additional propulsion system is required to prevent a post encounter with the swingby planet. This pro- pulsion system and associated systems must perform reliably after a long coast phase (many months). • The problems of long-time sta- bility of the solar orbit and contain- ment system integrity are unresolved, although these problems would be less important than for the previously dis- cussed destinations which are closer to the Earth. • There is an abort gap past Earth escape velocity. Launch opportunities for either a Venus or Mars swingby appear to be quite limited. Such an operation would be expensive in terms of re- quired Shuttle fleet size, number of launch facilities and use of ground crews. (For example, the reusable Space Shuttle is expected to have a two-week turn-around-time between launches.) These swingby missions offer no outstanding advantages over the 0.90 AU solar orbit (which can be launched on any day) . Solar System Escape • Since both Earth orbit and solar orbit destinations have uncertainties regarding long-time orbit stability and containment system integrity, solar system escape and solar impact should also be considered as possible waste package destinations. Of the two, it takes less energy to escape the solar system, and this case will be discussed first. Direct Solar System Escape . This can be achieved with a single propul- sion burn from low Earth parking orbit with all the propulsion and guidance provided by the launch vehicle. Advantages of Direct Solar System Escape • The waste package is removed from the solar system. • The waste package can be passive and requires no additional propulsion or astrionics systems. • There is a launch opportunity any day. Disadvantages of Direct Solar System Escape • An 8.75 km/sec Delta-V is required This is high in comparison to the Delta-Vs required for high Earth orbits and solar orbits. • There is an abort gap past Earth escape velocity. There is a small variation in in- jection Delta-V depending on the launch day. The most efficient tra- jectories will be in or near the ecliptic plane of the earth's path and consequently will fly through the asteroid belt. There is no diffi- culty in targeting the trajectory to miss the outer planets. As a point of interest it takes approximately 20 years for the waste package to reach the mean orbital distance of Pluto (the planet in our solar system which is the farthest from the Sun) but it will take over a million years to reach the distances of the nearest stars. Solar system escape is the most attractive destination discussed thus far for assurance of long-term waste disposal. 8.35 BNWL-1900 Solar System Escape via Jupiter Swi ngby . Solar system escape can be achieved with a properly designed swi ngby of Jupiter using a single pro- pulsion phase from low Earth orbit. As a result of using a Jupiter swi ngby, the Delta-V required to achieve solar escape energy is some- what less than that required for a direct solar system escape mission. Advantages of Solar Escape via Jupiter Swingby • The waste package is removed from the solar system. Disadvantages of Solar Escape via Jupiter Swi ngby • The Delta-V required is approxi- mately 7.01 km/sec, which is still high in comparison to some of the other destinations. • The launch opportunity is limited, occurring only once every 13 months with perhaps a 60 to 90-day launch durati on or "wi dth . " • A midcourse trajectory correction capability is needed as was the case for the Venus and Mars swingbys. • There is an abort gap past Earth escape velocity. The Jupiter launch opportunity is sufficiently limited that many facili- ties and personnel would be required to support the anticipated number of launches required, and in general the Jupiter swingby could be more re- strictive than the Mars and Venus swingbys. It would be simpler to use a direct solar system escape, even though the Delta-V is some 1.74 km/sec higher than for the Jupiter swingby. • Solar Impact A solar impact can be achieved di- rectly or via Jupiter swingby. Again the purpose in using a Jupi- ter swingby is to reduce the Delta-V. Direct Solar Impact . This can be achieved with a single propulsion phase out of a low Earth orbit with all the propulsion and guidance pro- vided by the launch vehicle. Enough Delta-V must be provided by the launch vehicle to cancel the Earth's orbital speed about the Sun, so that the waste package "falls into" the Sun. For a direct impact a Delta-V of approximately 24.08 km/sec is re- quired. A grazinq impact into the edge of the Sun could reduce the Delta-V requirement to about 21.34 km/sec . Advantages of Direct Solar Impact • The waste package is destroyed. • The waste package can be passive. • There is a launch opportunity any day. Disadvantages of Direct Solar Impact • The Delta-V required is extremely high. • There is an abort gap past Earth escape velocity. For this mission the Delta-Vs re- quired are far beyond the capability of current or currently planned launch systems and therefore are con- sidered impractical until more ad- vanced systems are developed. Solar Impact via Jupiter Swingby . As was the case for the solar system escape via Jupiter swingby, a solar impact can be achieved with properly designed swingby Jupiter using a sin- gle propulsion phase from a low Earth parking orbit. By using a Jupiter swingby to achieve a solar impact, the Delta-V required is appreciably 8.36 BNWL-1900 less than that required for a direct solar impact. For this mission the Delta-V required is above 7.62 km/sec . Advantages of Solar Impact via Jupiter Swing by • The waste package is permanently removed from earth and is destroyed . Disadvantages of Solar Impact via Jupiter Swingby • The Delta-V is still high in com- parison to some of the other destinations. • The launch opportunity is limited, occurring only once every 13 months with perhaps a 30 to 60-day launch duration or "width." • A midcourse trajectory correction capability is needed which in- creases mission complexity. • There is an abort gap past Earth escape vel oc i ty . The Delta-V required for this case is about 1.13 km/sec less than that required for a direct solar system es- cape mission. Many facilities and personnel would be needed to support the required number of launches dur- ing the limited launch opportunity. • Other Destinations Many other space destinations in addition to those discussed have been suggested. Examples include de- positing the waste packages on the Moon, on planets, in orbits of other planets, on asteroids, at Lagrangian equilibrium points and so forth. These destinations were not consid- ered in the detailed analysis al- though in some cases they could war- rant further investigation. The general arguments used by NASA against these destinations include (1) the regions are unexplored and/or are of scientific interest, (2) some of the regions could be of future value from an alternative applica- tions standpoint, (3) launch opportu- nities are limited, (4) deep space propulsion is required, and (5) in many cases the retro Delta-Vs are high for soft landings. Costs and energy requirements for placement on the moon or other extra- terrestrial bodies are expected to be between those for high earth and so- lar orbits and direct solar escape. It would be technically possible to use such repositories with state-of- the-art vehicles. • Comparison of Destinations To summarize the destinations dis- cussed, Table 8.10. lists the typical Delta-V requirements for the various missions and their principal advanta- ges and disadvantages. The Delta-Vs shown are representative for each des- tination, although there will be some variation depending on the particular launch opportunity and details of the mission profile. The Delta-V for high Earth orbits is an upper value for orbits between synchronous and lu- nar orbit altitudes. The Earth es- cape (solar orbit) Delta-V includes some provision for additional Delta-V in an effort to minimize the probabil- ity of a subsequent Earth re-encoun- ter as was discussed earlier. The Delta-Vs for the other solar orbits include the Delta-Vs required by the waste package after departing from Earth. Passive waste package im- plies that it will require no spec- ial space propulsion, midcourse or associated astrionics systems. The abort possibility past Earth escape 8.37 BNWL-1 900 TABLE 8.10 . Summary of Potential Space Destinations Destination Delta-V, km/ sec 4.11 Advantages High-Earth Orbit Low Delta-V Launch any day Passive waste package Can be retrieved Solar Orbits Via: Single burn beyond Earth escape 3.65 Circular Solar Orbit 4.11 Venus or Mars Swingby 4.11 Solar System Escape: Direct 8.75 Via Jupiter Swingby 7.01 Solar Impact: Direct 24.08 Via Jupiter Swingby 7.62 Low Delta-V Launch any day Passive waste package Low Delta-V Launch any day Low Delta-V Launch any day Passive waste package Removed from solar system Removed from solar system Package destroyed Launch any day Passive waste package Package destroyed Disadvantages Long-term container integrity required. Orbit lifetime not proven. Long-term container integrity required. Earth re-encounter possible (may not be able to prove otherwise). Abort gap past Earth escape velocity. Long-term container integrity required. Orbit stability not proven. Requires space propulsion system. Abort gap past Earth escape velocity. Long-term container integrity required. Limited launch opportunity (3 to 4 months every 19 to 24 months). Requires midcourse systems. Need space propulsion or have possibility of unplanned encounter. High Delta-V Abort gap past Earth escape velocity. High-Delta-V, Limited launch opportunity (2 to 3 months every 13 months). Requires midcourse systems. Abort gap past Earth escape velocity. Extremely high Delta-V. Abort gap past Earth escape velocity. Not possible with present vehicles. High Delta-V. Limited launch opportunity (1 to 2 months every 13 months). Requires midcourse guidance systems. Abort gap past Earth escape velocity. Note: Delta-V is the incremental velocity required to leave a low-earth orbit. An abort gap is a short time period wherein a controlled abort of the mission cannot be accomplished if the flight is off-course. 8.38 BNWL-1900 velocity is a disadvantage asso- ciated with all destinations beyond the Earth. The conclusions reached thus far indicate the primary candidate mis- sion destinations are direct solar system escape or possibly circular solar orbits for long-term disposal, or high Earth orbits for short term disposal, as shown previously in Fig- ure 8.5. The capability of possible launch systems for each candidate mis- sion destination is discussed in the next section. 8.2.5.2 Potential Space Transpor - tation Vehicle Perfor - mance and Cost • Basic Considerations Only the larger current and planned launch vehicles are consid- ered in this study. The vehicles con- sidered are shown in Figure 8.11. The Titan IIIE/Centaur is the expend- able booster that will launch the 1975 Viking mission to Mars. The Sat- urn V is the three-stage expendable Apollo booster. Its two-stage ver- sion was used to launch Sky lab. The Space Shuttle is primarily reusable and is expected to be operational in the 1980's. It is planned as a re- placement for virtually all the na- tion's space boosters in operation today. One of the most important factors in assessing the feasibility of space disposal is cost. The costs pre- sented in this section are those for the launch vehicles and their opera- tions which were used for assessment of technical feasibility. Conversion of these costs to those used for total waste management operating and capital costs is presented in section 8.5. These data can be used for com- parative purposes for preliminary de- termination of the best launch vehi- cles and the most promising mission destinations. • Expendable Launch Vehicle Perfor - mance and Cost Performance and cost data for the Titan IIIE/Centaur and the Saturn V (and the Space Shuttle) are listed in Table 8.11 for the high earth orbits and solar orbits. Performance data are based on a launch due East from the Eastern Test Range (ETR) into a 1 85- k i 1 ometer parking orbit. The upper stage of the launch vehicle pro- vides the Delta-V needed to acceler- ate the payload to higher energies from the parking orbit. The direct solar impact mission (24.08 kilome- ters per second) is not shown because it is well beyond the capability of current launch vehicles. The costs of the expendable launch vehicles are highly use-rate depen- dent. The Titan IIIE/Centaur cost is about $27 million at a production rate of four per year. At the higher launch rates expected for space dis- posal of radioactive waste, the cost would be expected to be considerably lower. For this study, it is assumed that the cost of the Titan IIIE/Cen- taur at high launch rates, can be re- duced about 30 percent and its cost is taken at $19 million as shown in Table 8.11. Similarly, the costs of the Saturn V and Saturn V/Centaur are taken at $150 and $155 million, re- spectively. Note that the costs used in this section include only the 8.39 BNWL-1 900 100 r- 80 60 o 40 20 TITAN 1 1 IE/CENTAUR SATURN V SPACE SHUTTLE FIGURE 8.11 . Possible Launch Vehicles for Space Disposal of Radioactive Waste costs of the launch vehicles and their operations. They do not in- clude operational costs associated with handling the nuclear waste con- tainer at the launch site or the inte- gration of the waste package with launch vehicle. These latter costs are included with those for total waste management in section 8.5. • Space Shuttle/Third Stage Perfor - mance and Cost The Space Shuttle by itself can de- liver payloads only to low Earth orbit. Missions beyond low Earth 8.40 BNWL-1900 TABLE 8.11 . Space Launch Vehicle Performance and Cost for High Earth Orbits and Solar Orbits Payload, kg r Launch Cost Launch Vehicle 10° Dollars Dollars/kg Expendable Vehicles: Titan III E/Centaur Saturn V Saturn V/Centaur 3,860 32,660 35,290 19 150 155 4,920 4,590 4,390 Space Shuttle: Reusable Tug Current Size 4,170 12.25 2,940 Reusable Tug Optimum Size 4,670 12.25 2,620 Centaur Current Size 6,490 16 2,460 Centaur Optimum Size 8,480 16.3 1,920 Note: a. Delta-V = 4.11 km/sec. b. For direct solar escape (Delta-V = 8.75), the payload for the Titan III E/Centaur and Saturn V = and for the Saturn V/Centaur is 7480 kilograms. orbit must be accomplished by having the Space Shuttle carry both a pro- pulsion stage and the mission payload to Earth orbit in its cargo bay. The propulsion stage is generally re- ferred to as a Space Shuttle third stage. After the third stage and pay- load are deployed in Earth orbit from the Shuttle Orbiter, the third stage will inject the payload to its desti- nation. Existing expendable upper stages are currently being evaluated for early use as Space Shuttle third stages. These staaes would be ex- pended on each flight. However, it is planned to eventually develop a new reusable Space Tug explicitly for use as a Space Shuttle third stage and having the capability of being re- covered and reused. The Space Shut- tle would launch the Tug and payload into low Earth orbit. After the Tug and payload are deployed from the or- biting Shuttle, the Tug will inject the payload to its mission destina- tion. Followinq the injection burn, the payload is separated from the Tug and the Tug does a series of burns to return to the waiting Shuttle Orbiter for recovery and reuse. 8.41 BNWL-1900 Several Space Shuttle/Third Stage options were considered in this study. These include: (1) One of the reus- able Space Tug concepts under study by NASA. It is designed to have the capability of performing a round-trip mission to geostationary (synchronous) orbit with a 1360-kilogram payload. It is a hydrogen-oxygen fueled stage with an engine specific impulse of 470 seconds and has a propellant ca- pacity of approximately 24 ,040 kilo- grams. (2) A similar reusable Tug but optimally sized for the waste dis- posal mission. (3) The existing ex- pendable Centaur stage. It also uses hydrogen-oxygen propellants and has an engine specific impulse of 444 sec- onds and a propellant capacity of about 13,610 kilograms. (4) A simi- lar expendable Centaur stage but re- sized for the waste disposal. With these various Space Shuttle/ third stage options useful payloads are achievable to only high Earth orbit and solar orbit destinations. Their performance, which is based on a mission Delta-V of 4.11 kilometers per second, is shown in Table 8.11. The performance data are based on a Space Shuttle delivery capability of 29,484 kilograms into a due East 185- kilometer orbit, which is a Shuttle specification. Note that the opti- mally sized third stages have higher payl oads . For the high Earth orbits and solar orbits, the reusable Tug, at its current size, can deliver a pay- load of 4,170 ki 1 ograms, whereas the optimally sized Tug (about 20,870 kilograms propellent) can deliver a payload of 4,670 kilograms. The current-size Tug performance is lower than the resized Tug because its pro- pellant capacity is too large for the higher payload weight of waste dis- posal missions, and the Tug must be off-loaded. In the case of the Cen- taur, the propellant capacity is too small to utilize the full orbital ca- pability of the Space Shuttle. The performance of the Centaur stage can be improved if its propellant capac- ity is increased. For the high Earth orbits and solar orbits, the current Centaur stage can deliver 6,490 kilo- grams. An optimally sized Centaur (about 17,240 kilograms propellant ca- pacity) can deliver a payload of 8,480 kilograms. It should be recognized that the higher payload capability shown for the Centaur stage is a consequence of its being expended rather than recov- ered. For the reusable Tug, a por- tion of its propellant is required to return to the Shuttle Orbiter waiting in low Earth orbit. For the expend- able Centaur stage, all the propel- lant is used to achieve the desired mission Delta-V, and its payload is accordingly higher. If the Tug were expended, its performance would be comparable to that for the optimally sized Centaur stage. The cost per Space Shuttle flight is currently estimated at approxi- mately $10.5 million. In addition, the cost per reusable Tug flight is assumed to be $1.75 million, which in- cludes operations, refurbishment and amortization of a unit production cost of $20 million/Tug. Totalinq 8.42 BNWL-1 900 the two, the cost per flight of a Space Shuttle/reusable Tug is about $12.25 million. The cost of the ex- pendable Centaur stage at the high launch rates required for waste dis- posal would be about $5.5 million. In total , the cost of a Space Shuttle/expendable Centaur launch is about $16 million. • Launch Vehicle Performance/Cost Com pari son Except for the Saturn V/Centaur, the launch vehicles considered thus far can only deliver useful payloads to high Earth orbit or solar orbit destinations. In order to provide an overall vehicle comparison for these destinations, the payload, cost per flight and cost per kilogram of pay- load delivered to a Delta-V of 4.11 km/sec were summarized in Table 8.11. These data should be used only for making preliminary com- parisons since other factors will have to be considered in making a vehicle selection. For example, there are limits on the desired waste package size. Also, the nuclear waste is only a small fraction of the total waste package weight, and this fraction will vary with waste package size. These and other factors will influence the choice of a launch vehi- cle for a particular destination. Nonetheless, Table 8.11 shows that the Space Shuttle vehicles are more cost effective than the current ex- pendable launch vehicles. The cost per pound of total payload delivered for the Space Shuttle vehicles is on the order of one half of that when using expendable launch vehicles. For the shuttle launched missions, it appears worthwhile to resize the upper stages for the waste disposal mission. The improved performance and cost effectiveness should readily justify the nonrecurring costs asso- ciated with resizing the stages. For the high Earth orbits or solar orbits the cost per pound of payload de- livered for the resized Centaur stage is about 25% lower than for the re- sized reusable Tug. This indicates that an expendable Shuttle stage would be more cost effective than a reusable stage. This conclusion is sensitive to the required mission Delta-V. If the required mission Delta-V were below about 3.4 km/sec, a reusable Shuttle third stage (Tug) could be more cost effective than an expendable stage (Centaur). In addi- tion to performance and cost, safety considerations and specific mission details can influence the final choice of a Shuttle third stage. From this reasoning, both reusable and expendable Shuttle third stages were considered for further eval uations . • Multiple Space Tug Configuration Performance and Cost The only launch vehicle consid- ered thus far that has a useful pay- load capability for the direct solar escape mission is the Saturn V/Cen- taur. As shown in Table 8.11, it can deliver a payload of about 7,480 kilo- grams to this destination. At a launch cost of $155 million, this re- sults in a specific cost of 20,720 dollars per kilogram, or roughly an order of magnitude higher than for the Shuttle launched cases to high Earth or solar orbits. One possibility for providing a more cost effective solar escape capability is 8.43 BNWL-1 900 to use several Shuttle/Tug launches to assemble a larger vehicle in Earth orbit. This same approach could also be used to provide higher payloads for the Earth orbit and solar orbit destinations. The procedure would be to use sev- eral Shuttle launches to place sev- eral Tugs in low Earth orbit along with the payload. The Tugs, which have the inherent capability of being able to rendezvous and dock with each other, would be assembled in orbit to form a tandem vehicle. In performing the mission, the Tug stages will burn sequentially, and each stage, if it is to be recovered, will return to its waiting Shuttle Orb iter. In this preliminary evaluation of a tandem vehicle, only the fixed size Tug concept was investigated. It is assumed to be available in both reus- able and expendable configurations. The Tug and Shuttle performance param- eters and costs are the same as dis- cussed earlier. The one exception is the cost of an expended Tug. The ex- pected unit cost of the reusable Tug is on the order of $20 million. If the waste disposal mission required expending a Tug, the cost of the ex- pendable Tug could be considerably lower. The production rate for an expendable Tug would be much higher than for a reusable Tug since each disposal mission would require a new Tug. The high use rate would prob- ably justify development of an ex- pendable Tug incorporating only the features necessary for accomplishment of the waste disposal mission. As an alternative approach, a modified ver- sion of the existing Centaur stage could be used as an expendable Tug. An accurate cost for the expendable Tug cannot be established at this time, but for the purposes of this study it is taken as $6.0 million per flight. Several tandem Tug configurations can accomplish the direct solar escape mission, two of which are con- sidered here for illustrative pur- poses. The first tandem configura- tion considered consists of two stages, a reusable Tug plus expend- able Tug, and requires two Shuttle launches. The first Shuttle launch carries a full Tug to orbit and the second carries an off-loaded expend- able Tug plus payload. The second tandem configuration considered con- sists of three stages, two reusable Tugs plus an expendable Tug, and re- quires three Shuttle launches. The first two Shuttle launches carry full reusable Tugs to orbit and the third carries an off-loaded expendable Tug plus payload. In both configurations the recoverable Tugs are the lower stages (burned first) since this is an optimum arrangement. It is as- sumed that the recoverable Tugs are brought back to Earth with the Shut- tle Orbiters used to initially launch the Tugs. That is, no additional Shuttle cost is charged for returning a Tug. The ideal payload performance capa- bility to solar escape for the two- and three-stage tandem configurations is 3,900 and 6,080 kilograms, respec- tively. However, gravity losses will significantly reduce the actual per- formance of these multi-Tug configu- rations. The gravity losses have been determined for these configura- tions assuming the Tug has a thrust 8.44 BNWL-1900 level of 88,940 Newtons. The actual payload capability of the two-stage and three-stage configurations for direct solar escape is 2,270 and 3,040 kilograms, respectively. A higher Tug thrust level could be used to reduce the gravity losses, but it is not expected that the new Tug engine will have a thrust level higher than 88,940 Newtons. Another approach to reducing the gravity losses is. to use a technique referred to as perigee propulsion. This is operationally more complicated and necessitates carrying the waste pack- age once around the Earth in an ellip- tical orbit between Tug burns. How- ever, using perigee propulsion increases the payload capability of the two-and three-stage configura- tions for direct solar escape to 3,270 and 4,400 kilograms, respectively. An overall comparison of launch vehicles performance and cost for the direct solar escape mission is shown in Table 8.12. The expendable Saturn V/Centaur provides the highest payload weight, but at a cost of about $20,700 per kilogram. The mul- tiple Shuttle/Tug configurations using perigee propulsion achieve lower payloads but at a cost of about $9, 000 per ki 1 ogram . • Space Transportation System Recommended The currently planned Space Shut- tle is more cost effective than cur- rent expendable launch vehicles by about a factor of two. The Space Shuttle will require a third stage to perform the disposal missions. De- pending on the particular mission, this could be either a reusable stage, such as the Space Tug, or an expend- able stage such as a Centaur. In either case, the third stage should be resized for the selected disposal mission. In fact, the launch rates required for waste disposal are ex- pected to be sufficiently high that it would probably be worthwhile to develop a version of the entire launch vehicle dedicated to providing maximum performance, higher' relia- bility for the disposal mission, and 1 owes t cos t . In this study, only current or planned space transportation systems were considered. It should be recog- nized, however, that the need for waste disposal is expected to extend far into the future and new space technology and systems development can be expected. These new develop- ments should be reviewed periodically for their application to space disposal . 8.2.6 Energy Bal ance The energy required to place waste material in a space disposal trajec- tory is compared on a fundamental basis here, with the beneficial elec- trical energy obtained from the pro- duction of the waste in the nuclear reactor . The incremental velocity required to escape the solar system from a 100-nautical-mile earth parking orbit is 8.8 kilometers (28,700 ft) per f 1 2 ) second. ' The velocity required for the earth parking orbit is 7.8 kilometers (25,600 ft) per second. The total velocity required for solar system escape is thus 16.6 kilometers (54,300 ft) per second. With 30 kilo grams of transuranium waste resulting 8.45 BNWL-1 900 TABLE 8.12 . Space Launch Vehicle Performance/Cost for the Direct Solar Escape Mission Launch^ Payload, , Cost, Cost Launch Vehicle kg 10 Dollars Dollars/kg Saturn V/Centaur 7,480 155 20,720 (2,1,1)^ Shuttle/Tug/Tug Configuration Without Perigee 2,270 28.75 12,660 Propul sion With Perigee 3,270 28.75 8,790 Propulsion (3,l,2) (b ) Shuttle/Tug/Tug Configuration Ideal Delta-V 6,080 41.0 6,740 Without Perigee 3,040 41.0 13,490 Propul sion With Perigee 4,400 41.0 9,320 Propulsion Notes: a. Delta-V = 8.75 km/sec. b. (— > --. --) represent the number of Shuttle launches, the number of expendable Tugs, and the number of reusable Tugs, respectively. from the operation of a 1000 MWe reac- solar escape trajectory is approxi- tor for 1 yr and a weight factor of mately 3 x 10 times the nuclear 20 required for shielding, the energy electrical energy obtained in gener- required to place the waste in a ating the waste. a ' The value was obtained as follows: 1 2 1 Energy = j MV = j • 67 lb (30 kg) of waste x 20 shielding factor # , 54 o 00 \2 = 6.2 x 10 10 ft-lb = 8.5 x 10 9 kg-m The electrical energy obtained from operation of the reactor is: 1,000 MWe x 365 days/yr x 24 hr/day = 1,000,000 kW x 8,760 hr g = 8.76 x 10 kW-hr (Assumed 100% operating efficiency) 8.76 x 10 9 kW-hr x 2,655 x 1 6 ft-lb/kW-hr = 2.32 x 10 16 ft-lb = 3.2 x 1 1 5 kg-m The energy ratio is thus Energy required for solar escape of Transuranics/Energy Produced . 8-5 x 10 9 kq-m = ZJ x 1Q -6 3.2 x 10 15 kg-m 8.46 BNWL-1 900 8.2.7 Safety The following analysis (Sections 8.2.7.1 through 8.2.7.5) of the safety considerations of nuclear waste disposal in space was developed in Reference (12). These sections are included as authored, with minor changes to fit the format of this report . The fundamental philosophy of nu- clear safety requirements for radio- active waste disposal missions in space can be stated as follows: Potential radiation exposure and harm- ful contamination of individuals, the population at large, and the ecology shall be negl ig i bl e . The statement in the previous para- graph includes also celestial bodies as required by the treaty to promote peaceful exploration and use of outer space which 60 nations, including the United States, have signed. For operations during all phases of a nuclear waste disposal mission, the exposure and contamination values should be negligible for the popula- tion at large and within' the permis- sible standards for the personnel in- volved in the mission. Criteria governing radiation exposure to hu- mans during ground handling and trans- portation which may be applicable can be found either in 10 CFR, Parts 20 and y\ K ' or in Reference (11). The nuclear waste disposal mis- sions discussed here are similar to other space shuttle/space trip mis- sions carrying radioactive materials to synchronous earth orbit or to plan- etary orbits. The primary difference lies in the comparatively large amount of radioactive material con- taining actinides being transported during the waste disposal missions. The purpose of this analysis is to evaluate the response of the nuclear waste package to the potential acci- dents that could occur during the various phases of waste disposal mis- sions. It was possible only to per- form a qualitative evaluation because the mission hardware and systems and the mission proper are only in a pre- liminary state of definition. And furthermore, definitive design cri- teria do not exist for this particu- lar type of nuclear waste disposal method . 8.2.7.1 Nuclear Safety Requ i remen ts The nuclear waste contained within the waste package imposes certain nu- clear safety requirements on the pack- age design, its supporting equipment, and mission operations. Some of these waste package requirements are listed below: • Sub-critical design under all con- ditions Encapsulation of waste material Shielding for external radiation Re-entry protection Impact protection Fire and explosion protection System to transmit location of capsul e Shuttle orbiter supporting equipment needs are • Temperature Controls • Monitoring Equipment Operation requirement needs are • Parking orbit altitude control with long decay time 8.47 BNWL-1900 • Recovery preparedness • Future encounter avoidance control • Orbital retrieval means. 8.2.7.2 Accident Model Consid- erations In this section general potential accident cases are discussed qualita- tively. Each of the cases includes a series of environments which could be present at various times during the accident. Development of probability numbers is beyond the scope of this study. • Ground Handling Provided proper procedures are es- tablished, the probability of an acci- dent while handling the nuclear waste package, including installation within the Shuttle cargo bay,will be extremely sma 1 1 . • Launch Pad Abort The assumed potential accident would occur at or immediately after launcKj causing the Space Shuttle to explode and burn, thus exposing the nuclear waste package to an adverse environment. The environment cre- ated by such an accident would be: blast overpressures, residual liquid and solid propellant fires. The residual fire of the solids could last about five minutes at a tempera- ture of approximately 2600°K. • High Velocity Impact If a malfunction caused the shut- tle to make a 180° change in direc- tion soon after launch, a powered impact could occur. The resulting environment would be similar to those encountered during launch pad abort except the impact would be at high velocity. Furthermore, a portion of the propellants would have been used, the residual fires would have shorter duration, and the fireball would be smaller. The presence of a crew on board would reduce the probability of occurrence to a very small number. The actual impact velocity for the nuclear waste package itself would be lower than that of the Shuttle be- cause of the cushioning effects of the shuttle structure. • Failure During Ascent Many ascent failures and malfunc- tions which would lead to catastro- phic failure in an unmanned space vehicle would require mission changes or aborts, but not an accident. How- ever, there remains a possibility, although small, that a catastrophic accident would occur on board the Shuttle. If the Shuttle remains in- tact, the result of the failure would likely terminate with relatively low impact velocity. If a major explo- sion occurs on board, the nuclear waste package would impact the ground at a high velocity. If the package impacts while still attached to the Tug, there may be an explosion with blast overpressure, fragments, fire- ball, and liquid propellant fire but at a considerably lower magnitude than during a launch pad abort be- cause of the relatively small quan- tity of Tug propellants. • Crash Landing There is a possibility that the Shuttle Orbiter will make a crash landing. If there was insufficient time to dump the Tug propellants prior to the landing, overpressures, impacts of fragments, fireball, and propellant fires could occur. 8.48 BNWL-1 900 • Uncontrolled Re-entry and Impact The potential accident modes during re-entry considered here are those which might occur after the nu- clear waste package has been deployed from the Shuttle. Basically there are then two possibilities: The pack- age re-enters by itself or it is still attached to the Tug. The envi- ronment encountered by the package or the Tug/package (or Tug/Tug/package ) depends on. the entry velocity and entry angle into the atmosphere. (In this exploratory study only the ver- tical entry was analyzed). The nu- clear waste package would be exposed to re-entry heating and thermal stress. If the heat shield of the package were to fail, the waste mate- rial encapsulation could fail also and expose the radioactive waste to the re-entry and impact conditions. (Note that the purpose of the heat shield is to preclude this latter f ai 1 ure mode. ) • Post Impact After an intact impact of a nu- clear waste package, it is exposed to environments which could cause over- heating and melting, oxidation, and corrosion, which may eventually cause some release of the radioactive mate- rial and/or an increase in the exter- nal radiation dose. However, for any post-impact condition the external radiation dose represents a hazard. The dose for an intact package was assumed to be 1 rem/hr at 1 meter from the surface of the package. For any degree of breaching of the pack- age the external radiation is in- creased, whether or not radioactive material is released. 8. 2.7.3 Analytical Results of Accident M otte 1 Only a qualitative evaluation was made of the possible release of radio- active materials and the potential hazards resulting from external radia- tion. A quantitative evaluation and the determination of probabilities can only be made after a more de- tailed hardware, system and mission definition has been made. Some of the analytical methods have been confirmed with experiments, such as large sphere impact tests of more than 300 m/s velocity and fragmen- tation tests. * Nuclear Waste Package Response The basic nuclear waste package de- sign used in the analyses was the sin- gle package design for disposal of transuranium waste containing 1% fis- sion products and shown in Figure 8.3, Where possible, the analytical methods checked against experiments that were closely related to the acci- dent conditions. • Overpressure An accident at the launch pad which results in an explosion of the main liquid propellants could pro- duce a blast overpressure of approxi- mately 150 atmospheres (assumes a mixing mode yielding 20% TNT equivalent.) In the analysis it was assumed the re-entry shell is stripped away, leaving the spherical stainless steel impact vessel to take the full overpressure. This assumption repre- sents a conservative approach. The spherical shape is an ideal shape to withstand the pressure. The yield stress of the stainless steel is 8.49 BNWL-1900 2400 atmospheres (35,000 lb/in. 2 ). An empty vessel with a radius of 0.68 meters and a 2 . 54-centimeters- thick wall can withstand an external pressure of 175 atmospheres without yielding. Since this is greater than the overpressure, the nuclear waste package will not be breached by the blast overpressure. However, because of the differences in compres- sive strengths of materials, the ex- ternal pressure may cause the outer impact shell to shift relative to the internal waste, thus deforming the impact shell and reducing the shielding thickness. • Fragments During accidents of the explosion type, fragments of varying sizes and of various materials (predominately aluminum) could impact the nuclear waste package at varying impact ve- locities. An analysis was made as- suming aluminum fragments (sharp and blunt) impacting at 1520 m/sec (5000 ft/sec). The results indicated the containment vessel would not be penetrated. In addition to the analytical study an experimental test was set up with aluminum pellets fired at a stainless steel sphere with a wall thickness of 1.58 centimeters. • Firebal 1 No calculations were performed for this condition. Comparison with other capsules involved in fireball tests indicated that because of the short duration (seconds) and the large mass of the nuclear waste pack- age, no serious damage should result to the nuclear waste package due to the fireball. • Residual Propellant Fires Of the two types of fires, the solid propellant produces the higher temperature, 2300°C (4200°F). To evaluate the response of the nuclear waste package to the. solid propel- lant, a heat transfer model was es- tablished which consisted of 72 nodes for the various layers of mate- rial. It was assumed that the re-en- try shell had been stripped away by the overpress ure . Initially the surface heat flux is high since the temperature of the impact shell is low. As the surface temperature rises, the radiation ef- fect from the propellant fire dimin- ishes and the heat flux drops. The temperature of the impact shell ap- proaches the melting point in about five minutes. Because of the high temperature it is possible to breach the outer impact shell if the solid fire can last for five minutes adjacent to the package. This assumes the solid fuel burns from one side and that the full thickness of the solid fuel is in effect. Although it is possi- ble to cause melting of the outer shell, other layers will still pre- vent release of the radioactive nu- clear waste. However, the shield- ing (lithium hydride) could be lost and the external dose rate would be increased. This situation would re- quire a shielded or remotely oper- ated retrieval vehicle. • Atmospheric Re-entry In the event of an aborted mis- sion, the nuclear waste package could return to earth in an uncon- trolled manner (i.e., not on board 8.50 BNWL-1 900 the shuttle). For this type of acci- dent there are many combinations of velocity and angle of re-entry. The case which was analyzed and which re- sulted in establishing the required thickness for the re-entry shell was the vertical re-entry angle at 11 km/sec (36,000 ft/sec). This type of re-entry exposes the re-en- try shell surface to a peak heat flux of 300 kW/cm 2 . Although the high heat flux lasts a short time it was felt that this represented a critical condition. For the above condition the re-en- try shell has sufficient thickness to prevent melting through to the im- pact shell. The calculations indi- cate that the impact shell tempera- ture does not increase. Because of the extremely high heat flux it will be difficult to perform a test of this kind to con- firm the calculations. In addition, other re-entry conditions may impose other more severe conditions on the package (i.e., skip re-entry) al- though the probability for this type of re-entry will be yery low. • Impact Following an uncontrolled abort (i.e., the nuclear package is not brought back with controlled Shuttle) the package will impact the Earth. Based on the design of the re-entry shell, the impact velocity should be 300 m/sec or less. This 1 velocity would be exceeded only if there was an abort with a tug attached and thrusting in. Based on the experiments and analyses it appears that in most im- pacts on Earth the package will be buried in the ground with relatively little damage to the outer shell. However, if the package lands on a surface such as a solid, noncrush- able one that does not absorb any of the energy, the nuclear waste pack- age will probably be breached. This condition may or may not result in release of radioactive waste, since it is a low percentage of the matrix and is encapsulated with tungsten pro- tection layers. It can be assumed that for the harder surfaces the nu- clear waste package is deformed and possibly breached, and in either case the shielding will be reduced, pro- ducing the possibility of higher than designed for external dose rates. • Post Impact After an impact the nuclear waste package will either be intact or breached and either buried beneath the surface, partially buried, or on top of the ground. A series of cal- culations was performed on various degrees of burial. The results indi- cated that in the no-burial or par- tial-burial cases the vessel would not rupture within 23 days (approach- ing equilibrium condition). For deep burial, all cases except those con- taining waste that produced less than 2 kilowatts of thermal power, re- sulted in rupture of the impact ves- sel. This rupture was due to the in- creased pressure from helium released in the decay process and from disso- ciating lithium hydride. The rupture would in all probability be a minor crack to relieve the pressure. No experimental data were avail- able to determine whether or not the surfaces would melt. The calcula- tions did not account for any mate- rial changes in the soil (conductiv- ity was varied with temperature). 8.51 BNWL-1900 If the impact vessel remains in- tact there will be no oxidation and no corrosion for extended periods of time. If the outer vessel is breach- ed, there would be some loss in the effectiveness of the shielding. The waste material already in the oxide form should not react with the sur- rounding environment. The radiation level in the immediate area would in- crease, due primarily to the loss of shielding. Because of this, it would be desirable to locate and retrieve the nuclear waste package as soon as possible. 8.2.7.4 Recovery of Nuclear Waste Package It is desirable to locate an abort- ed waste package as soon as possible. For those aborts near the launch pad and prior to reaching a parking orbit the vehicle will be tracked and its location predetermined. For these cases recovery procedures can be worked out satisfactorily to recover the package without undue hazards to people in the vicinity. For those aborts which can occur approaching or past Earth escape Delta-V's, in which the nuclear waste package could re- encounter the Earth at some future date, knowledge of impact position may not be immediately known. It would be helpful if some kind of de- vice for transmitting a signal could be incorporated in the package to help in locating the package prior to impact. missions, as described in this report appear feasible. The waste package design concept with the various pro- tective shells provides a means for preventing release of radioactive waste constituents in most hypotheti- cal accidents. Additional testing would be required to confirm the concept . The safety study on accident mod- els and package responses point out certain key issues to maintain over- all nuclear safety for space disposal of nuclear waste packages. These key issues are : 1. The waste package should be de- signed to maintain integrity without releasing significant amounts of its radioactive contents throughout all potential hazardous events. 2 . The external radiation of the package, whether intact or damaged, has to be held to a minimum so that recovery can be accomplished without undue exposure to the population. (This might require development of location and retrieval means for Earth- impacted waste packages.) 3. Potential accident conditions that could lead to uncontrolled re-en- try of the waste package have to be minimized. This would be accom- plished by careful selection of tra- jectories, by use of highly reliable vehicles and by developing space re- trieval means (vehicle) for retrieval of these packages from accidental or- bits beyond Earth escaoe. 8.2.7.5 Summary of Safety Analyses by NASA Within the framework of this explo- ratory study, nuclear waste disposal 8.2.7.6 Space Contamination One of the concerns expressed about disposal in space is the possi- ble contamination of space. However, 8.52 BNWL-1 900 the dispersion factors achieved in event of radionuclide release in space disposal are enormous. As an indication of the magnitude of the dilution factor achieved, as- sume that waste material is dropped into the sun and subsequently forced back into the solar system. Assume that it is evenly distributed throuqh- out the volume of a sphere with a ra- dius equal to the earth's distance from the sun (93,000,000 miles or 150,000,000 kilometers). The volume of this sphere is 1.4 x 10 40 cm 3 .' a ' By the year 2000, the total accu- mulation of transuranics and fission products (obtained from Volume 1 of this report series) is: Transuranics 1,770 M Ci Fission Products 143,000 M Ci If the total waste accumulated to the year 2000 were uniformly dis- 40 3 persed in this sphere (1.4 x 10 cm volume), the concentration would be 1 .06 x 10" 23 yCi/cc. ^) The allowable concentrations in air of several elements, from 10 CFR 20 are ,-14 Pu-238 7 x 10 yc/ml Cs-137 2 x 10" 9 yc/ml 1-129 2 x 10" 11 yc/ml Assuming an allowable concentra- tion comparable to that of Pu-238, the actual versus allowable concentra^ tion would be 1 .06 x 10 -23 ,-14 = 1.5 x 10 -10 7 x 10 For comparison purposes, assume that the following quantities of stronti um-90 , cesium-137 and pluto- nium-239 which have been put into the earth's atmosphere by weapons test- • (42) ,. , .... ., ing are dispersed within one mile of the earth's surface (actually much of this is on the surface or in the oceans now ) : Stronti um-90 Cesium-1 37 PI u toni um-239 Total 20,000,000 Ci 34,000,000 Ci (°) 400,000 Ci 54,400,000 Ci The volume within 1.6 kilometer (one mile) of the earth's surface (both above and below) is about 5.3 x 10 23 cc ,d) Volume is calculated as follows 4 „3 4 ,„„ „„„ „„„J |ttR" = yrr(93,000,000) 3.37 m24 ., 3 x 1 miles There are 161,000 cm in a mile or 4.17 thus : p/i -5 3.37 x 10^ mi^ x 4.17 x Concentration is calculated as follows 1n 15 3, .3 10 cm /mi 10 15 cm 3 /mi 3 = 1 .4 x 10 40 cm 3 The vol ume i s 1 ,770 x 10 6 + 143,000 x 10 6 Ci 1 44 , 770 x 10 6 Ci x 1 6 yCi/Ci 1 .4 x 10 40 cc 1 .4 10 40 cc 1 .06 x 10 17 10 TO = 1 .06 x 10 23 y C i / c c Taken as 1.7 times that of Sr-90 Volume is calculated as follows: ? 2 ttD x 2 mi = tt (8,000 mi x 161,000 cm/mi) x 2 mi x 161,000 cm/mi = 5.3 x 10 23 cc. 8.53 BNWL-1 900 The total hypothetical concentra- tion in this volume is thus 1x10 uCi/cc. ' This hypothetical concentration compares with the space dispersion concentration as follows: Present weapons test isotopes dispersed in 3.2 kilometer (2 10 mile) thick layer around earth 1.0 x 10 uCi/ml Isotopes to vear 2000 -23 disoersed around sun 1.06 x 10 .. C i /ml The present hypothetical concentra- 1 3 tion is thus 1 x 10 times as great as the hypothetical space concentra- tion. Thus, there is much room for concentration factors in the space disposal system before concentrations around the earth begin to approach the present status. A more definitive analysis of the potential for space contamination is found in Appendix E. 8.3 RESEARCH AND DEVELOPMENT REQUIREMENTS The following section summarizes the estimated Research and Develop- ment requirements for extraterrestri- al disposal of radioactive waste transurani cs . Key Research and De- velopment milestones for a space dis- posal program (estimated at about 20 years duration) are summarized in Figure 8.12. Costs for research and development are estimated to be about 50 million dollars plus special space flight and vehicle development spe- cific to waste disposal. This latter cost element was very difficult to es- timate and is not estimated by NASA. The BNW study staff feels that the cost is in the order of 100 million dollars (a figure not discussed with NASA). Another disposal Research and Development cost is that for the re- maining fraction of the waste. From other disposal concept analyses in this report, these costs are expected to be in the range of 50 to 200 mil- lion dollars. Total Research and De- velopment costs for management of radioactive waste are therefore ex- pected to be in the range of 300 mil- lion dollars. Research and Develop- ment costs are highly preliminary estimates of near-low costs required for a viable program. A larger Re- search and Development program would probably be pursued to optimize costs and safety. The Research and Develop- ment costs shown here are only those related to transuranic waste disposal, and do not include the significantly higher costs required for general space vehicle development ($6.6 bil- lion estimated for first space shut- tle flights'- ■'), which are assumed to be part of the general space devel- opment program. Of the specific Re- search and Development tasks dis- cussed in the following sections, the space flight studies (both those gen- eral to space flight development and those specific to waste disposal), are expected to be the most time and money consuming. The concentration is calculated as follows 5.4 x IP 7 Ci x 10 6 u Ci/Ci ,23 1.0 x 10" 10 yCi/cc 5. 3 x 10 cc 8.54 BNWL-1900 TASK NO. TASK TITLE 1 WASTE CONTENT 2 DISPOSAL PRIORITY 3 PARTITIONING 4 CAPSULE DESIGN 5 ENCAPSULATION PROCESS 6 GROUND TRANSPORTATION 7 SPACEFLIGHT 8 ECONOMIC ANALYSIS 9 SAFETY ANALYSIS 10 PUBLIC RESPONSE YEAR FROM BEGINNING 10 15 20 1 2 3 4 5 6 7 8 9 10 11 12 KEY MILESTONES 1. WASTE CONTENT ESTABLISHED 2. DISPOSAL PRIORITY ESTABLISHED 3. PARTITIONING PROCESS DEVELOPED 4. CAPSULE DESIGN COMPLETED 5. CAPSULE DESIGN TESTED 6. ENCAPSULATION PROCESS ESTABLISHED 7. ENCAPSULATION PROCESS DEVELOPED 8. SPACE SHUTTLE FLOWN 9. SPACE TUG FLOWN 10. SPACE FLIGHT FULLY DEVELOPED 1L SAFETY ANALYSIS COMPLETED 12. PUBLIC RESPONSE SATISFIED F I GURE 8.12 . Research and Development Program Space Disposal of Transuranics 8.3.1 Waste Content At present the principal code used to establish isotopic content of dif- ferent wastes is 0RIGEN. The accu- racy of this code is suitable for the type of investigative work done to date. However, if extraterrestrial disposal is to be used for the acti- nide or transuranic wastes, a more accurate picture of true. isotopic con- tent is desirable than that obtain- able in present codes. Because of uncertainty and physics effects for extended exposure in dif- ferent types of reactors, existing data on the isotopic content of spent fuel is limited. A program to ana- lyze spent fuel from several differ- ent fuel cycles is required. This analysis is something that should be done for almost any part of the Waste Management Program. 8.3.2 Disposal Priority The purpose of this task is to put into more clear perspective the safety incentives and problems for space disposal of certain radio- isotopes. With such information available, better choices can be made, regarding their disposal needs. Though a number of analyses of the 8.55 BNWL-1 900 relative potential hazards of differ- ent isotopes exist, additional infor- mation would be advantageous for the analysis of space disposal. Past com- parisons of the potential hazards in space of one isotope, such as Pu-238 with another such as Pm-147, have not covered the total range of variables for full evaluation of the incentives for space disposal. Beyond the actinides, there are differences of opinion as to whether the very long-life fission products such as iodine-129 and techneti um-99 are sufficiently hazardous to require exotic disposal techniques. 8.3.3 Parti ti oni ng Partitioning is a fundamental re- quirement for execution of space dis- posal. Therefore, the timing and suc- cessful completion of partitioning process development is critical to space disposal. It is expected that suitable partitioning processes could be developed within about five years and that a commercial plant could be in operation five years from that time. Thus the availability of a com- mercial partitioning operation could be expected in about 10 years. Parti tioni ng, or separation of cer- tain constituents from the waste for special management, is discussed in Section 7 and those costs for parti- tioning research and development (3 to 5 million dollars) are not in- cluded in the total research and de- velopment costs for space disposal. The current primary possibility con- sidered most likely for the disposal of the long-lived isotopes in space is to separate the waste stream into fission products and transurani cs . Separating the transurani c and fis- sion product stream could provide a waste fraction for disposal in space which appears to be in reasonable form. Nonetheless, it may prove de- sirable to remove the curium from this product to minimize heat prob- lems for space disposal. It might also be advantageous to extract the neptunium for conversion to Pu-238 for other purposes. If long-lived fission products such as iodine-129 are truly shown to be a long-lived hazard, it might be beneficial to re- move the iodine from the fission prod- uct stream and ship it to space with the transurani cs . 8.3.4 Capsul e Design Flight qualified capsules have been designed and tested in periods ranging from 1 year to 3 years. With the existing background in capsule de- sign and available test facilities, it is expected that a suitable, fully tested capsule design could be pro- vided within 10 to 15 years. Higher priority could perhaps improve on this schedule if necessary. It appears possible today to de- sign a capsule which would satisfy stringent technical and safety cri- teria. However, there are a number of areas in which additional data and refinement of data and methods could improve the safety and reduce the overall cost of disposal. Typical of these are: • Criticality and Radiation Neutron - ics Calculations Additional code and experimental studies are required to refine criticality data and the radiation dose rates. 8.56 BNWL-1 900 • Heat Transfer Analysis Heat transfer under a wide variety of circumstances is one of the more critical aspects of design for space disposal. The capsule must be able to withstand such di- verse circumstances as deep burial in sand, re-entry, impact in space vacuum, and explosion of the rocket on the launch pad. Optimi- zation of materials, configuration and environment will require a great deal of refinement. • Refined Structural Analysis Some data exist on impact resis- tance of large capsules, but sig- nificantly more needs to be done to make this a practical system. Structure analysis starts with the containment of helium buildup within individual containers and extends through the impact resis- tance required to withstand re- entry, impact on solid materials, and extreme temperature conditions • Impact Testing Testing for structural considera- tions, shielding, burial corrosion resistance, and any other possible circumstances will be required. The Research and Development costs may be estimated by comparison of a somewhat similar past experience with radioactive heat sources for space. The design, fabrication, process development and testing of a single design of pi utoni um-238 isotopic heat sources for space missions has cost on the order of $500,000. ( 44 ^ However, because of the much larger size and quan- tities of waste capsules, and the consequently greater costs for impact in the fabrication and test- ing programs, it is expected that the Research and Development costs for waste packaging will be greater by at least a factor of 10. 8.3.5 Encapsulation Process Develop - ment Encapsulation processes are ex- pected to be similar to those exist- ing for space applications of radio- nuclide heat sources. Some of the key steps requiring development are: • Particle Preparation Processes exist for the prepara- tion and micro-encapsulation of particles. In general, however, these processes are applicable to small scale production. Develop- mental work is required to provide producti on- type processes. • Compacting of Matrix The development of processes is re- quired for compacting the matrix surrounding the fuel established in the capsule design. • Contai nment The containment shell surrounding the capsule will also require some process development studies. Ex- tension of existing fabrication techniques could likely be used, particularly if the shell is stain- 1 ess steel . All of the required steps for an encapsulation process have been test- ed, although not necessarily on the same materials that will be used. The application of existing processes to new materials has been success- fully demonstrated in the past in pe- riods from a few months to a year. Pilot plants for encapsulation of 8.57 BNWL-1 900 radioactive materials have been built in periods of 5 years or less. As an example of the program costs and times incurred in this type of de- velopment, a radi opromethi urn heat (45) source capsule was designed, ' the fabrication process developed and the source fabricated and tested in less than one year for less than $100,000. Testing for this capsule included rocket sled testing at terminal ve- locity and actual drop testing. In Fiscal Years 1964, 1965, and 1966 a program for the development of cermet fuels which involved the spheroi d i za ti on of uranium dioxide fuel particles, coating the particles with tungsten and consolidating them in the form of complex fuel shapes was conducted for NASA and the AEC. The fabrication process development portion of this work totaled $780,000 over the three year period. ' Again, because of the increased size, quantity, and complexity of transuranic waste capsules and the need for greater integrity, more ex- tensive process development by fac- tors of 5 to 10 than that required for smaller capsules is expected to be required for nuclear waste cap- sules. 8.3.6 Ground Transportation While conventional ground trans- portation systems will probably be applicable, some minor studies will be required to confirm analyses. 8.3.7 Space Flight As discussed in Section 8.2.4, there are vehicles which could be used today. Since the space shuttle is a vehicle which is likely to be used for waste disposal, operational use of the space shuttle is antici- pated. It is expected that the space shuttle will be operational in the early 1 980' s. The space tug is required for eco- nomical space disposal. A firm sched- ule for an operative space tug has not been established. However, NASA personnel have indicated that it can be operable by 1985. It could prob- ably be operational prior to this time if there was sufficient incentive. Developmental work in this area is likely to be all under NASA. Re- search and Development costs or times for this category are difficult to estimate. However, it is believed that development costs will be in the order of hundreds of millions of dol- lars beyond the currently planned space shuttle development costs. Some of the primary studies are ex- pected to be: • Disposal Trajectory The determination of a disposal trajectory which has no significant unknowns is one of the major tech- nical needs of this total program. Development in this area might in- volve propelling several waste containers into the sun and monitor- ing for detectable transuranics which might return during the next 5 to 30 years. Another experimental program could involve sending cap- sules in an earth escape trajectory or a high earth orbit and monitor- ing and analyzing their course. If such a path could be followed for 20 to 100 years it could increase the reliability of predictions on the future of such capsules. 8.58 BNWL-1900 • Vehicle Development NASA analysis to date has been based on existing concepts. Some de- velopment would undoubtedly be re- quired to adapt and optimize present concept vehicles to waste disposal. * Advanced Systems Advanced systems such as particle acceleration and solar sails have been examined only briefly. If space disposal is considered to be practical with current systems, it is probable that significant economic gains could be achieved through the development of advanced systems. While potential safety improvements are not apparent, there is some proba- bility that additional safety could also be achieved through the utiliza- tion of advanced systems. It is therefore expected that a much more detailed analysis of ad- vanced systems would be pursued be- fore space disposal can be imple- mented. If advanced systems show significant promise, they could war- rant a complete and extensive develop- ment program comparable to that for conventional vehicles described above . 8.3.8 Economic Analysis The economic analyses of space dis- posal alternatives will involve both a detailed engineering assistance study and a broad overview cost- benefit analysis. The engineering as- sistance work will provide continuous correlation of the economic effects of various technological approaches. 8.3.9 Safety Analysis The principal constituent of safety analysis is the execution of the space flight including potential accident considerations. Some other studies are required for the addition- al processing which must be done to put material in the form for flight. The safety aspects of flight will include the effect of earthly acci- dents such as explosion on the launch pad, failure to achieve orbit, or re- entry from a partial orbit. Space flight effects will include establish- ing realistic probabilities and conse- quences for malfunctions or accidents Safety studies will also include estimation of effects of space dis- posal on our environment. 8.3.10 Public Response Many arguments about nuclear waste are more emotional and philosophical than they are technical. A serious study of the emotional and philosoph- ical aspects of space disposal of waste must be aimed at developing con- crete conclusions for positive action. Action would involve communication in- terchange and education of the public and the technical community regarding the concerns of both. If space disposal is to become vi- able, national and international pol- icy must be examined and new policy establ i shed . 8.4 TIME REQUIREMENTS FOR COMMERCIAL OPERATION The time required to establish space disposal as a commercial opera- tion is dependent principally on the development studies discussed in the preceding section. Thus the time for commercial operation could be in the order of 20 to 25 years from the time 8.59 BNWL-1900 of actively starting Research and De- velopment? this includes up to 5 years ' from ful 1 development to full operation. Should experimental flights such as impact into the sun or monitoring a vehicle enroute to Pluto (20-year flight time) and beyond be considered necessary to the establishment of policy, then the time for commercial operation could be extended beyond 20 years . 8.5 CAPITAL AND OPERATING COSTS The costs of transuranic waste dis posal by extraterrestrial means in- clude three main components--parti - tioning, encapsulation, and flight. Basic capital and operating costs in 1974 dollars for waste management are developed in this section, and are converted to units of nuclear elec- trical power and nuclear fuel in Vol- ume 1 of this report series. Partitioning costs summarized below are taken from the separate sec- tion of this report which discusses partitioning in depth. Encapsulation costs are discussed below. Flight costs were discussed previously in part 8.2 of this report (technical feasibility), and are summarized here. All additional cost data were devel- oped in Volume 1 of this report and are summarized here. 8.5.1 Partitioning Partitioning costs incremental to reprocessing costs without partition- ing are estimated to be: Transuranium Elements (U and Th removed) in a package along with 1% of the fission prod- ucts cost $15,000/MT of fuel. Transuranium Elements (U and Th removed) in a package along with 0.1% of the fission prod- ucts cost $20,000/MT of fuel. 8.5.2. Encapsulation Costs For a reference reprocessing plant capacity of 1825 MT/year of LWR fuel, the waste transuranic through- put was calculated to be 1,288 kg/ year. To achieve a reasonable plant size it is assumed that the encapsula- tion plant will handle the transura- nium waste from two 1825 MT/year re- processing plants, or approximately 2,600 kg/year. Using the process of Figure 8.9, preliminary encapsulation costs were developed and are detailed in Appen- dix F. The total cost of encapsu- lation and heat shield is estimated at $4,700/kg of actinide waste, or $3,300/MT of original LWR fuel. The estimate assumes 1% fission product contamination, with heavily shielded hot cells used for each process step. Since with either 1% or 0.1% of fis- sion products in the transuranics, most of the work would require remote operation in hot cells, the costs would be similar for the 0.1% case. For checks of the order of costs, the fabrication cost of reactor fuel elements containinq high exposure plu- tonium has been estimated at $52 to 8.60 BNWL-1 900 $55/ kq of fuel material (uranium oxide and plutonium oxide) in a one- (47 ) ton-per-day plant. ' The cost of purifying and encapsulating cesium- 137 at $0.05/curie (48) is $4350/kg.^ a ^ Direct labor and eguipment reguire- ments are estimated to be, 17 person- nel (total cost $170,000/yr) and $675,000, respectively. These are detailed in Appendix Table F. 2 . The manpower and eguipment reguire- ments are based on exDerience in the spheroidization, coating and encap- sulation of materials such as pluto- nium, promethium, and polonium. These costs are detailed in Appendix Table F.3, based on the material make-up of a capsule as given in Appendi x Table F. 4. Direct material costs are esti- mated to be $340/kg transuranic. The costs for materials could be somewhat higher than those estimated because of unigue form or purity re- quirements. However,, a qreat effect on the total encapsulation cost is not expected. Indirect manufacturing expense in- cludes items such as inspection, testing, and maintenance. This is es- timated to be conservatively high at 100 percent of direct labor cost. Overhead costs are calculated at 100% of direct labor plus indirect manufacturing expense. Building capital costs are esti- mated by comparison with the Waste Encapsulation and Storage Facility (WESF) recently constructed at Hanford.^ ' This facility is de- signed for the annual encapsulation of 30 megacuries each of cesium-137 chloride in 60,000-curie packages, and strontium-90 fluoride in 150,000- curie packages. This radioactivity amounts to about 340 kilograms of cesium-137 and 210 kilograms of strontium 90 annually. The completion was late 1973 at a cost of $10.75 million. Using a 0.6 power scaling factor to convert from 530 kg/yr of (Cs+Sr) to 2600 kg/yr of transuran i cs , would provide a cost for the transuranic plant of about $27 ,000 , 000 . ^ C ^ Be- cause of the additional complexity of the transuranic waste encapsula- tion process, a figure of $40,000,000 is used for plant capital cost. As a comparison of the size order of these costs, the capital cost of a one ton/day high-exposure plutonium mixed-oxide fuel plant was estimated at $6,400,000 in 1966. ( 49 > 50 ) 8.5.3. Costs for Space Transporta t i on This section was outlined by NASA (1 2) Lewis ' and was taken from Section 8.2. Minor modifications in their report have been made to conform to the format of this report. a. Using 87 curies of Cs-137 per gram. b. Using 87 curies of pure Cs-137/g and 142 curies of pure Sr-90/g c. Cost 2600 kb = ($10,750,000) / 2 600\ ° • 6 \^530 J $27,000,000, 8.61 BNWL-1900 The vehicle costs, launch costs and operations costs are essentially a function of the number of vehicles, the launch rate, and the destination. Two representative destinations were estimated; high earth orbits and so- lar escape. The Kennedy Space Cen- ter was used as the reference launch facility. Beyond about the year 1990, significant additional space flight facilities will be needed. The transportation costs pre- sented below include technical sup- port and operations. 8.5.3.1 . Launch Costs The most cost effective vehicles were selected from Table 8.11 for determining the space transportation costs . • High Earth Orbit or Solar Orbit The most cost effective vehicle was the Shuttle with a Centaur of op- timum size (using approximately 17,240 kg of propellant). The launch cost for this vehicle is esti- mated at $16.3 million ($10.5 million for the reusable Shuttle launch and $5.8 million for the optimum expend- able Centaur). These costs are based on 100 total flights and not more than 40 flights per year. The Shut- tle launch cost of $10.5 million in- cludes propellant costs and opera- tional costs. If the flight rate is increased from 40 to 140 per year, the unit operational cost per flight would be reduced to approximately 75-90% of that for 40 flights/yr. • Solar System Escape This mission requires one Shuttle per Tug and at least two Tugs per nuclear waste payload. Table 8.12 indicates that the most cost effec- tive method for the solar escape des- tination involves perigee propulsion and consists of two Shuttles, a reus- able and an expendable Tug for a total cost of $28.8 million dollars per mission. The same conditions on launch rate apply to this cost as ap- plied in the Shuttles for the high earth orbit destination above. 8.5.3.2 Ground Facilities Costs Many of the existing facilities at Kennedy Space Center could be used on the basis of 20 flights per year. Beyond that, additional facili- ties would be required. In addition to launch and operations facilities, a new facility for receipt and in- spection of nuclear waste packages and for storage of packages in a con- trolled environment would be required Cost of this facility is estimated at approximately $4 million. As the number of launches per year increases past 20, additional facili- ties are needed. Projecting to 100 launches per year, an additional $140 million (- 20%) in mobile launchers, a new maintenance and checkout facili- ty, new crawler and maintenance fa- cility, a new launch pad and a new solid rocket booster disassembly fa- cility would be needed. After the year 2000 some other ar- rangement for launching facilities may be required since it would be dif- ficult to handle more than 120 to 140 launches per year at Kennedy Space Center . 8.62 BNWL-1 900 8.5.3.3 Total Space Transporta - tion Costs After the launch, other facilities come into use, such as tracking and monitoring stations and recovery teams and possible facilities for han- dling the Shuttle at other locations. Most of these facilities are assumed to already be in existence and would probably need only modifying. It would be expected that these modifica- tions would be an order of magnitude less than the costs for the 'facili- ties discussed above. The total transportation costs for both representative destinations are presented in Table 8.13 for launch rates of 40 per year. The facility cost per launch, based on a 30-year time period, increases from less than 0.1 to about 0.2 million dollars if the launch rate goes up by 100 per year, or a small amount. 8.5.4 Cost Summa ry The estimated cost to put material into a high-earth orbit or solar orbit is $1,950 per kilog-ram of gross payload, as shown in Table 8.14. The comparable estimated cost to escape the solar system to deep space is $8,800 per kilogram of gross payload. Encapsulation will increase the weight of transuranic actinide wastes and thus the unit costs by factors of about 15 to 30. Based on the above figures, ap- proximate costs which might be ex- pected for space disposal of transu- ranics with 0.1% fission product contamination are summarized in Table 8.15. These costs convert to 0.2 mills/kW-hr for high-earth orbit and 0.5 mills/kW-hr for solar system escape. These costs, when discounted to the time of reprocess- ing and allowing 0.04 mills/kW-hr for disposal of the remaining waste frac- tion (fission products) are 0.15 Mills/kW-hr and 0.34 mills/kW-hr, respectively. When compared with costs for the production of electric- ity of about 10 mills per kilowatt hour, the cost of space disposal of waste transuranics is thus likely to be less than b% of the cost of pro- ducing electricity. Additional costs would be incurred in the management of the remaining fraction of waste which is not disposed of into space. Based upon analysis of other poten- tial means of disposal elsewhere in this report, this incremental cost should not exceed 1% of the cost of nuclear electricity. 8.6 PUBLIC RESPONSE, P01ICY AND ENVIRONMENTAL CONSIDERATIONS 8.6.1 Public Response Public reaction to the concept of space disposal is unknown but is ex- pected to be mixed. The possibility of space disposal has been suggested many times in the past, but little factual information has been pre- sented which the public could use in forming an opinion. The fact that space disposal has the potential to remove certain radio- active waste constituents permanently off the earth (very distant from man's 8.63 BNWL-1900 TABLE 8.13 . Space Transportation Cost for Disposal of Transuranic Waste Basis: Reusable Shuttle 40 Payl oads/Year Cost/Pay! oad $Millions Facilities and Flight Preparation <0.1 High Earth Orbit Shuttle (1) 10.5 Centaur (Optimum) (1) 5.8 Total 16.4 Solar System Escape Shuttle (2) Tug-Expendable (1) -Reusable (1) Total 28.8 21 .0 6 .0 1 .75 TABLE 8.14 . Estimated Space Transportation Cost for Disposal of Transuranic Waste Basis: Space Shuttle is the base vehicle. Transuranium elements contain 0.1 or 1% of the fission products. Capsules are shielded for 1 Rem/hr from surface of package. Percent Fission Product in Waste Total Payload/Transuranic Weight, kg/kg Transportation Cost, $/mission Transportation Cost, $/kg of Payl oad Transportation Cost, $/kg of Transuranics High Earth Orbit or Solar r b i t ( a ) Solar System Escape 1 .0 0.1 1 .0 0.1 8400/288 8400/447 3270/113 3270/191 16.4xl0 6 16.4xl0 6 28.8xl0 6 28.8xl0 6 1950 1950 8800 8800 56.8xl0 3 36.5xl0 3 254.0xl0 3 150.0xl0 3 a. Transuranic waste is divided into three equal packages. 8.64 BNWL-1 900 TABLE 8.15, Cost for the Disposal of Transuranium Elements in Space Cost, $per kg Transuranic Cost, $per MT fuel Partitioning Encapsulation Space Flight Cost /payload= \ I 20 x actinide wt/ Total High Earth Escape from Orbit Solar System 14,000 4,700 36,500 55,200 14,000 4,700 150,000 168,700 High Earth Escape from Orbit Solar System 20,000 3,300 26,000 49,300 20,000 3,300 107,000 130,300 environment or population) is expec- ted to be a strong positive factor. The principal objections of crit- ics of space disposal appear to be concerned about high cost, unfavor- able energy balance, and safety dur- ing launch and space flight. Ques- tions regarding cost and energy are relatively straight forward to deter- mine, and therefore could be amenable to general public understanding. Questions regarding safety are much more complex and emotional, and there- fore would be expected to be more dif- ficult to achieve relatively uniform favorable public response. Other items which we might expect to be of concern to the public are the proof of high certainty that events will proceed as planned and phi 1 osophi cal -techni cal questions re- garding sending materials into "un- known" areas. 8.6.2 Policy Conflicts International and National poli- cies which could impact extraterres- trial disposal are discussed in sec- tion 3 of this report series and are reviewed briefly here. Any of the extraterrestrial dis- posal concepts would come under the provisions of the International Treaty on Outer Space , which was ratified by the United States in 1967. Article VII of this treaty provides that "Each State Party to the Treaty that launches or procures the launching of an object into outer space, including the moon and other celestial bodies, and each State Party from whose territory or facil- ity an object is launched, is inter- nationally liable for damage to an- other State Party to the Treaty or to its natural or juridical persons by such object or its component parts on the Earth, in air space or in outer 8.65 BNWL-1 900 space, including the moon and other celestial bodies." Any launch opera- tions or mission aborts could entail international liabilities under this treaty. An orbit degradation which caused a capsule to eventually return to earth might also be subject to these provisions. Liability for damage from any of the space concepts could be incurred under the terms of the 1972 United Nations' Convention on International Liability for Damage caused by Space Objects . The United States has not yet ratified this treaty, which de- fines in some detail the responsibili- ties of countries launching objects into space. For example, Article XII of the Convention specifies "...such reparation in respect of the damage as will restore the person, natural or juridical, state or international organization on whose behalf the claim is presented to the condition which would have existed if the dam- age had not occurred." While this convention will not materially extend the liability already accepted under the Treaty on Outer Space , it will, if ratified, mean acceptance of a pre- cise method for defining, assessing and paying for space-related damages. Article 25.1 of the International Convention on the High Seas provides that "Every State shall take measures to prevent pollution of the seas from the dumping of radioactive waste, tak- ing into account any standards and regulations which may be formulated by the competent international organi- zations." Any radioactivity which reached the ocean from a failed space flight might be deemed in violation of this convention. Any contracting party to the con- vention can request a revision five years after it has entered into force. The request for revision is consid- ered by the General Assembly of the United Nations. The International Nonpro 1 i f era ti on Treaty , to which the U.S. is a party, could conceivably affect any ultimate disposal concept. The basic objec- tive of the treaty is the limitation of nuclear weapons through the effec- tive control of the flow of source and special fissionable materials. Article III of the treaty states in part: "Procedures for the safe- guards required by this article shall be followed with respect to source or special fissionable material whether it is being produced, processed or used in any principal nuclear facil- ity or is outside any such facility. The safeguards required by this arti- cle shall be applied on all source or special fissionable material in all peaceful nuclear activities within the territory of such State, under its jurisdiction, or carried out under its control anywhere." While mixed fission products and transuranic elements miqht be consid- ered outside the scope of this treaty on the technical basis of cost and risk of separation of the fissionable materials, any concept dealing with separated transuranic element dis- posal would almost certainly fall under the agreements of this treaty. It would seem, then, that those proce- dures and operations would have to conform to International Atomic En- ergy Agency safeguards. Likewise, they would be subject to the appropri- ate international inspection and observation. 8.66 BNWL-1900 Any amendment to the treaty must be approved by a majority of all par- ties to the treaty and by all nuclear weapon states party to the treaty. Twenty-five years after the treaty is in force a conference is to be held to decide whether it should be ex- tended indefinitely or renewed for a fixed time period. This decision will be made by a majority of the par- ties to the treaty. Each party to the treaty has the right to withdraw if it decides that extraordinary events have "jeopar- dized the supreme interest of its country . " 8.6.3 Environmental Considerations The principal environmental consid- erations of the extraterrestrial dis- posal system will be in regard to the space flight operation. There will be some side effects such as the addi- tional processes required by parti- tioning and encapsulation. There will be additional fabrication of space flight vehicles and additional production of rocket fuel beyond that required without space disposal of nu- clear waste constituents. There will be additional transportation require- ments. If a disposal plan which in- volves only the disposal of transuran- ics is used, disposal would still be required for the fission products in the waste. However, it is possible that the fission products would be handled in a manner different from the combined waste since the lifetime requirements are much different. The incremental environmental ef- fects with reprocessing plants, the rocket fabrication plants and fuel production plants should not be unique. The principal effects considered for the environmental impact of space flight are land, resource use, water, air, ecologic impact, aesthetic impact and transportation impacts. Land The existing Kennedy Space Center site should be adequate until about the year 2000, assuming disposal of 10-year-old waste. Beyond that time, additional land and facilities roughly comparable to those at the Kennedy Space Center will be required about every 10 years. The expected effect of the Shuttle system on the Kennedy Space Center has been evalu- ated by NASA. Much of the discussion below is extracted from their pre_ f 51 ) liminary analyses. ' Large areas of land surrounding the launch facilities are required for supporting services and for a buffer between these activities and the surrounding community. At the Kennedy Space Center, maintenance of environmental stability and planned multiple land use have been stressed. For instance, under an agreement with the Bureau of Sport Fisheries and Wildlife, the boundaries of the Merritt Island Wildlife Refuge and the Kennedy Space Center are now co- extensive. This agreement provides that the Bureau, subject to certain conditions, exercise primary adminis- tration over all property (except the Space Program facilities) for all pur- poses unrelated to the Space Program. Legislation has also been introduced which, if enacted, would allow the 8.67 BNWL-1900 joint use of the Space Center area north of the Haulover Canal by the National Park Service as a National Seashore Park (rather than as a part of the present Wildlife Refuqe). The multiple land use was considered in the evaluation of candidate Space Shuttle launch and landinq sites, and activation of the selected building sites. Future activity will continue to stress land-use patterns compatible with the use of the area as a Wild- life Refuge and/or a National Seashore Park. Resource Use With space disposal, there would be potential adverse effects from the ultimate loss of some materials from the Earth. For example, the world re- serves of tungsten are 1,200,000 met- ric tons, with a potential world re- ( 52) source of 51,000,000 metric tons. At a launch rate of 100 capsules/year, over 100 metric tons per year of tung- sten would be irretrievably lost to space. Water In the space flight system, with planned recovery of all elements of the Space Shuttle except its fuel tank, the potential impact of the pro- gram flight aspects on water quality is related to: • On-pad accidents and propellant spills which may result in run-off of propellants to local drainage systems . • In-flight failures which may re- sult in vehicle hardware and propel- lant landing in the ocean. • Controlled re-entry of spent booster and Shuttle hydrogen and oxygen tanks . • Construction and operation of faci 1 i ties . Provisions such as dikes and catch basins are made for containing on-pad spills and disposing of the spilled propellant without major contamina- tion of the water (or air) environ- ment. Infrequent on-pad vehicle fail- ures would normally be expected to result in a fire that consumed most or all of the propellants. Thus, they are discussed in the section on air quality. Any unconsumed propel- lant would be treated in the same way as a spill. Potential sources of pollutants to the marine environment and the major pollutants would be: hardware, solid propellants, liquid propellants, lu- bricants, and hydraulic fluid- hydrocarbons . Possibilities of water pollution are primarily associated with water soluble, toxic materials which may be released to the water environment. Rocket propellants are the dominant source of such materials. Impact of the Shuttle fuel tank would release liquid hydrogen and liquid oxygen which would burn or evaporate rapidly into the atmosphere. Toxic materials contained in the Shuttle would be re- turned to the launch site. However, if the Shuttle were forced to abort to a water landing, these materials would enter into the water. The quan- tities from these infrequent occur- rences would dilute to non-toxic lev- els of concentration within the area affected by the emergency landing. The ammonium perchlorate in solid propellants is mixed in a rubber bind- er and would thus dissolve slowly in 8.68 BNWL-1 900 water. Toxic concentrations would be expected only in the immediate (within a few feet) vicinity of the propellant, if they occur at all. Oils and other hydrocarbon materi- als which are essentially immiscible with water, if released, may float on the surface of the water. Ouantities of hydrocarbons used are small. Jettisoned or re-entered hardware will corrode and thus contribute vari- ous metal ions to the environment. The rate of corrosion is slow in com- parison with the mixing and dilution rate expected in a marine environment, and hence toxic concentrations of metal ions are not expected to be pro- duced. The miscellaneous materials (e.g., battery electrolyte, hydraulic fluid) are present in such small guan- tities that, at worst, only extremely localized and temporary effects would be expected. In the immediate off- shore areas at Kennedy Space Center there is ample current to ensure dis- persion of these materials. The ground water sources in the area have been examined; no adverse effects are anticipated on the ground water . In 1970, Kennedy Space Center used approximately 3.8 million liters (one million gallons) per day out of a peak of 100,000,000 liters (26 mil- lion gallons per day supplied by the Cocoa Municipal Water Supply. In 1980, it is predicted that Kennedy Space Center will reguire about 3-4 million liters /day (0.9 million gal- lons/day) out of an estimated output of 130 million liters/day (35 million gallons/day). The fresh water re- quirements of Kennedy Space Center for the operational period are pro- jected to be less than that required for the Apollo program and will amount to only a relatively small por- tion of the overall demand for other resources . The chief potential for pollution on surface water is the propellants. In a normal launch essentially all propellants or propellant products are injected into the atmosphere, and the hardware, except for the propel- lant tank, is recovered. In the case of an in-flight failure in the early stages of flight, the booster and pro- pellant tank would probably impact in- tact. The shuttle would be expected to separate intact and return to the launch site. Handling of propellants at the Kennedy Space Center will follow the same procedures established and proven by the successful operations for the Apollo and other program ef- forts. The Kennedy Space Center has not had a fuel or oxidizer spill in excess of 3,8 liters (one gallon) ex- cept for a spill that occurred in 1969 during preparation for launch of Apollo/Saturn 505. Approximately 20,000 liters (5,300 gallons) of pro- pellant were captured in the spill pond provided for the purpose, col- lected and turned over to the Air Force to dispose of in accordance with accepted and established procedure . The construction of Shuttle fac- ilities and their operation is not ex- pected to affect aquatic areas. Air The quantity of combustion prod- ucts released durinq a flight is shown in Table 8.16. The result of most importance is the history of the 8.69 BNWL-1 900 TABLE 8.16 . Combustion Products of Concern Emitted by the Space Shuttle into Selected Atmospheric Layers Single Mission Quantity Emitted, kg Altitude Range Combustion Solid Rocket Shuttle Atmospheric Layer Kilometers 0-0.5 Product Motor Surface Boundary Layer CO co 2 37,200 6,600 HC1 31,900 ci 2 92 A1 2 3 43,300 H 2 15,900 19,500 Troposphere 0.5-10 CO co 2 HC1 ci 2 A1 2 3 113,000 20,100 96,900 278 132,000 H 2 48,100 62,200 concentration of the pollutant at ground level downwind of the launch point should wind currents move a por- tion of the cloud to the ground. In all normal launch cases, the peak con- centrations are well below the appli- cable maximum allowable 10-minute con- centration levels to industrial workers . Exhaust cloud concentrations of CO, HC1 , and A1„0- have been calculated as a function of distance downwind of the launch pad for an abort with burn- ing of solid rocket propellants. Peak concentrations are about 5 to 10 times larger for this case than for the normal launch, but would still be below the 10-minute maximum allowable concentration levels to industrial workers within the controlled area. At some locations downwind the rec- ommended limits for 10- and 60-minute exposures to hydrogen chloride for the general public may be exceeded as the result of a pad abort or fire. The time dependence of the concentra- tion at these locations is such that the time-averaged concentration is less than the recommended limits. The action taken to assure protection of the public will depend on exposure measurements within these zones. The National Academy of Sciences/ National Research Council Report sum- marizes the known effects of HC1 on wildlife. There would be no effects 8.70 BNWL-1 900 of even the predicted peak ground level concentrations of about 9 parts per million for a pad abort. NASA has analyzed the lower atmo- spheric effects and upper atmospheric effects on rain; no adverse analyses of pollutant effects of Shuttle opera- tions are foreseen. During rain possible precipitation scavenging of hydrogen chloride from the solid rocket exhaust cloud has been analyzed by NASA in an over-land trajectory. If possible harmful ef- fects of rain containing hydrogen chloride after the launch were antici- pated, the launch would be postponed. Ecologic Impact Except for the effect of noise on birds and animals, the major impacts on the ecology have already been suf- fered during construction of the Ken- nedy Space Center. During additional construction comparably small ecolog- ic effects would be expected for each comparable construction addition. Aesthetic Impact The visual impact on the community will be small since the facilities are in a reservation. The high vi- sual impact occurring at launch is of short duration. Noise The major source of noise associ- ated with the Space Shuttle program will be the noise generated by the rocket engine exhaust flow during en- gine tests and launches and* that of the sonic boom. The nature of the rocket engine noise may be generally described as intense, of relatively short duration, and spectrally com- posed of predominately low frequency energy. Noise pollution as a result of construction projects is not an- ticipated to be a problem, because all projects will be carried on within the boundaries of the Kennedy Space Center or equivalent, which in- cludes a large buffer zone for rocket launches. Space Shuttle operational person- nel within the launch area will be protected by personnel protective equipment or by isolation so that reasonable limits will not be ex- ceeded. Throughout the Apollo/Sat- urn V Program vehicles generated fre- quencies and intensities of the same order as those predicted for the Space Shuttle. Operational Observ- ers are stationed 3,500 meters (11,500 feet) from the launch pad in a small enclosure, and emergency crews are located approximately 550 meters (1,800 feet) from the launch site in standard armored personnel carriers. None of these personnel has suffered injury. Structural damage is possible with low frequency, hi gh- i ntens i ty noise. Therefore, structures within the con- trolled area will be designed to with- stand the noise environment to which they are to be exposed. For uncontrolled areas, a general noise exposure criterion of a maximum overall sound pressure level of 115 decibels, referenced to 0.00002 New- 2 tons/m , for both man and structures has been established by the Launch and Landing Site Review Board with re- spect to rocket engine noise. Nor- mally the acoustic energy which prop- agates into this region is of low frequency content, i.e., 100 Hertz and below. 8.71 BNWL-1 900 The Kennedy Space Center launch site meets the above criterion. The center has existing land area that is adequate for the Space Shuttle zones, and the noise generated will not af- fect the local area populace to any higher degree than previously experi enced . The runway orientation for Shuttle and logistics aircraft landing is such that overflights of populated areas and interference with existing civil air routes are minimized. Noise pollution affecting neighbor- ing communities is estimated to be significantly less than that emitted from commercial jet carrier opera- tions at the neighboring Titusville/ Cocoa Airport. The Kennedy Space Center buffer zone extends well into the Indian River west of the pro- posed new runway where the Shuttle is planned to land using commercial- type jet engines. Space Shuttle sonic booms occur during the ascent after launch, dur- ing booster descent after separation, and during Shuttle descent after re- entry from orbit. The most severe booms result when a vehicle engages in certain types of maneuvers that tend to amplify the overpressures. These maneuvers cause focusing of the sound pressures over a very small but predictable area on the surface. Focus on populated areas can be avoided by properly programming the flight maneuvers of the vehicles. Depending on mission orbit, return opportunity, and maneuverability, Shuttle re-entry sonic boom, however, may occur over land. The sonic boom characteristics for the returning shuttle have been calculated based upon extensive analytical work throughout NASA and on an exhaustive experimental program conducted by the Ames Research Center. In summary, the low overpressures, infrequent occurrence, and public awareness of sonic boom resulting from Shuttle return to the Kennedy Space Center lead to the conclusion that no significant environmental ef- fect to people, structural and natu- ral condition results. Some distur- bance to eagles, ospreys and other wildlife may occur. The degree of disturbance is expected to be minor, but can be evaluated more fully and accurately when Space Shuttle launch- ings begin in the existing planned space program. Transportation The transportation effects would be considerably greater from the large space launch components then from the "payload" capsul es carryi ng the transuranic waste constituents. In this environmental consideration, as in all others, it should be re- membered that only the disposal of the transuranic elements in waste are being considered here. If space dis- posal is used, all of the environ- mental considerations discussed for disposal of the remaining waste con- stituents will also be incurred. 8.72 BNWL-1 900 REFERENCES R. E. Hyland et al, Feasibility of Space Disposal of Radioactive Nuclear Waste, I, Executive Sum - mary , NASA TM X-2911, NASA Lewis Research Center, December 1973. K. Z. Morgan, W. S. Snyder, and M. R. Ford, "Relative Hazards of the Various Radioactive Mate- rials," Heal th Physi cs , vol. 10, pp. 151-169, 1964. 11 M. J. Bell and R. Long Term Hazard S . Dillon, The of Radioactive Waste 'Produced by the Enriched Uranium, Pu-238 U, and 233-U/Th Fuel Cycles , ORNL-TM-3548 , Oak Ridge National Laboratory, Oak Ridge, TN, November 1971. J. W. Healy, Surface Contamina - tion: Decision Levels , LA-4558-M, Alamos Scientific Laboratory, Alamos, NM, September 1971. Los Los M. J. Szulinski, Daughters of the Effect of the Actinides During Long Term Storage Hazard Consid - erations , Atlantic Richfield Han- ford Company, Richland, WA , Feb- ruary 8, 1972. 6. C. M. Slansky, "Ultimate Manage- ment of Radioactive Liquid Wastes," Chemical Engineering Progress Symposium Series , vol. 65, no. 97, 1969. 7. J. F. McCarthy, Jr. et al., Concepts for Space Disposal of Nuclear Waste, M assachusettc; in- stitute of Technology, October 1972. 8. W. Archer, Space Disposal of Nu- clear Waste, the Boeing Company, personal communication to Kirk Drumheller, Battel 1 e-Northwest , July 1972. 9. R. H. Miller, "Thinking Hyper- sonic," Astronautics and A eronau- tics , August 1971 . 10. J. P. Nichols et al, Projecti ons of Fuel Reprocessing Require - ments and High-Level Solidified Wastes from the U.S. Nuclear Power Industry , 0RNL-TM-3965 , Oak Ridge National Laboratory, Oak Ridge, TN, September 1972. 12 13, 14 15 16, 17. 18, 19, Department of Transportation, A Review of the Department of Trans - portation Regulations for Trans - portation of Radioactive Mate - rials , December 1972, Jr. 15. R. E. Hyland et al., Feas i bi 1 i ty Study of S p ace D i s p o sUTi of N u - ~WA37T cl ear group Waste NASA TM X TTE7T 2912 study vol . R. E. Hyland et al., Study of Ex - terrestrial Disposal of Radioac -" ti ve Wastes , Part 1 1 , ''Prel imi nary Feasibility Screening Study of Ex- traterrestrial Disposal of Radio- active Wastes in Concentrations, Matrix Materials and Containers Designed for Storage on Earth," NASA-TM-X-68147, Lewis Research Center, Cleveland, Ohio, October 1972. R. L pact Puthoff , Test of a A 640 Ft/Sec Im- Two-Foot Diameter Model N uclear Reactor Containment System without Fracture , NASA TM X-67997, Lewis Research Center, Cleveland, OH. H. T. Fullam and F. P. Roberts, Reactions of the Sesquioxides p.f Pm, Nd, and Dm with Water , BNWL-1421, Battelle-Northwest, Richland, WA, June 1970. G. P. Dix, "Nuclear Safety of Space Nuclear Power Systems," Na- tional Topical Meeting, American Nuclear Society, p. 358, Hunts- ville, AL, April 28-30, 1970. Mound Laboratory, Mound Labora - tory Isotopic Power Fuels Pro - grams: Apri 1 -September 197T T MLM-1827, Miamisburg, January 16, 1972. OH, J. E. Selle and K. L. Breakall , J4e ta 1 1 ographic Investigation of Two Large Radioisotopic Heat Source (LRHS) Capsules ,' MlM-1853, Mound Laboratory, Miamisburg, OH, September 29, 1 971 . R. E. Zielinski and D. E. Etter, Investigation of Supplementary Materials Used in the Pioneer RTG s and Capsule _S h i p_p j n g Contain - ers . MLM-1882, Mound Laboratory, Miamisburg, OH, January 12, 1972. .73 BNWL-1 900 20, 21 22 23. 24. 25 26, 28. F. L. Baker, R. C. Cranfill neer RTG Safety Evaluation: Phase I I mpa ct and Pio - Launch Abort Sequential Tests , SC=DR=71 01 01 , Sandia Laboratories, Albuquerque, NM, May 1971 . Baker, R E n t r e k i n C . Cranf i 1 1 , LASL Dart Mini thruster Safety Evaluation Jests, SC - R R - 7 1 00 1 2, Sand i a la bora to r 1 es Albuquerque, NM, February 1971. A. P. Fraas, G. Samuels, Isotope Kilowatt Program Quarterly Pro- gress Report for Period Ending June 30, 1971, ORNL-TM-3491 , Oak Ridge National Laboratory, Oak Ridge, TN, August 1971. E . Lamb, ORNL Isotopic Power Fuels Quarterly Report foj^ ^.e_r_i_p_d Ending' March 31 ,"1971 , " 0RNL"-4672 . Oak Oak Ridge National Laboratory, Ridge, TN, June 1971 . R. A. Robinson, "Prototype 244 Cm 2 3 Heat Source," Chemical Engineering Progress Symposium — 62- Series , vol 68, 1970. 66 , no . 106, pp . Forrest , Kernel for of the W. Ruehle and D. L "The Compact Co-60 Space Power," Proc Fourth Intersociety Energy Con - version Engineering Conference , pp. 321-329, Washington, DC 699036, September 22-26, 1969. Donald Kubose, Ming Lai, Harry Goya , Radioactivity Release from Radionuclide Power Sources IX. Release from Th ulium 170/1 1 Oxide and Promet hi urn- 1 47 Oxide to Seawater , N0LTR-71 -206, Naval Ord Laboratory, White Oak, Silver Spring, MD, December 8, 1971 . 27. W. J. Dalby, Status Report Im- pact Tests--Radio isotope Fuels and S imulants , SC-DR-710218, Sandia Laboratories, Albuquerque, NM, April 1971 . C. G. Anderson et al., Protec ted Radioisotopic Heat Source, U Patent 1969. 3,569,714, November S 14, 29. Martin R. Scheve, Low Ballistic Coefficient Radioisotope Heat Source , U.S. Patent 3,570,784, October 20, 1967. 30. S. E. Bramer et al., "Reentry Protection for Radioisotope Heat Sources," Nuclear Technology , vol . 11 , p. 232, June 1971 . 31. R. D. Baker, Quarterly Status Report on PI utoni urn- 238 Space - Electric Power Fuel Development Program ,' LA-4647 MS, Los Alamos Scientific Laboratory, Los Alamos, NM, July 1 - Septem- ber 30, 1970. 32. R. E. Hyland, M. L. Wohl , P. M. F i n n e g a n , Study of Extraterres- trial Disposal of Radioactive Wastes. Part III - Preliminary Feasibility Screening Study of Space Disposal of the Actinide Radioactive Wastes with 1% and 0.1% Fission Product Contamina- tion. NASA TM X-68216, Lewis Research Center, Cleveland, OH, April 1973. 33. Code of Federal Regulations , Title 10, Part 71, "Packaging of Radioactive Material for Trans- port," U.S. Government Printing Office, Washington, DC, Janu- ary 1 , 1972. 34. B. Griggs, Ba ttel 1 e-Nor thwes t , personal communication to Kirk Drumheller, Ba ttel 1 e-Northwes t , Richland, WA . 35. C. L. Brown, Battel 1 e-Northwes t , personal communication to Kirk Drumheller, Ba ttel 1 e-Northwes t , Richland, WA . 36. K. Drumheller, "Radioactive Source Encapsulation," Chapter 3, pp. 135-179, Radioisotope Engi - neering , Geoffrey G. Eichholz, editor, Marcel Dakker, Inc., NY, 1972. 37. D. E. Thomas, W. 0. Harms, R. T. Huntoon, editors, "Sympo- sium on Materials for Radio- Isotope Heat Sources," Nuclear Metal 1 urgy , vol. 14, Proc. of the 1968 Nuclear Metallurgical Symposium held in Gatlinburg, TN. , October 2-4, 1968. 38. Mound Laboratory, Mound Labora- tory Isotopic Power Fuels Program: January-March 1971 , MLM-1817, Miamisburg, June 15, 1971 . OH, 8.74 BNWL-1900 39 40, 41 42 43 44. 45. 46. D. L. Prosser, Ex amination of SNAP-27 .Heat _So ufc e 1 CA 4__ and Preparation of Fuel for FCA8 , MLM-1831, Mound Laboratory, Miamisburg, OH, July 23, 1971. F. D. Postula, N. B. Eisner, M . R . E m k e n , Lightweight , Low-Ba 1 1 i sti c Coe fficient , Radioisotop e Fjje lj d Ca psule Prototy pe Development Program - Final Report . GA-8998, Gulf General Atomic Co., San Diego, CA. , May 1971 . F. D. Postula, N. B. Eisner, M . R . E m k e n , Lightweight, Low - Ba 1.1 .1 st i_c_ Co ef fi cj.ent L jR a d j. pj_s_o - tjope Fu£lsd Capsule Prototype _D.e- velooment Program , Fina l , Report , S_u pp . B, Test Dataand AnaJ ysis , GA-8998, Suppi . ' B~, Gifjf General Atomic Co., San Diego, CA., May 1971 . J. H. Harley, "Worldwide Pluto- nium Fallout from Weapons Tests," Proceedings of Environmental Plu- 47 48. tonium Symposium , August 4-5, 1971 . USAEC-LASL "Next in Space: A Commuter Special," U.S. News and World Report , Apri I 22, 1974. R. K. Robinson, Exxon Nuclear, Richland, WA, personal communica- tion to Kirk Drumheller, Bat- tel 1 e-Northwes t , Richland, WA . N. C. Davis and D. W. Brite, " Promethi um-1 47 Radioisotope Ap- plication Program," AMSA Heat Source Final Report , BNWL-994, Battel le-Northwest, Richland, WA , March 1969. Personal Notes, K. Drumheller, 3R, 3P, and 3Z Programs, Bat- tel 1 e-Northwes t , Richland, WA . 49, 50, 51 J. B. Burnham, L. G. Merker, Com - parative Costs of Oxide Fuel Ele - ments, Vol. 1 - The Computed Costs of Fuel Elements (Vibra - tional Compacted or Pellet)~n - riched with U-235 or Low Expo - sure Plutonium or High Exposure Plutonium , BNWL-273 , Battel 1 e- Northwest, Richland, WA , July 1966. R. W. McKee, Techn i cal , Eco- nomic, and Pol i cy Consi derations Affecting Future Production, Mar - keting and Use of Cesium-137 and Strontium-90 , BNWL-1686, Bat- tel 1 e-Northwest , Richland, WA , December 1972. J . B . Burnham, L . Comparative Costs G. of Merker, Oxide Fuel Elements, Vol. 3 - The Effect on Fuel Element Costs When Certain Reference Values are Varied , BNWL-273, Battelle-Northwest, Richland, WA, August 1966. J. B. Burnham, D. E. Deonigi, L. G. Merker, Comparative Costs of Oxide Fuel Elements - Vol. 2 , Process Facility Description and Cost Estimates , Battelle- Northwest, Richland, WA , Decem- ber 1966. R. Hyland et al., personal com- munication of preliminary infor- mation, NASA Lewis Research Center, Cleveland, OH, to Kirk Drumheller, Battelle-Northwest, Richland, WA . 52. United States Mineral Resources , U.S. Geological Survey Profes- sional Paper 820, U.S. Govern- mental Printing Office Stock No. 2401-00307, 1973. BNWL-1900 SECTION 9: TRANSMUTATION PROCESSING Section 9 Contributors R. C. Liikala, Study Leader J. B. Burnham B. F. Gore B. R. Leonard, Jr. D. L. Lessor C. W. Lindenmeir T. I. McSweeney E. T. Merrill W. C. Wolkenhauer 9 . i i i BNWL-1900 TABLE OF CONTENTS 9.3 TRANSMUTATION PROCESS I NG/ EL IM I NAT I ON TRANSMUTATION WASTE MANAGEMENT SYSTEMS 1 Overall Criteria 1.1 Overall Energy Balance 1.2 Overall Waste Balance 1.3 Specific Transmutation Rate 1.4 Total Transmutation Rate 2 Description of the Concept Systems 2.1 Accelerator Devices 2.2 Fission and Thermonuclear Explosive Devic 2.3 Fission Reactors 2.4 Fusion Reactors TECHNICAL FEASIBILITY . 1 Accelerator Devices 1.1 Charged Particle - Nuclear Reactions 1.2 Beta Decay Acceleration by Coulomb Excitation 1.3 Photon Transmutation Processes 1.4 Spallation Accelerators . 2 Fission and Thermonuclear Explosive Devices 3 FissionReactors 3.1 Fission Product Transmutation 3.2 Actinide Transmutation 4 Fusion Reactors ..... 5 Candidate Transmutation Concept System 5.1 System Description .... 5.2 Requirements for the Concept System ESTIMATED RESEARCH AND DEVELOPMENT REQUIREMENTS 1 Task 1 - Actinide Recycle in Fission Reactors 1.1 Phase 1 - Evaluate Efficacy of Actinide Transmutation in Fission Reactors 1.2 Phase 2 - Development and Demonstration of Actini Recycle in Fission Reactors .... 2 Advanced Concept Evaluations .... ESTIMATED TIME FOR REQUIREMENTS FOR OPERATION CAPITAL AND OPERATING COSTS 1 System Characteristics for Cost Bases 2 Estimated Costs ....... 3 Alternate Fission Schemes Which Might Reduce Cost PUBLIC RESPONSE POLICY CONSIDERATIONS ENVIRONMENTAL CONSIDERATIONS .... 1 Incremental Increase in Fuel Cycle Processing . de 1 1 2 2 2 3 3 3 4 5 5 8 10 11 11 12 12 13 15 16 16 17 21 23 23 24 25 26 9.27 29 30 30 31 32 33 35 36 36 9.37 9.37 9. i v BNWL-1900 9.8.2 Incremental Impact from Increased Uranium Requirements 9.8.3 Incremental Impact of a Combined Reprocessing and Fuel Fabrication Plant for LWR Actinide Transmutation Fuel 9.8.4 Environmental Impact from Waste Solidification Processing 9.8.5 Environmental Impact from Transportation . 9.9 SAFETY ASPECTS OF ACTINIDE RECYCLE IN LWRs . 9.9.1 Effect of Actinide Recycle in LWRs on Accidents in Fuel Processing Plants ..... 9.9.2 The Effect of Actinide Recyle in LWRs on Accidents Other Fuel Cycle Components .... REFERENCES APPENDIX 9. A .Transmutation by Accelerators APPENDIX 9.B Transmutation by Fission and Thermonuclear Explosive Devices ..... APPENDIX 9.C Transmutation by Fission Reactors . APPENDIX 9.D Transmutation by Fusion (CTR) Reactors . APPENDIX 9.E Comments of Peer Reviewers 9.38 9.38 9.41 9.41 9.42 9.42 9.43 9.44 9.A.1 9.B.1 9.C.1 9.D.1 9.E.1 9. v BNWL-1900 LIST OF FIGURES 9.1 Concept for Transmutation by Accelerator Devices 9.2 Concept for Transmutation by Fission and Thermonuc Explosives .... .... 9.3 Present Nuclear Fuel Cycle . .... 9.4 Concept for Transmutation by Fission Reactors . 9.5 Concept for Transmutation by Fusion (CTR) Reactors 9.6 Actinide Recycle Strategies Studied 9.7 Short-Term Cumulative Hazard of Actinide Waste fro Operation of a Typical PWR . .... 9.8 Long-Term Cumulative Hazard of Actinide Waste from Operation of a Typical PWR . .... 9.9 Transmutation Waste Management Strategy 9.10 Projected Nuclear Power Economy in U.S.A. . 9.11 Overall System Requirements for Managing High-Leve Radioactive Wastes by Transmutation Concept 9.12 Estimated Research and Development Program for Act Transmutation in Fission Reactors ear ni de 60-Year 60-Year 9.4 9.5 9.6 9.7 9.9 9.18 9. 19 9.19 9.23 9.24 9.25 9.28 LIST OF TABLES 9.1 Summary of Transmutation Device Feasibility ... 9.2 Transmutation Rates Using Thermal Neutrons from a SpallationAccelerator . . 9.3 Properties of Several Important Fission Product Nuclides and Time Required for 99.9 Percent Reduction of Their Inventory by Decay and Neutron Transmutation ... 9.4 Comparison of Actinide Inventories for Two Recycle Strategies Using U0 2 Fuel in a PWR . . 9.5 Effect of Recycling on Hazardous Radionuclides in the Reactor and the Processing Plant 9.6 Estimated Annual Incremental Fuel Cycle Costs for Trans- mutation of Actinides in LWRs Assuming Remote Fabrication of 10 Percent of the Fuel . 9.7 Incremental Fuel Cycle Cost for Increased Enrichment Due to Recycle of Actinides . . 9.8 Additional Uranium Required to Supply 1,825 MT/Year of Actinide Recycle Fuel . . 9.9 Summary of Incremental Environmental Considerations for Uranium Mining, Milling, Fluoride Conversion and Enrichment for Actinide Recycle in LWRs . 9.10 Uranium Fuel Reprocess ing-Fabri cati on Plants at Individual Sites - Gaseous Radioactive Effluents from Processing 1 ,825 MT/Year 9.11 Transmutation Fuel Reprocess ing-Fabri cati on Plant Gaseous Radioactive Effluents from Processing 1,825 MT/Year 9.12 Radiological Discharges from Waste Solidification . 9.10 9.14 9.17 9.21 9.33 9.34 9.34 9.37 9.39 9.40 9.40 9.41 9.1 BNWL-1900 9.0 TRANSMUTATION PROCESSING/ELIMINATION A possible approach to the manage- ment of radioactive waste is the use of nuclear processes themselves to change (transmute) the toxic long- lived radioactive cons ti t uten ts in high-level waste into short-lived ra- dioactive or nonradioactive isotopes. Transmutation is generally defined as any process whereby a nuclide ab- sorbs or emits radiation and is thereby changed into another nuclide. It is proper that in the study of waste management alternatives consid- eration be given to the possibility of eliminating the waste by trans- mutation. Since many of the toxic materials were formed from transmu- tation processes, it is conceivable that they can be destroyed by the same processes. More practically, the transmutation process can accel- erate the decay rate of radioactive waste by converting long-lived radio- isotopes to other isotopes which have shorter decay times. If this can be achieved, the quantity of waste containing long-lived radio- nuclides could be reduced signifi- cantly and the time required for safely storing treated radioactive waste may be substantially shortened. The transmutation concepts consid- ered in this study were limited to managing high-level radioacti've waste from nuclear power. All constituents of high-level waste for example, short-lived fission products, are not amenable to treatment by transmuta- tion processing/elimination (here- after referred to as transmutation). Therefore, this study mainly focused on the problem of reducing the long- term toxicity of high-level waste. This section of the report de- scribes known processes which could potentially be used for transmuting toxic isotopes into nontoxic iso- topes. The study is extensive enough to determine which fundamen- tal processes are theoretically not possible in light of current tech- nology. At the same time the study was not exhaustive in that little or no optimization of possible pro- cesses was attempted. Thus, these initial conclusions of technical and/or economic infeasibility of transmutation should not dismiss future research entirely since it is the only method that has the poten- tial for accelerating natural aging processes of eliminating waste. 9 .1 TRANSMUTATION WASTE MANAGEMENT SYSTEMS On the basis of an extensive lit- erature review and private communica- tions with other professional inves- tigators the various ideas proposed for transmuting nuclear waste were collected. The waste management transmutation ideas were divided into four categories based upon the type of physical device conceived for accomplishing the transmutation. A tentative description of each concept system was developed and criteria were developed as a basis for judging technical feasibility. The four concepts are the use of: 9.2 BNWL-1900 1) Accelerators 2) Thermonuclear Explosives 3) Fission Reactors 4) Fusion Reactors Waste transmutation processes con- sidered in concept (1) involve the use of several different nuclear and photonuclear particles as sources for accomplishing transmutation. In par- ticular, charged particle bombardment employing both protons and electrons has been studied. Photon radiation, both bremss trah 1 ung and stimulated emissions, has also been investigated. Finally, neutron radiation is a candi- date transmutation source. The study of concepts (2) through (4) were con- sidered only on the basis of using these devices as neutron sources. 9.1.1 Overal 1 Cri teri a To establish the relative merits and specific technical feasibility of the various approaches, a number of criteria were developed and applied to the proposed concepts. These include : overall energy balance, overall waste balance, specific transmutation rate, and total transmutation rate. Throughout this section, the merits derived from transmutation are viewed in terms of toxicity indices, ' permitting easy comparison of a 1 terna t i ves . 9.1.1.1 Overall Energy Balance Overall energy balance is a funda- mental criterion which applies to all advanced concepts. For a process to be practical it must not consume more energy than was originally obtained when the waste was created. This cri' teri on is strictly valid only so long as fission power reactors are contrib- uting significantly to the energy economy. If fission reactors are fur- nishing the energy for the transmuta- tion process and if this criterion is not satisfied, as much waste is being created as destroyed. If fis- sion power were no longer necessary, this criterion is no longer one of feasibility but of practicality. It might then be acceptable to the pub- lic to violate this criterion if the energy source were sufficiently plen- tiful and environmentally clean. In considering two processes, the one which consumes the least energy per transmutation is the more attractive for this criterion. 9.1.1.2 Overall Waste Balance The second criterion applied to the various transmutation concepts was the overall waste balance. For a process to be feasible it must re- move more long-term toxic waste than it creates. This criterion is of a. Toxicity index is defined as the base 10 logarithm for the amount of air or water in cubic meters required to dilute the present amount of a given isotope to levels defined in the Code of Federal Regul ati ons (10 CFR - Part 20) as the maximum permissible concen- tration. The toxicity index provides only an approximate compari- son of radiological risk, since it does not allow for accumulation or reconcen trati on of a nuclide in environmental media, nor for the total impact of a number of nuclides. For limited comparisons, it is an acceptable alternative to dose calculations if used with cauti on . 9.3 BNWL-1900 particular concern in transmutation because the process is similar to that which originally created the waste. The assurance that this criterion is met depends upon a rather detailed analysis. Some of the elements nec- essary for this analysis are knowl- edge of the nuclear reactions used and energy and type of source nu- cleon. In addition, detailed knowl- edge of the proposed mother-daughter chain must be available. In summary, it needs to be demonstrated that selection of a particular transmuta- tion path does not result in the creation of descendant isotopes which are ultimately of comparable toxicity to the waste itself. 9.1.1.3 Specific Transmutation Rate The third criterion applied to the various concepts was the specific transmutation rate of the process. Any isotope in waste which is a can- didate for transmutation has a natu- ral rate of decay. Any transmutation process has a rate of transmutation which is the product of the nucleon flux times the probability of the specific transmutation reaction. (The latter parameter is known as the cross section of the reaction.) For a transmutation process to be feasi- ble the transmutation rate must be of the same order as the natural decay rate so that the actual removal rate is increased. To be a highly success- ful process, the transmutation rate should be several times that of the natural decay rate. 9 .1.1.4 Total Transmutation Rate For a process to be feasible it must be capable of transmuting a sig- nificant fraction of the expected inventory of radioactive waste. In addition, adequate transmutation sources must be available in suffi- cient quantities over the total pro- posed history of the process for the process to be feasible. Here, the total rate of the power industry to transmute its own waste is consid- ered. For example, if neutron trans- mutation is the process under consid- eration, this criterion is concerned with the available neutron supply as a function of t i me . Application of this criterion re- quires the examination of the avail- able sources for transmutation as a function of time. If alternative sources cannot be shown to be avail- able in sufficient quantities when primary sources are no longer avail- able, then transmutation will not be f eas ibl e . 9.1.2 Description Of The Concept Systems As mentioned above, the ideas pro- posed were categorized in terms of accelerator devices, thermonuclear explosive devices, fission reactors, and fusion reactors. A description of each concept system was developed In the course of study it was con- cluded that the use of accelerators and thermonuclear explosive devices was not feasible. Therefore, only brief descriptions are given for these two concept systems. 9.4 BNWL-1900 9 .1.2.1 Accelerator Devices The concept of this system, de- picted in Figure 9.1, has a parti- tioning stage and a target fabrica- tion stage between reprocessing and transmutation using the accelerator. The bulk high-level waste from a re- processing plant, either in aqueous or solid forms, could not be effec- tively transmuted using accelerator devices. The waste partitioning fa- cility shown in the figure is to separate the bulk high-level waste by partitioning the waste into sepa- rate chemical element streams or a mixture thereof. The feasibility study of transmutation of only the cesium and strontium fission product components of the waste by the neu- trons from a spallation accelerator has indicated that one accelerator would be required for every three or four fission reactors of the same thermal power. If other components of the high-level waste were also competing for these neutrons, the number of required accelerators per fission reactor would make the pro- cess no longer feasible. The con- cept would fail both the criteria of energy balance and waste balance. Therefore, the waste would have to be processed (i.e., partitioned) to chemically separate the waste into fractions with different character- istics. After irradiation by the acceler- ator neutrons, the target material would contain some fraction of the original radioactive waste since it would not be feasible to irradiate until 100 percent of the waste iso- topes had undergone transmutation. In addition, the target material would contain highly radioactive short half-life daughter isotopes and radioactive elements due to trans- mutation in any structural material which contains the sample. These residual radioactive materials would be transported to the retriev- able waste storage facility for decay and ultimate disposal using other means . WASTE PARTITIONING \ TARGET FABRICATION FACILITY / BULK HIGH-LEVEL WASTE TRANSPORTATION CASKS ACCELERATOR DEVICE ♦ STABLE ASH TO DISPOSAL RADIOACTIVE MATERIALS •0 I HE I Hit / RETRIEVABLE STORAGE ULTIMATE DISPOSAL FIGURE 9.1 . Concept For Transmutation By Accelerator Devi ces 9.5 BNWL-1900 9.1.2.2 Fission and Thermonu - clear Explosive Devices The overall pictorial representa- tion of this concept is shown in Fig- ure 9.2. A partitioning step to segregate the various toxic speci- mens is shown since it would prob- ably be required in this concept. This explosive device concept does not appear feasible in part because of the large number of explosions estimated to be required to achieve the required total transmutation rate. If the desired constituents of the high-level waste were not separate from the rest, the material present in the target which was not a candi- date for transmutation would compete for the available neutrons. Thus more explosions would be required and the concept would be less feasi- ble than it appears to be when only selected constituents of the waste are considered as targets. Selected materials would be shipped to a tar- get fabrication facility and, hence, to the disposal site. The targets would be lowered into a drilled hole along with the thermonuclear explo- sive device, the hole sealed, and the device set off. The neutrons created in the explosion would trans- mute some constituents of the waste. If it were possible to use the solid- ified encapsulated waste from the reprocessing step directly as the target, the waste canisters received from the reprocessing facilities would be sent directly to the trans- mutation site, eliminating the parti- tioning and fabrication stages. 9.1.2.3 Fission Reactors The Present Fuel Cycle . The trans- mutation of actinide waste in fission reactors has been shown to be a feasi- ble concept. Review of the present fuel cycle shown in Figure 9.3 helps to put this actinide transmutation concept in perspective. Uranium is TARGET FABRICATION FACILITY WASTE PARTITIONING FACILITY \ rSHHS^-B BULK HIGH-LEVEL WASTE TRANSPORTATION CASKS TARGET AND EXPLOSIVE PLACEMENT FACILITY % \Tit FISSION OR THERMONUCLEAR EXPLOSION CHAMBER SEALING DEVICES FIGURE 9.2 . Concept For Transmutation By Fission And Thermonuclear Explosives 9.6 BNWL-1 900 ENERGY FABRICATION FACILITY i URANIUM TO ENRICHMENT FACILITY REPROCESSING FACILITY RETRIEVABLE STORAGE FACILITY FIGURE 9.3 . Present Nuclear Fuel Cycle received at the fabrication facility from an enrichment facility and fab- ricated into fresh U0„ fuel which is shipped to the reactor. After irra- diation in the reactor to expected exposures of ^27,500 MWD/MT^ and ^33,000 MWD/MT for boiling and pres- surized water reactors (BWRs and PWRs ) , the spent fuel is sent to reprocessing where U and Pu are re- covered. Present reprocessing recovery efficiencies for U and Pu are around 99.5 percent. The radio- active waste to be sent to a federal repository, ' consists of the resid- ual U and Pu, all other actinide elements, and the fission products. The uranium recovered from commercial light water reactors (LWRs) is usu- ally stored at the reprocessing facil- ity. After a cooling period it may be sent to enrichment facilities to be used as feed material. It can also be sent directly to a fabrica- tion facility to be incorporated into fresh fuel. The recovered uranium from fast breeder reactors (FBRs) will be recycled for blanket mate- rial to produce plutonium. The re- covered plutonium is also stored at the reprocessing facility. After storage it may be sent to a fabri- cation facility to be used in making fresh L)0 2 -Pu0 2 fuel . A key item in the present nuclear fuel cycle is the economics of re- covering uranium and utilizing the plutonium. If the economics dictate Megawatt days per metric ton of fuel. 9.7 BNWL-1 900 reprocessing the uranium because of its residual enrichment value, then the plutonium may either be stored for fast breeders or recycled in thermal reactors. If economics indi- cate that the uranium is not worth recovering and that the plutonium can be saved for fast breeder reactors, then this fuel may not be reprocessed until the time the plutonium is needed in fast breeder reactors. In this event, the spent fuel would prob- ably be stored in water basins for the time period between discharge from the reactor and the advent of the breeder market. Another alterna- tive is to reprocess LWR fuel to re- cover plutonium irrespective of the worth of the uranium for recycle. In any event, fissionable material has inherent value in that it can produce energy. Assignment of this economic worth will be the crucial factor in determining when to reprocess LWR fuel. These economic arguments have not taken into account the costs of waste management. The costs of manag- ing waste may have some impact on the timing of reprocessing and choice of i sotopes to recy cl e . Transmutation Cycles . This con- cept considers recycling constituents of high-level waste back to the reac- tor for subsequent transmutation to less toxic species. The concept is depicted in Figure 9.4. Specific constituents in spent fuel coming from the reactor are recovered dur- ing reprocessing and partitioning and are sent to a fuel fabrication facility for incorporation into rods for subsequent insertion in the reactor . FABRICATION FACILITY REACTOR URANIUM FABRICATED FUEL TRANSPORTATION CASKS SPENT FUEL RECYCLE MATERIALS -q REPROCESSING FACILITY PARTITIONING FACILITY NON RECYCLE MATERIALS _ RETRIEVABLE STORAGE FACILITY FIGURE 9.4 . Concept For Transmutation by Fission Reactors 9.8 BNWL-1900 The technical feasibility for this concept is assessed by considering the constituents of the waste as being three categories of radionu- clides. These categories are based upon the length of time the waste must be securely stored such that the remaining quantities of radionu- clides represent no significant threat to the health and safety of the general public in the event of uncontrolled releases. The catego- ries are : 1. 1000 years or greater 2. 100 to 1000 years 3. Less than 100 years In high-level waste the actinide elements and perhaps the fission product 1-129 fall in the first cate- gory. The fission products Sr-90 and Cs-137 are the primary constitu- ents of the second category. Most of the other fission products fall in the third category. The reason certain fission products occur in the high-level waste in the amounts they do is that the probability for their undergoing transmutation in fission reactors is very small (i.e., small nuclear cross sections). As a re- sult, they do transmute in-situ. The radionuclides in Category 3 are basically of this type and therefore are not viable candidates for transmu- tation in fission reactors. Some of the fission product radionuclides in Categories 1 and 2 are likewise of this type. A potential scheme for transmu- tation appears to be chemically re- processing the fuel, recovering all of the uranium and plutonium (i.e., eliminate process losses), segregat- ing those species of high-level waste which represent significant interme- diate and long-term toxicity from those which rapidly decay to stable isotopes, and treating the signifi- cantly toxic species via transmutati on . As discussed in Section 7 of this volume, current technology for chem- ical reprocessing of nuclear fuel does not, however, provide product or waste streams according to these cate- gories. Therefore, technological im- provement is needed to separate acti- nide elements, and certain of the fission products (e.g., Sr-90 and Cs-137) from the balance of the waste stream. This partitioning is needed to recover these materials for trans- mutation in fission reactors. If partitioning and transmutation can be achieved, then the toxicity of these wastes may be effectively reduced to levels commensurate with those of Category 3 materials. Category 3 materials could then be managed by means such as storing in a retriev- able storage facility. 9.1.2.4 Fusion Reactors Fusion reactors, commonly referred to as Controlled Thermonuclear Reac- tors (CTRs), potentially can provide a copious supply of neutrons which could be used for effective transmu- tation of nuclear waste. The mean energy of neutrons produced in fis- sion reactors is about 2 MeV, whereas the mean energy of neutrons produced in deute ri urn- tri ti urn fusion reactors is about 14 MeV. The higher energy neutrons in a fusion reactor represent a better source for neutron- induced transmutation since high- energy neutron reactions can be used 9.9 BNWL-1900 directly or the neutrons the rmal i zed to provide an intense source of 1 ow- energy neutrons. A schematic of this concept is presented in Figure 9.5. As in the case of fission reactor transmutation, it is likely that the waste stream from the reprocessing plant would have to be partitioned to obtain separate product streams for the actinide elements, strontium and cesium, and perhaps others in order to be managed in a CTR transmutation device. The factors which must be consid- ered in the cycle of processing, fab- rication, and irradiation in the blan- ket of a fusion reactor are similar to those of transmutation in a fis- sion reactor. The irradiation time in the fusion reactor is limited by two factors. One is the mechanical integrity of the canning material due to the intense neutron radiation field. A second factor is the de- creased efficiency of the transmu- tation process with irradiation time because of the decreasing concentra- tion of target nuclei and the compe- tition of daughter nuclei for the available neutrons. After some ir- radiation period, which has not yet been defined, the waste target must be discharged from the reactor. The discharge material will probably con- tain a significant fraction of the original target material. In addi- tion, it will contain some short half- life daughter isotopes and radioac- tive container material. It appears likely that the toxicity of the original target material will not be reduced sufficiently in one irradia- tion step that it could be sent to storage. More likely, the material would be returned to and reprocessed in the partitioning facility. After one cycle of irradiation, the parti- tioning facility would probably have effluent streams of nonradioactive compounds and of radioactive short- lived compounds. The latter would be sent to storage and, subsequently, PARTITIONING FACILITY TRANSPORTATION CASK TARGET FABRICATION TRANSPORTATION FACILITY CASK CTR TRANSMUTATION PRODUCTS RETRIEVABLE STORAGE FACILITY SHORT LIVES AND STABLE NUCLIDES I ► ULTIMATE DISPOSAL FIGURE 9.5. Concept For Transmutation By Fusion ( CTR i Reactors 9.10 BNWL-1900 to the federal repository. Since a sustained controlled thermonuclear reaction has not yet been achieved, use of this particular concept for waste management requires a break- through in this technology, in con- trast to the fission reactor concept which requires only technological ref i nement . 9.2 TECHNICAL FEASIBILITY Evaluation of technical feasibil- ity was made for each transmutation concept system. The results of the evaluation are summarized in Table 9.1. For accelerator systems, only the spallation type accelerator showed real promise for feasibility. A spallation accelerator with a neu- tron-producing target could meet the feasibility criteria under certain conditions for the transmutation of actinides to fission product. How- ever, this process would be very costly at best and is judged to be much less attractive than the use of fission reactors or of fusion reac- tors. Certain lonq-lived fission products might possibly be trans- muted with the neutrons from a spall- ation accelerator. This concept would, however, require chemical separation, treat only a small frac- tion of the fission products, and require a large capital investment in a large number of accelerators. Ten- tative calculations indicate the more likely feasibility of transmutation of Category 2 fission products when TABLE 9 . 1 Summary of Transmutation Device Feasibility Technically Feasible for Transmutation Device Accelerators • Electron Accelerator • Proton Accelerator • Spallation Accelerator Thermonuclear Explosives Fission Reactors Fusion Reactors Fission Produc ts Actinides Category' 3 3 i Cateqory( a ) 2 Category( a ) 1 Category (a) 1 No No No No No No No No No Possibly Possibly Possibly No No Possibly( b ) Yes No No No Yes No Possibly Yes Yes a) Category 1 Category 2 Category 3 b) For specific isotopes Storage required for >1 000 years Storage required for 100 - 1000 years Storage required for <1 00 years 9.11 BNWL-1900 they are used directly as the spall- ation target. This possibility ap- pears to be expensive and will re- quire improvements in accelerator technology beyond that which already exists. Underground explosion of thermonuclear devices is estimated to be feasible for transmutation of actinides and marginally so for some separated isotopes of fission product. It is not clear that the concept is feasible from a seismic standpoint since large numbers of 100-kiloton (kt) explosions would be required annually. The trans- mutation of actinides in fission reactors is technically feasible, could reduce the actinide inventory to be managed by larae factors, and would incur only a modest cost pen- alty. Fission product transmutation in thermal fission reactors does not meet the criteria. However, transmu- tation of certain long-lived fission products in fast reactors might be possible if the fission products can be separated. It is technically fea- sible to transmute moderate and long- lived fission products and actinides by the neutrons of fusion reactors if their development is successful. Supporting details on feasibility of each concept system are provided in Appendices 9. A through 9.D of this vol ume . 9.2.1 Accelerator Devices Basically four schemes were eval uated: • Direct bombardment by charged par- ticles from accelerators having ener- gies of tens of MeV • Acceleration of the beta decay pro- cess by Coulomb excitation • Use of two processes of photon transmutation: electron bremsstrah- lung and stimulated gamma emission • Use of a h i ah energy (>_ 1 BeV) pro- ton accelerator to produce by spalla- tion an intense source of neutrons for transmuting radionuclides. The last of these, namely the BeV proton-induced spallation device is the only concept showina any pro- mise. The others are ruled out on a technical basis since they do not meet the criterion of enerqy balance. Each of these is discussed briefly below. Additional details are pro- vided in Appendix 9. A of this volume. 9.2.1.1 Charged Particle - Nu - clear Reactions The penetration of medium-mass nu- clei by protons requires that the protons have energies of tens of MeV. Nuclei with higher atomic numbers re- quire even more energy. To achieve one nucl i de-el i mi nati ng reaction with a charged particle would require an energy input of ^5000 MeV (i.e., 5000 MeV/reacti on ) . The energy re- leased in the power reactor where the waste was created is ^200 MeV/fission or about 1000 MeV per FP of radioac- tive waste. Clearly more energy is expended in transmuting the waste than was acquired in creating it. For waste transmutation by beams of charged particles from accelerators to become more attractive requires increases in both the probability of a reaction and the number of nuclei transmuted per reaction. 9.12 BNWL-1900 9.2.1.2 Beta Decay Acceleration By Coulomb Excitation Beta decay from certain excited states of some nuclides proceeds more rapidly than beta decay from the ground state or, in some instances, more rapidly from the ground state than from certain metastable states. If this is the case for a radioactive fission product, it may be possible to reduce the effective lifetime of a nuclide by inducing a transition to a more rapidly decayina state. The fission product isotope Kr-85 is representative of these circumstances and therefore was studied. Based upon an assumed energy bal- ance, which reflects an optimistic amount of energy spent on transmuting Kr-85 nuclei, the reaction cross sec- tion needed to accomplish this was derived as a function of the energy of the particle (proton) causing the excitation. Comparison of the reac- tion cross section required for trans^ mutation to occur with estimates of the physical values, shows differ- ences of about three orders of maqni- tude (10 ), the physical values being smaller. Thus, the energy balance criterion is not met and this scheme is not technically feasible. 9.2.1.3 Photon Transmutation Processes Two different photon processes were considered for radionuclide transmutation. These were photodis- integration of nuclei by electron bremsstrahl ung , and stimulated gamma emi ssi on . Electron Bremsstrah 1 ung . Bombard- ing targets with electron beams in the tens to low hundreds of MeV range produces a "shower" of photons from bremsstrah 1 ung , from annihilation of electron-positron pairs, and from other processes. Some of the photons in the "giant" resonance energy re- gion will undergo nuclear interac- tions such as ( y » n ) , (y,p), ( y » 2 n ) or ( Y > n p ) . With nuclear waste material as targets, these nuclear reactions might lead to stable or less toxic nuclei. The feasibility criteria for such a system are the reaction yield, which consists of the number of reactions per incident electron, and the energy supplied per reaction. Assuming electrons accelerated to 34 MeV, the yield of photons was found to be too small and the energy required per transmutation reaction was found to be roughly two orders of magnitude larger than that gained (i.e., 200 MeV/fission in the reactor) Thus the energy balance criterion was not met, and this scheme is concluded to be not technically feasible. Stimulated Gamma Emission . The stimulated emission of light was dis- covered in 1960. This process is known as the Laser (L_ight Amplifica- tion by Stimulated Emission of Radia- tion). In this process excited en- ergy states of atoms or molecules decay very rapidly when placed in an intense field of electromagnetic ra- diation of the proper wavelength. One mode of the decay of excited en- ergy states of nuclei is by gamma ray emission. Since gamma rays are elec- tromagnetic radiation, as is light, the possibility of stimulated gamma ray emission, a Graser, has been dis- cussed since the discovery of the 9.13 BNWL-1 900 Laser. In the Graser concept, a ma- trix material contains the waste isotope(s) and other radioactive ma- terial of higher intensity but shorter half-life. Hence the proba- bility exists that emission by a given nucleus increases with increas- ing intensity and exceeds the natural decay rate of the waste isotope. The two major problems which tend to dis- courage hopes for this concept are: • According to the best available theory, stimulated emission i-s possi- ble only for neutron boson emission. For practical purposes, this means photon emission and no stimulated emission is possible for alpha or beta decay of nuclei. • The low attenuation material me- dium needed for stimulated gamma decay may be unattainable. The concept of stimulated gamma emission as a transmutation method depends on the existence of nuclear states which could be reached by gamma excitation and which decay to a stable or short-lived isotope more rapidly than natural decay. of the fission product. No such states have been identified. In addition, if such states were found, then stimulated emission requires reduc- ing the attenuation of gammas by several orders of magnitude in order for it to be a technically feasible scheme . 9.2.1.4 Spallation Accelerators High energy (> 1 BeV = 1,000 MeV), large current proton accelerators f 2- 5 ) have been proposed v ' which would provide the most intense, continu- ously operating source of neutrons yet attained. The Intense Neutron Generator (ING) which was studied by the Atomic Energy of Canada, Ltd., ( 2 ) . . group v ' was the most serious investi- gation of a large spallation accel- erator. The high energy neutrons pro- duced through the spallation process in a Pb-Bi target were to be moder- ated in a surrounding DpO medium to provide a thermal neutron flux of at 16 2 least 10 n/cm -sec. Thermal flux values of this magnitude are required for the transmutation of low neutron cross section fission product (FP) isotopes. Several groups^ " ' have studied the possible use of spalla- tion accelerators for the transmu- tation of fission products over the years. Studies have also been made of the use of spallation accelerators to transmute fertile to fissile mate- rial, a concept which is closely re- lated in technical feasibility to the FP transmutation concept. The energy invested in transmuting one fission product nucleus was esti- mated to be between 23 and 110 MeV. Since these values are less than the 200 MeV acquired in creating the fis- sion products, the energy balance criterion may be met. Candidate fis- sion products for transmutation in a spallation accelerator are Cs-137, Sr-90, and Tc-99. The energy needed to transmute these isotopes is less than the energy invested in their creation . The inventory of fission products such as Sr-90, Tc-99 and Cs-137 can be reduced significantly only by very high neutron flux levels, on the order of 10 16 to 10 17 n/cm 2 -sec. On the basis of a proton beam power of 9.14 BNWL-1900 65 MW (the ING proposal), ^ it is estimated that roughly two spallation accelerators are needed to handle the inventory of one LWR. The rates of transmutation of Sr-90, Tc-99, and Cs-137 were calcu- lated and compared to the natural decay rates. The results are shown in Table 9.2. As shown, the rate is highly dependent on the thermal flux level attained with the device. It is very effective in accelerating the decay of Tc-99 because of the magni- tude of the Tc-99 cross section (^22.6 barns). Clearly, thermal flux levels greater than 10 n/cm -sec are needed to make the concept attrac- tive for transmutation of Sr-90 and Cs-137. It is assumed that daughter products may have to be removed from the system frequently to keep them from competing for the neutrons. Ma- terials problems in the transmutation target, including its clad, and other parts of the system can be antici- pated to be severe. In examining the waste balance, it is found that a spallation accelera- tor may create more waste than it can transmute. The spallation protons with energy of 1 BeV on the lead tar- get creates more short-lived radioac- tive product nuclei. The approxi- mately 20 neutrons produced per incident BeV proton will come from on the order of 5 to 10 parent nuclei, with almost all of the daughters being left in radioactive precursor states. Almost all of the radioac- tive waste created, however, has half- lives much shorter than the fission products which would be transmuted. This would not be true for a uranium target . Another conceptual use of the spallation accelerator is to use the radioactive fission product waste as the proton target of the accelerator. In this concept the fission product nuclei are transmuted directly by spallation and further by nuclear processes induced by the secondary TABLE 9.2. Transmutation Rates using Thermal Neutrons from a Spallation Accelerator Time (yrs) To Eliminate 99% of the Material Isotope Transmuted Natural Decay Transmutation Using Accelerator 10 15 n/cm 2 - -sec io 16 n/cm 2 -sec Sr-90 191 83 14 Tc-99 1.40xl0 6 6.6 0.66 Cs-137 199 170 80 9.15 BNWL-1 900 neutrons. A study team of the Japa- nese Atomic Industrial Forum has speculated that eighty-five Cs-137 nuclei could be transmuted per inci- (5) dent proton. The Los Alamos Scientific Laboratory group which has reviewed this study suggests that 200 FP nuclei might be transmuted per incident proton (Appendix 9.E of this vol ume ) . 9.2.2 Fission and Thermonuclear Explosive Devices Because thermonuclear explosive de- vices are known to produce large yields of neutrons, the use of these neutrons for the transmutation of ra- dioactive waste has been pro- ( fi -ft \ posed. ' A preliminary evaluation by PNL indicated that this concept was not technically feasible. The evaluation was by Los Alamos Scien- tific Laboratory (LASL) personnel who strongly disagreed with the PNL Study findings and concluded that the con- cept may be very attractive. The de- tails of the LASL review are con- tained in Appendix 9.E of this volume Subsequently, PNL reevaluated the con- cept based on the LASL review. The details of the PNL evaluation are given in Appendix B of this section. The concept which was evaluated consisted of the explosion in a hole (1.5km deep) of a fission-actuated thermonuclear device with an explo- sive yield of 100 kilotons. The ex- plosive yield was taken from the LASL review as the largest which could be seismically decoupled from adjacent explosions since a large number would be required on an annual basis. The main conclusions of the PNL evaluation of the feasibility of the concept are : • The concept can be considered only for radioactive waste with half-lives much greater than the 12.3 year trit- ium because of significant production of tritium residue in the device. Hence, the transmutation of short- lived radioactive waste is not tech- nically feasible. • The estimated cost of transmuta- tion of the long-lived fission pro- duct isotopes, TC-99 plus 1-129 plus CS-135 in elemental form, is almost twice the cost of the electricity produced in their creation. Hence, the transmutation of long-lived fis- sion product elements or structural material waste is not technically f easi bl e . • The technical feasibility of the transmutation of separated isotopes of long-lived fission products cannot be obtained from the PNL evaluation but would require more rigorous cal- culation of the neutronic behavior of the concept. • The estimated cost of transmuta- tion of the actinide waste of Np, Am and Cm is estimated to be less than 20 percent of the electrical cost of their creation. The concept is, thus, technically feasible for actinide transmutation. This concept is, how- ever, judged to be much less attrac- tive than the concept of actinide re- cycle in fission reactors both on a cost basis and also on the large annual number of thermonuclear explo- sives required for the transmutation. 9.16 BNWL-1900 9.2.3 Fission Reactors Since fission reactors produce neutrons, it has been sug- [ 3 9 10) gested v ' * ' that radioactive waste might be transmuted in fission (3 9) reactors. Early studies v ' ' pro- posed burning fission products in high-flux reactors. More recent stud ies^ ' conducted at Oak Ridge Na- tional Laboratory (ORNL) have consid- ered transmuting fission products and actinides in fission reactors. Clai- borne 1 ' did an extensive investi- gation of transmutation of actinides in a prototypical commercial pressur- ized water power reactor (PWR) and showed that the toxicity could be re- duced by one to two orders of magni- tude by recycling actinides in PWR U0 2 fuel. The ORNL studies concluded that transmutation of fission pro- ducts in fission reactors was not technically feasible, but that trans- mutation of actinides was feasible. Independent analyses made at PNL con- firm both of these conclusions. Kubo^ 11 ) and Kubo and Rose^ 12 ^ at Massachusetts Institute of Technology have studied Claiborne's results and similarly have concluded that acti- nide recycle in fission reactors is not only technically feasible but an attractive waste management concept. Kubo and Rose further conclude that actinide recycle in fast breeder re- actors would be even more attractive. This conclusion, however, is largely based on intuition and is, as yet, unsupported by any definitive cal- culations. Their extension to Clai- borne's work has been primarily de- voted to analysis of the further reduction in potential toxicity which would accrue from improved separations efficiencies for acti- nides higher than plutonium. The technical feasibility for transmuting fission products and acti- nides is summarized below. Support- ing details are given in Appendix 9.C of this vol urn e. 9.2.3.1 Fission Product Transmu - tation Transmutation of fission products in fission reactors is not techni- cally feasible because the waste in- ventory and transmutation rate crite- ria cannot be met. The cross sections for the fission products and the thermal neutron flux levels attained in fission reactors are both too low. Indeed, if the cross sec- tions were appreciable, the fission products would not exist in the quan- tities they do since they would burn out in-situ. Table 9.3 was extracted from Reference 10 to illustrate why fission product transmutation is not technically feasible. As shown in the table, the neutron cross sections are small, thus requiring a high neu- tron flux. The data given in the bottom of Table 9.3 show that flux levels of the order of 10 to 1 7 2 10 ' n/cm -sec are required to achieve a reasonable gain in trans- mutation rate relative to natural de- cay. In summary, it is concluded that transmutation of fission pro- ducts in thermal fission reactors is not feasible since none of the tech- nical feasibility requirements appear to be met. 9. 17 BNWL-1900 TABLE 9.3. Nuclide Properties of Several Important Fission Product Nuclides and Time Required for 99.9 Percent Reduction of Their Inventory by Decay and Neutron Transmutati on ( 1 ) (a) (b) Half-life, years Burnout cross section, barns Curies/metric ton in spent fuel (c) Relative hazard in spent fuel ■> m air at RCG/metric ton 3 m water at RCG/metric ton Time required for 99.9% >,\ decay and burnout, years Decay only * = 10 14 n/cm 2 -sec^ $ = 10 15 n/cm 2 -sec^ * = 10 16 n/cm 2 -sec^ * = 10 17 n/cm 2 -sec^ Sr-90 28.9 1.2 77,600 2.6 x 10 2.6 x 10 Cs-137 30.2 0.17 108,000 2.1 x 10 5.4 x 10- 14 Kr-85 10.74 1.8 11,400 3.8 x 10 H-3 10 12.33 nil 708 3.5 2.3 10' 1-129 1.6 x 10' 35 0.0367 1.8 x 10 S 6.1 x 10 5 288 302 107 123 1.6 x 10' 249 295 106 123 63 112 245 98 123 6.3 17 91 57 123 0.63 1.8 12 11 123 0.06 a. Effective thermal cross section in typical spectrum of a PWR having average thermal flux of 2.91 x 10 13 n/cm 2 /sec. b. Per metric ton of uranium charged to a PWR having average specific power of 30 MW/metric ton and burnup of 33,000 MWd/metric ton. c. Volume of air and water potentially contaminated to RCG (10 CFR 20) by the content of a metric ton of spent fuel . 6 □ d. Indicated times are doubled and tripled for reduction of inventory by factors of 10 and 10 , respectively. e. Average thermal flux assuming spectrum typical of that in a PWR. 9.2.3.2 Actinide Transmutation Claiborne^ ' has conducted ' the most extensive study to date of actinide transmutation in LWRs. The result of Claiborne's study are used exten- sively here to delineate technical feasibility of actinide transmutation in fission reactors. Some prelimin- ary calculations were made at PNL of actinide recycle using Claiborne's strategy and examining some'other strategies. These three strategies are depicted in Figure 9.6. The re- sults of the PNL calculations confirm that actinide recycle in thermal fis- sion reactors is technically feasible, since it meets all of the selection criteria. Review of Claiborne's Work . Claiborne's strategy, shown as strat- egy 1 in Figure 9.6, was to store the U and Pu recovered during chemical processing and recycle the other ac- tinides in new PWR U0 ? fuel assem- blies. Thus each fuel rod in every PWR fuel assembly contained these ac- tinides. His analysis covered sev- eral chemical processing extraction efficiencies. He showed that the actinide inventory in accumulated 9.18 BNWL-1900 1. Reference Strategy (Proposed by Claiborne Dol Spent Fuel £ Reprocess T Actinides Other than U & Pu Hold Recovered U & Pu 2. Alternative Uranium Strategy Fuel Fabrication — r~ Fresh Enriched UO. New Fuel PWR Same as 1 except recycle actinides in 10% of new fuel rods, 90% of new fuel normal U0 ? fuel 3. Plutonium Recycle Strategy Spent Fuel 1 Actinides In- cluding Pu New Fuel 1 Reprocess Fuel in Fabrication PWR I Natural UO J J Hold Recovered U ..««.«•■ -■ ^ FIGURE 9.6. Actinide Recycle Strategies Studied New Enriched UO- Fuel waste is substantially reduced when actinides are recycled. His results are cast in terms of potential relative hazard, ' defined as the amount (volume) of water (or air) required to dilute each nuclide of a mixture to its Radiation Con- centration Guide value (RCG). The cumulative hazards of actinide waste for the short and long terms are dis- played in Figures 9.7 and 9.8, respec- tively. The top curve on these fig- ures is the potential hazard of the actinide waste accumulated if the ac- tinides from a 1000 MWe PWR, operat- ing at 80 percent capacity for 60 years, are not recycled in the reac- tor (i.e., other disposal schemes are used 10 years after reprocessing). The bottom curve is for the recovery of 99.5 percent of all actinides dur- ing chemical processing, with the U and Pu stored, and all other actin- ides recycled. The difference be- tween the two curves represents the reduction of the potential relative hazard of actinide waste. Claiborne also shows that the reduction is even more substantial (factors of 200) if recovery of 99.9 percent of U and Pu can be achieved in chemical process- ing. The postulated reductions in We prefer to use the term toxicity index rather than hazard index because its connotation is clearer. However, since we are taking information directly from Claiborne's report, (10; we use his notation here. 9.19 BNWL-1900 1 - y^C - / NO RECYCLE - 9». 5 / OF 11 i Pu REMOVED - 1 - A / RECYCLE OF 99.5' OF ALL ACTINIDES / EXCTPT U AND Pu 1 1 1 1 I 1 TIME AFTER FUEL IS FIRST DISCHARGED FROM THE REACTOR, YR FIGURE 9.7 . Short-Term Cumulative Hazard of ActinideWaste from 60-Year Operation of a Typical PWR (10) cumulative hazard may not be meaning- ful on an absolute basis. What they do reflect is the reduction in quan- tity of actinide waste which must be managed or otherwise disposed of. There is an implicit assumption that if this quantity is reduced, the future hazard is reduced proportionately. PNL Studies . Survey calculations were made at PNL using the ALTHAEA ( 1 3 ) code v ' for the three strategies shown in Figure 9.6. The first strat- egy which was calculated was that as- sumed by Claiborne in order to have a verification of his results. For the first strategy of recycl- ing in e^/ery U0 ? rod, the calculated amount of each actinide element dur- ing recycle was consistently larger than the values obtained by Claiborne o E J 1 10' c 10 10 io- 10' J ' I I I I ll \ NO RECYCLE - 99.55 OF U + Pu REMOVED RECYCLE OF 99.5% OF ALL ACTINIDES EXCEPT U AND Pu J I I I I I ll J ■ I i ■ ■ i I J I ' I i i i i io J i(T TIME AFTER DISCHARGE, YR 10- 10' FIGURE 9.8 . Long-Term Cumulative Hazard of Actinide Waste from 60-Year Operation of a Typ- ical PWR (10) 9.20 BNWL-1900 Of the numerous possibilities exist- ing to cause differences in calcu- lated values, the most logical being cross section differences and that the calculations made by Claiborne may not have included neutron spatial self shielding effects, since he used the ORIGEN code^ 14 ^ which does not have this provision. Comparison of neutron multiplication values shows that the values obtained by Claiborne are consistently lower but with the same trends as those obtained in the PNL study. This tends to support the contention of differences due to spa- tial self shielding effects. The differences in concentrations indi- cate that the hazard reductions in transmutation of actinides presented above from Claiborne's study might be overestimated. A penalty of about 4 percent in the enrichment of U-235 is encountered for equilibrium cycles when actinides are recycled in every U0 2 fuel rod. A second set of calculations was made assuming that the actinides would be recycled in every tenth U0„ rod rather than in every rod. The actinide element inventories for this case along with those of recycling in every UO2 rod are presented in Table 9.4. Comparison of the differences 1n values between cycles for each strategy shows that the actinide in- ventory is not reduced as much by re- cycling the actinides when concen- trated in a few rods as compared to recycling in every U0 ? rod. The en- richment penalty for this strategy ranges from O percent in U-235 con- tent at second recycle to -v5 percent at equilibrium. The incentive for re- cycling actinides in fewer rods is to reduce the economic penalty which would occur if remote fabrication of UO2 fuel containing actinides is re- quired. In this event the additional enrichment penalty would be a small fraction of the savings derived from reduced fabrication costs. The third strategy calculated was the storage of the recovered U from chemical processing and the recycle of the rest of the actinides in LWR Plutonium recycle fuel (UO^-PuOj). The plutonium recycle scheme employed was typical of current fuel manage- ment schemes for recycling self- generation plutonium. The inventory of the actinides Np, Am, and Cm fall in between the values obtained for the UO2 fuel strategies, thus making it a technically feasible concept. However, the plutonium generated con- tains large quantities of Pu-238 which would dictate expensive fabri- cation in a remote facility. The fis- sile content also becomes so low that the fuel would eventually have to be driven by other fuel in the reactor. In summary, actinide transmutation in fission reactors appears to be technically feasible and is attrac- tive in that significant reductions in the toxicity index of actinides can be achieved. The cycle could be accomplished in L)0 ? rods, either ev- ery rod or some fraction thereof, or in plutonium recycle rods. This study indicates that using existing chemical separation and recovery effi- ciencies, the recycle of actinides in light water power reactors can achieve a decrease in the short-term toxicity index of about a factor of 10 and 9.21 BNWL-1 900 TABLE 9.4 . Comparison of Actinide Inventories for Two Recycle Strategies Using U0 ? Fuel in A PWR. Actinide Inventory (gms/MT of Heavy Metal in Reactor) Am CUT Recycle No. 1 2 3 4 5 6 7 8 9 Strategy ,1a) !« 521 521 784 812 921 994 993 1118 1031 1193 1052 1253 1063 1300 1068 1334 1071 1360 1073 1380 1 2 1 2 144 144 32.0 32.0 176 176 92.0 92.5 183 187 135 141 184 192 161 177 184 194 176 202 184 195 184 222 184 195 189 233 184 196 191 239 184 196 192 242 184 196 193 243 a. Recycle actinides in every U0 2 rod as shown in Figure 9.7. b. Recycle actinides in one-tenth of the UOp rods. 11 about a factor of 50 decrease in long- term hazard index. These reduction factors may be significantly improved by achieving higher separation effi- ciencies and better optimization of the reactor irradiations. As Kubo points out, a greater reduction in the toxicity indices may also be possible by recycling actinides i\n FBRs. There are, however, no calculations presently available to support the latter specu- lation. Lastly, as discussed in peer group review meetings (see Appendix 9.E of this volume), a special pur- pose reactor designed specifically to destroy actinides (i.e., convert them to fission products) has merit and should be evaluated. Such a facility might be constructed and operated by the Federal government. 9.2.4 Fusion Reactors The unique features of a fusion re- actor, or Controlled Thermonuclear Reactor (CTR), as a waste transmuta- tion device are the high energy of the neutrons available and the high *i * • • j. j (15,20) T . flux anticipated. ' The most troublesome fission product nuclei to be considered for transmutation are those which have relatively small neu- tron reaction cross sections. Due to 9.22 BNWL-1900 the high energy, (n,2n) reactions can contribute significantly to the pro- cess, while the high flux and high source strength makes waste transmu- tation at reasonable rates appear pos- sible a priori. Studies were made of waste trans- mutation in blanket regions of a CTR. Included were: 1) the use of fast neutron flux for transmutation of Sr-90 and Cs-137, 2) the use of ther- mal neutron flux for transmutation of the fission products Sr-90, Cs-137, Kr-85, 1-129, and actinides, and 3) transmutation of large amounts of Cs-137 in a moderating blanket. A summary of the principal find- ings of these studies is given below. Supporting technical details are given in Appendix 9.D of this volume. CTR power plants can, in principle, transmute all of the Sr-90 and Cs-137 created by the electrical economy, even if the economy assumed were all nuclear power. The daughter products produced in the transmutation process appear to approach well defined equi- libriums after a year or so of opera- tion. They should, therefore, cause few perturbations on the characteris- tics of operating CTR power plants. In addition, the fission product nu- clei can contribute to additional neu- tron production in the CTR blanket. Both fission products and acti- nides can be effectively transmuted with thermal neutrons using dilute target samples. For a representative value of a fusion plasma power den- sity, calculations indicate that the toxicity ha 1 f -1 i f e^ a ' is reduced by about 5 orders of magnitude for the actinides and by 8 orders of magni- tude for 1-129. For the other fis- sion products the toxicity half-life is reduced roughly in the range be- tween 5 and 100. Quantities of Cs-137 could be transmuted under the projected CTR (19 20 ) blanket loading conditions^ ' . The reductions in Cs-137 toxicity are, however, projected to be at most a factor of about three. The irradiation of actinides in the blanket of a CTR would greatly re- duce the cumulative toxicity index due both to the high energy of the D-T fusion neutrons and to the in- tense neutron sources expected for fu- sion reactors. Calculations made for irradiations of small samples of acti- nides in CTR blankets indicated reduc- tions of toxicity indices by factors of 1,000 to 10,000 for expected neu- tron source levels. Definitive calcu- lations have not yet been made for more massive loadings of actinides. but this concept is clearly capable of reducing the toxicities by a fac- tor of at least 10 to 100 below that of actinide recycle in LWRs. In summary, the technical feasibil- ity of transmutation of actinides and selected fission products seems clear- ly established. The limitations which will be imposed by factors such as radiation damage and heat transfer Defined as the time required for the toxicity index for a particular iso- tope and its daughters to decay to one-half of its original value. 9.23 BNWL-1900 will require definition when the at- tainment of a viable CTR appears to be closer than it is at present. 9.2.5 Candidate Transmutation Con - cept System Since it is technically feasible to transmute actinides in fission re- actors and CTRs, and certain fission products in CTRs, these two reactor technol ogi es ' combi ne to form a poten- tially viable long-term strategy for waste management. This strategy is described below. 9.2.5.1 System Description The layout of the system is given in Figure 9.9. In the near term (year 2000 or 2010), the actinides obtained from the partitioning pro- cess would be recycled in fission re- actors and the fission products placed in a retrievable storage fa- cility. In the long term, if the development of present concepts of CTRs is successful, the fission prod- ucts would be retrieved from storage and recycled along with the actinides in the CTR. A pictorial representation of the projected nuclear power economy is given in Figure 9.10 to illustrate the impact of this strategy. The in- stalled capacity of fission reactors up to year 2000 is taken from the AEC projection given in Reference 21. The actinide inventory projected for this growth rate (shown in Volume 1 of this report) is also shown on Fig- ure 9.10. It is not unlikely that fission reactors will follow the bell shaped curve, since new technologies will be developed which will compete ENERGY ■*- URANIUM FUEL FABRICATION PLUTONIUM STORAGE ^ DEPLETED URANIUM CONTROLLED THERMONUCLEAR REACTOR (CTR) FRESH FUEL HEAVY ELEMENTS ctr products " fIssIon" products PLUS ACTINIDES REACTOR SPENT FUEL REPROCESSING. PARTITIONING ENERGY RADIOACTIVE WASTE RETRIEVABLE STORAGE ULTIMATE DISPOSAL P •NEAR TERM FAR TERM F I GU RE 9.9 . Transmutation Waste Management Strategy 9.24 BNWL-1900 . s ^1200 / • '"^ § / / s O / / \ O 1500 (A >• Z o -900 / '' < 0. < o oc ACTINIDE INVENTORY / / y A.FISSION REACTOR CAPACITY* K O UJ WITHOUT RECYCLE W 2 u 1000 UJ > o -600 / / / K z / z UJ / o > (A z / E 500 UJ " Q -300 // / Q z CTR,/ ' CTR CAPACITY UJ -j I- -j u / 2 < / (A / z o ••• * 1 ■ 1 1 ■ ' — ■ — 1970 1980 1990 YEAR 2000 BEYOND F I GURE 9.10 . Projected Nuclear Power Economy in U.S.A. more favorably in the power market- place. If actinide recycle is under- taken in fission reactors, the acti- nide inventory should tend to follow the fission reactor growth curve. The fission reactors being built in the latter period of this curve will have increasingly larger amounts of actinides recycled through them. How- ever, if the CTR becomes a viable power source, then actinide recycle can be transferred to these plants, eventually transmuting all of the ac- tinides into fission products. Then the most toxic of the fission prod- ucts can themselves be converted in the CTR to stable or less toxic species. 9.2.5.2 Requirements for the Con - cept System The general system flow diagram, shown in Figure 9.11, starts with bulk high-level liquid waste from the reprocessing plant. The waste may be chemically conditioned and stored for a 5-year period to provide for cool- ing before being chemically segre- gated into actinide and fission product streams. The actinides, long-lived fission products, and short-lived fission products are each solidified and transported to the fab- rication site. The transportation is expected to be done in conventional types of shielded and cooled casks via common carrier (truck or rail). 9.25 BNWL-1900 FUEL REPROCESSING WASTE MANAGED BY OTHER OPTIONS FRACTIONATE FISSION PRODUCTS F DESIGNATES THE MOST LIKELY ANTICIPATED PATH FOR TRANSMUTATION OF ACTINIDES IN FISSION REACTORS CTR DESIGNATES THE MOST LIKELY ANTICIPATED PATH FOR TRANSMUTATION OF ACTINIDES AND LONG LIVED FISSION PRODUCTS IN A CONTROLLED THERMONUCLEAR IFUSIONI REACTOR TRANSPORTATION TO REPROCESSING SITE TRANSPORTATION TO FABRICATION SITE TRANSPORTATION TO REACTOR SITE EMPLACEMENT OPERATION REACTOR OPERATION SOLIDIFY WASTE OPTION FOR ONE OR MORE WASTE FRACTIONS TO BE MANAGED BY OTHER CONCEPTS ILG. SHORT-LIVED FISSION PRODUCT IN RETRIEVABLE STORAGEI F IGURE 9.11 . Overall System Requirements for Managing High-Level Radioactive Wastes by Trans- mutation Concept Since transmutation of most of the short-lived fission products does not appear feasible, the option to use other waste management concepts such as retrievable storage is expected to be exercised. The actinides are fabricated, ei- ther into fuel or target rods for fis- sion reactors and/or fusion (CTR) re- actors. The fission products would be fabricated into CTR target rods in shielded process lines by remote fabrication. The fabricated rods would be shipped to the reactor site using con- ventional shielded shipping casks. Upon receipt at the reactor site, they are inserted into the reactor. The insertion into the reactor will be done with machines currently in use for refueling reactors. The actinides and fission products in the fuel and/or target rods are transmuted during reactor operation. Any residue, or heel, would be re- turned to reprocessing and the cycle repeated . 9.3 ESTIMATED RESEARCH AND DEVELOP - MENT REQUIREMENTS ^ Research and Development needs are primarily directed toward determining Members of the peer group and others who reviewed the initial draft of this report had numerous comments and suggestions on this section of the report. As a result, portions of this section presented in the draft have subsequently been revised to incorporate these suggestions. 9.26 BNWL-1900 if viable strategies for application of transmutati o-n exist. This is the objective of the Research and Develop- ment study described below. Parti- tioning is an integral part of any transmutation cycle. Thus, the via- bility of partitioning must be estab- lished early in any study to develop transmutation schemes. The near-term potential of actinide transmutation using fission reactors dictates that Research and Development emphasis be placed in this area. However, ad- vanced technologies, such as CTRs, hold future promise, and a sustained low-level Research and Development effort is defined to evaluate the ap- plication of these technologies for waste transmutation. The Research and Development needs are estimated to require 20 years and to cost about $133 million. The needs are outlined in two tasks. Task 1 includes the Research and De- velopment needs for developing the fission reactor transmutation concept and is estimated to require between 10 and 15 years and cost $130 million to complete. Task 2 includes the Research and Development needed to monitor advanced strategies and make preliminary evaluations of their ap- plicability as transmutation devices. This task is estimated to cover 20 years and cost $3 million. A fundamental assumption under- lying the estimates presented here is that transmutation will be accom- plished in commercially owned facili- ties. It should be noted that the possibility of transmutation in gov- ernment-owned special purpose acti- nide burning reactors merits consid- eration. In fact, it was the concensus of the peer review group that this should be studied (Appen- dix 9.E of this volume). However, estimates of the needs and costs re- quired for researching and developing this waste management strategy were not developed for this study. 9.3.1 Task 1 - Actinide Recycle in Fission Reactors The Research and Development study discussed below is directed toward the development of a viable strategy for application of transmutation in fission reactors. Research and De- velopment expenditures for actinide recycle in fission reactors are esti- mated to be: $3 to 5 million for partitioning; $50 million for terres- trial disposal of the residual waste; and $75 million for actinide recycle engineering. The latter investment of $75 million assumes that govern- ment provides $25 million and indus- try funds $50 million. The $ 2 5 - m i 1 - lion investment by government would be to develop base technology; whereas, investment by industry would implement the technology to a viable commercial operation. The needs rela- tive to terrestrial disposal of the waste fraction not transmuted are out- lined in Volume 2 of this report and will not be discussed here. The needs for the balance of the actinide recycle Research and Development study are outlined below. The Research and Development pro- gram for actinide recycle includes two phases. The first phase covers three to four years to evaluate the efficacy of actinide recycle in vari- ous types of central power stations currently operating (LWRs and HTGRs) 9.27 BNWL-1900 and being developed (LMFBRs and GCFBRs). The efficiency of actinide transmutation in each of these reac- tor types would be evaluated, the limits of technology for partitioning actinides would be determined, and the impact (in terms of risk) of im- plementing actinide recycle in these fuel cycles would be estimated. To assess the desirability of actinide transmutat'i on requires evaluation of the effect of recycling actinides on all waste streams in the fuel cycle (e.g., in the process steps of fabri- cation and reprocessing/partitioning) as well as the reduction of high- level waste via burnup in the reactor, If this first phase indicates the need, the program will proceed to the second phase. The second phase in- cludes the Research and Development needed to implement the concept on a commercial scale. Technical bases and criteria needed to' guide design would be developed. Experimental pilot programs, and a full-scale demonstration program would most likely be required to establish the viability of the concept. Process engineering studies associated with fabrication, reprocessing, and par- titioning will be needed to establish the technology and to develop guide- lines and specifications which assure safety during all fuel cycle opera- tions. A cooperative effort between governmental laboratories and private industries is required to meet the objectives of this program. The two phases of the Research and Development program in Task 1 and the activities within these phases are displayed in Figure 9.12 as a func- tion of program years. Each phase and the Research and Development activities within each phase are dis- cussed below. 9.3.1.1 Phase 1 - Evaluate Effi- cacy of Actinide Transmu- tation in Fission Reactors The objective of Phase 1 efforts is to confirm the success of trans- mutation of actinides in fission reac- tors and provide quantitative data for a basis to evaluate if the need is compelling relative to other waste management alternatives. The optimum nuclear fuel cycle for transmutation will be identified. The efforts are broken into three major activities which are shown in Figure 9.12. The first activity is to evaluate actinide transmutation, considering each of the various reactor types cur- rently operational or expected to be commercially viable as central power stations between now and the early 1990s to determine the efficiency of each in reducing actinide inventory of hi gh-1 evel waste . The reduction of toxicity in acti- nide recycle is directly related to the efficiency of partitioning acti- nides from the high-level waste stream. Therefore, Research and Development efforts are needed to determine the level of partitioning that is technically achievable, and these efforts are listed as the second major activity in this phase of the program. The third activity is to evaluate the impact of recycling actinides in 9.28 BNWL-1900 Years Phase 1- Evaluate Efficacy Activity 1 - Efficiency Activity 2 - Partitioning Activity 3 - Risk Phase 2 -Development and Demonstration Activity 1 - Technical Bases Activity 2 - Design Criteria and Process Specifications Activity 3 - Small Sample Irradia- tions Activity 4 - Full Scale Demonstra- tion 10 12 14 16 18 20 1 2 3 4 5 6 7 Milestones: 1. Identify optimum type of reactor for transmutation. 2. Evaluate total risk of implementation. 3. Phase 1 completed. Go or no go decision on transmutation relative to other waste management alternatives. 4. Initial decision on demonstration experiment based upon early data from capsule irradiations. 5. Final Decision on proceeding with demonstration experiment. 6. Demonstration completed. 7. Phase 2 completed. FI GURE 9.12 . Estimated Research and Development Program for Actinide Transmutation in Fission Reactors these reactors by assessing the incre- mental risks at the affected process stages in these nuclear fuel cycles. This information is needed to assure that advantages accruing to waste management objectives via destruction of actinides in the reactor are not negated by increases of actinides in processing waste streams such as in fabrication and partitioning. The data developed in these three activities will be used to determine the viability of transmutation for each of these nuclear fuel cycles in order to identify the most likely fis- sion transmutation cycle. This infor- mation can then be used in comparison with similar data on alternative disposal methods (e.g., geologic, seabed, extraterrestrial) to decide 9.29 BNWL-1900 whether transmutation is compelling in management of high-level waste. It is estimated that Phase 1 of the program would be completed in about three to four years at a cost of around $5 to $7 million. Perhaps about one-half of this investment will be needed for Task 2 on parti- tioning since fission reactor trans- mutation requires a partitioning step. Details of Research and Development for partitioning are given in Sec- tion 7 of this volume. Industry is expected to be actively involved in this evaluation, and industries' in- volvement in this phase aids in mak- ing the decision as to whether trans- mutation is viable and compelling in management of high-level waste. 9.3.1.2 Phase 2 - Development and Demonstration of Actinide Re eye 1 e in Fission Reactors Phase 2 of the program would be initiated only if the results of Phase 1 prove positive and it is de- cided there is need for transmutation in management of high-level waste. The objective of this phase, of the program is to demonstrate the via- bility of the concept of recycling actinides in the nuclear fuel cycle(s identified as most likely in Phase 1. The time required for accomplishing this objective is estimated at some- where between 10 and 15 years and the cost about $75 million. The efforts in this phase represent a large-scale cooperative program involving govern- ment and industry. It assumes that government will fund the development of base technology and the transfer of the technology to industrial capa- bility and that industry will fund the development and application of the engineering capability for recy- cling actinides in fission reactors. As shown in Figure 9.12, Phase 2 is outlined as four major activities. Technical bases are needed to in- sure that actinides can be adequately handled in all transmutation cycle process steps--f abri cati on , irradia- tion, reprocessing, partitioning, waste treatment, and transportation between these steps. Thus, the first major activity is to develop the broad technical base needed to imple- ment actinide recycle in fission reac- tors. Included in this activity is evaluation and development of data on the chemical and physical constants for actinides. In addition, experi- mental data will be developed to pro- vide bases for making accurate engi- neering predictions of mechanical, chemical, thermal, and nuclear behav- ior during all process operations in the transmutation fuel cycle and to assure that the basic theoretical knowledge is not limiting the develop- ment of actinide recycle engineering. The second major activity in Phase 2 of the program is the develop- ment of criteria and specifications to design, manufacture, transport, reprocess and partition the actinide- bearing fuel/or target elements to be irradiated in the power reactor. Process engineering studies are required to establish these technolo- gies as viable commercial processes and develop specifications which as- sure adequate safety margins in all steps of the nuclear fuel cycle. 9.30 BNWL-1900 In activity 3, shown in Fig- ure 9.12, irradiation experiments will be conducted to measure the be- havior of small samples bearing acti- nides in reactor environments. These data will permit early determination of the transmutation rates of acti- nides and provide data for assessment and normalization of design methods. Moreover, this information can be used in the first activity to aid in determining the adequacy of know- ledge concerning physical constants used in design analysis. The outcome of early measurements on irradiation of small samples bear- ing actinides will assist in deciding whether to proceed with the design of a large scale demonstration, listed as the last major activity of Phase 2. In the event that the outcome is fa- vorable, then activity 4 will proceed to a full scale demonstration of the concept. This demonstration could be conducted in an AEC-owned facility or, if available, a utility-owned and operated reactor. The program outlined in Phases 1 and 2 involves major expenditures of funds covering roughly 15 years of development. These efforts represent contributions which would be forth- coming from numerous governmental laboratories, universities, private companies, and public agencies. It is important that these efforts be coordinated to assure proper guidance and direction of the research program to meeting waste management objec- tives. Though program management has not been delineated as an activity, it should be recognized that it is needed. 9.3.2 Task 2 - Advanced Concept Eva! uati ons Since transmutation using CTRs shows promise, as does possibly spal- lation accelerators and nuclear ex- plosives, an effort, albeit low rela- tive to tasks of other phases in the program, should be sustained to evalu- ate these concepts. For example, it would be worthwhile to undertake a modest effort to better ascertain the accelerator requirements for transmu- tation in order to determine the break' through necessary in acceleration technology for this alternative to represent a technically feasible transmutation concept. In addition, other advanced schemes are expected to be brought to the attention of the waste management program in the future for evaluation of their feasibility. For example, the development of laser technology may lead to processes ap- plicable for waste transmutation. It is estimated that the evalua- tion of advanced concepts would each cost up to $3 million expended over about 20 years. If significant break- throughs are realized in these ad- vanced technologies and preliminary evaluation of transmutation shows merit, these efforts are expected to expand accordingly. 9.4 ESTIMATED TIME FOR REQUIREMENTS FOR OPERATION It is estimated that between ten and fifteen years would be required for implementing operation of acti- nide recycle in fission reactors. If transmutation of actinides were under- taken in commercial LWRs, then opera- tion could be implemented in about 9.31 BNWL-1900 ten years. Implementation of recycle in commercial fast reactors would re- quire longer time periods and depend upon the advent of commercial opera- tion of fast breeder reactors. If it appears advisable to recycle acti- nides in special purpose burner reac- tors, then sufficient time would be needed to design and construct such a facility. A time span of fifteen to twenty years is estimated to be re- quired for this case. Estimates of the time span needed in transmutation of actinides and fis- sion products in CTRs depends on achieving a viable CTR. Current esti- mates of a commercially viable CTR is in the period between the years 1995 and 2000. We estimate that it would require at least 5 years beyond the commercial implementation date for op- eration of actinide recycle in CTRs. Coupling the technological devel- opments required for actinide trans- mutation in fission reactors and the projected time scale of the introduc- tion of LWRs, HTGRs and LMFBRs in the production of electrical power suggests that the early 1990s repre- sent a reasonable target for appli- cation of actinide recycle in a commercial facility. This would al- low sufficient time for development and demonstration of partitioning. Application of actinide recycle in government-owned facilities could probably be accommodated in the same time frame. Thus, a date somewhere between 1990 and 1995 would seem to represent a conservative estimate ir- respective of the mode of operation. 9.5 CAPITAL AND OPERATING COSTS The transmutation of actinides in a commercial type light water power reactor (LWR) has been shown to be technically feasible. The potential reduction in long-term toxicity index of high-level waste is in the range of two orders of magnitude if acti- nides are recycled in a LWR. In HTGRs and FBRs , equivalent or more re- duction in toxicity index is expected In a CTR, transmutation of actinides and selected fission products (e.g., Sr-90 and Cs-137) appear technically feasible and more rapid reductions in toxicity appear attainable. How- ever, the CTR faces a technological breakthrough in attaining a sustained thermonuclear reaction. For purposes of estimating capital and operating costs, the reference case selected was actinide recycle in LWRs because these reactors are presently competitive with other electrical generation systems. The base case evaluated takes the Am, Cm, Bk, Cf and Np extracted from the high-level waste stream in the par- titioning step and recycles them back through the fabrication and ir- radiation step. This path is par- allel to the plutonium recycle path presently being utilized in LWRs. Wastes from the transmutation pro- cessing would be stored in a re- trievable storage facility until the time that CTRs are developed. These costs are not included since their projection is difficult in terms of if and when CTRs become available. 9.32 BNWL-1900 9.5.1 System Characteristics for Cost Bases Claiborne's analysis^ ' has been used as a base for much of the dis- cussion of actinide reaction in LWRs. A cursory analysis of the costs asso- ciated with Claiborne's recycle strategy shows that it is economi- cally unattractive. Once actinide recycle is initiated, all U0 2 fabri- cation would have to be remote. This remote handling would result in fabri- cation costs approximately five times higher than present U0 ? fabrication costs. Since fabrication costs repre- sent a significant fraction of the total fuel cost, this recycle strategy is far from optimum. However, obvious alternative strategies exist. If the actinides were recycled in a small fraction of the U0 2 rods (<10 percent), then at least 90 percent of the rods could be fabricated without a cost penalty. The remainder would have to be fabri- cated remotely. A fuel management strategy describing this alternative is as follows: 1. Partitioning will recover Th, U, Np, Pu, Am, Cm, Bk and Cf from the high-level waste stream from the re- processing plant. One percent of the fission products will be present in this stream. 2. The actinides will be recyled in 10 percent of the rods. 3. The actinide rods will be re- motely fabricated. 4. The actinide fuel elements in conjunction with the remaining UO^ elements will be designed to give the same reactor endurance (i.e., the same goal exposure) as a U0- fuel assembly. 5. The fuel assemblies remain in the reactor for the same length of time as U0 ? only assemblies. The cost bases for this alternative transmutation strategy are described in detail in the following para- graphs and will be used to obtain a preliminary cost penalty for the transmutation of actinide waste in a LWR. Estimates of radiation levels and associated fabrication costs for (22) plutonium recycle fuels v ' were used as a basis for determining if remote fabrication was a requirement. An estimate of the neutron emission rate to be expected from the acti- nide bearing U0 ? fuel was obtained from Claiborne's work. ' As can be seen from the values on Table 9.5, which is extracted from the refer- enced document, neutron dose rates of 10 n/sec-MT are realized after only a few cycles. By comparison, the "high exposure" plutonium con- sidered in BNWL-273^ 22 ^ has a neutron rate of about 10 n/sec-MT. The in- creased neutron yield of four orders of magnitude would make the manufac- ture of fuel containing the actinides intolerable in a glovebox facility. The transportation of these toxic materials can be minimized by having the actinide target manufacturing facility an integral part of the re- processing plant. In that way the shipping containers, which are nor- mally returned empty from the repro- cessing plant to the reactor, can be utilized for shipment of the acti- nides to the reactor site. Kubo and ( 1 2 ) Rose v ' have come to a similar con- clusion. Thus no significant cost 9.33 BNWL-1900 TABLE 9.5 . Effect of Recycling on Hazardous Radionuclides in the Reactor and the Processing PlantOO) Recycl e Number 1 2 3 4 5 10 15 20 25 30 40 50 60 Air Required for Dilution to RCG, m3/metric ton of fuel Reactor Discharge 1 .20 71 12 23 27 30 36 41 45 2.48 50 52 53 53 10 10 10 10 10 10 10 10 10 10 10 10 10 10 17 17 17 17 17 17 17 17 17 17 17 17 17 17 150 d, (No U Decay or Pu) 1 .41 02 x 30 x 86 x 15 x 31 5.80 6.20 50 74 88 02 7.08 7.09 , 16 , 16 , 16 , 16 I 16 , 16 , 16 , 16 I 16 l 16 I 16 , 16 , 16 ,16 Ne utron Yi n/sec eld Af metri c ter 150 ton of d. Dec fuel ay Acti No U 5.50 1 .53 4.66 1 .47 3.50 6.65 3.33 6.32 8.71 1 .04 1.15 1 .25 1 .30 1 .31 n i d e s or Pu ) 08 (a 0' I 10 I 10 I 10 I 11 , ]1 l 11 , 12 I 12 , 12 > 12 .12 Cf On - y 1 96 x 5 1 73 x o 8 2 77 X o y 1 24 X O 10 3 26 X 10 6 39 X I0 10 3 30 X I0 11 6 29 X O 11 8 66 X I0 11 1 03 X I0 12 1 14 X I0 12 1 25 X I0 12 1 29 X I0 12 1 30 X I0 12 Cm Only 4.76 1 .33 87 20 42 60 24 3.71 4.03 4.23 4.36 4.49 4.54 4.57 0' a. Based on 99.5 percent extraction of actinides. is expected in considering transpor- tation in this cycle. Likewise, existing facilities and machines (e.g., fuel handling) at reactors would be adequate for receiving and charging acti n i de-bear i ng U0~ fuel into and out of the reactor. The neutronic penalty incurred when actinides are recycled in fresh UO2 fuel is discussed in Section 9.2.3 and Appendix 9.C. The neutronic pen- alty associated with the case where the actinides are present in every tenth rod is shown in Figure 9.C.5. The penalty is expressed in terms of the additional U-235 enrichment re- quired to obtain the same reactivity endurance for the fuel rods contain- ing actinides. The incremental en- richment penalty after 10 recycles will be used to estimate the incre- mental cost penalty. Based on the calculations summarized in Fig- ure 9.C.5, the enrichment in all the fuel rods must be increased from the base of 3.3 wt% to 3.47 wt%. 9.5.2 Estimated Costs Recycling actinides in 10 percent of the U0 ? fuel was assumed as the reference case for estimating costs. The estimated annual incremental costs for transmutation of actinides in LWRs are summarized in Table 9.6. The annual fuel costs shown are for the processing and manufacture of acti ni de-beari ng elements. The costs are given for each 1,000 MT of fuel. 9.34 BNWL-1900 TABLE 9.6 . Estimated Annual Incre- mental Fuel Cycle Costs for Transmutation of Acti nicies in LWRs Assuming Remote Fabrication of 10 Percent of the Fuel Cost (millions of dollars/year)/1000 MT of Fuel 10 21 14 45 Component Partitioning Fabrication Enrichment Total The costs are approximately $45 mil- lion per year per 1,000 MT fuel. It was assumed that the partition- ing step, whereby the actinides are separated from the fission products in the reprocessing plant, adds $10/ kg or $10 million per year for the reference throughput. The incremental cost of fabricat- ing fuel elements containing the ac- tinides is very difficult to estimate since little is known about the char- acteristics of the material and there- fore, the fabrication process and plant design. Fuel fabrication costs of $70/kg were estimated for LWR fuel by escalating the pellet costs re- (22) ported by Burnham et al. 'to ob- tain current costs. Escalation rates of 5 percent per year for labor and 7 percent per year for building and equipment were used. These escalation factors are proba- bly conservative since increased throughput can more than offset these costs. For the purpose of this analysis it was assumed that the fuel would be remotely fabricated in a shielded cell facility. The costs for such fabrication are not well known al- though some experimental facilities have been operated. It is estimated that an annual fabrication cost pen- alty would be $140 - $280/kg of fuel fabricated remotely. Since only 10 percent of the fuel would have to be fabricated this way, the penalty would be approximately $21 /kg of to- tal fuel fabricated. The incremental enrichments per recycle and the associated incre- mental costs are shown in Table 9.7. The total cost for base power is es- timated to be around 6.5 mills/kWhe. Using the reference reprocessing plant throughput (5 MT/day) and as- suming the reactors operate at 85 percent capacity, the worst case for enrichment penalty (0.050 mills/kWhe) translate to an incremental penalty of about $14 million per year per 1 ,000 MT of fuel . TABLE 9.7 . Incremental Fuel Cycle Cost for Increased Enrich- ment Due to Recycle. of Acti nides Incremental U-235 Recycl e Number Cost (mills/kWhe) Enrichment (wt%) 0.0 3.30 1 0.022 3.375 2 0.032 3.402 3 0.0038 3.426 4 0.041 3.440 5 0.043 3.450 6 0.045 3.456 7 0.047 3.462 8 0.048 3.466 9 0.049 3.468 9.35 BNWL-1900 To put these costs in perspective, the incremental cost of $45 million (Table 9.6) is propagated to an in- cremental electrical cost. Sixty- 1,000 MWe PWRs require about 1,825 MT of fuel per year. These sixty PWRs operating at 85 percent capa- city produce 447 x 10 megawatt hours of electricity per year. Assuming electricity sells for 23 mills/kWh (i.e., $23/MWh), the total income would be $10.3 billion per year. From Table 9.6 the additional fuel cost required is approximately $80 million for these 60 reactors. Thus, the cost of electrical power would be increased by 0.8 percent per year. For comparison, it has been esti- (23) mated v ' that the cost of power for the Tennessee Valley Authority may double to meet all existing and planned pollution limits. 9.5.3 Alternate Fission Schemes Which Might Reduce Cost The case presented as a base case assumed that 10 percent of the U0 ? rods would contain all the actinides and would be remotely fabricated. In addition, it was assumed that these rods would be designed to be com- parable with a standard fuel cycle design. These are design constraints which affect the overall cost penalty. More detailed calculations may develop more optimum fueling se- quences and lower costs. It may be possible to place all the actinides in special target rods which are re- movable at the end of cycle but can be used initially for power shaping. These rods could have higher actinide concentrations and remain in the re- actor for longer periods of time. In this way the fabrication cost can be reduced and also amortized over a longer time period. Employing either one of these strategies could result in a significant decrease in the fa- brication penalty which is presently the controlling cost. If no Cf-252 and less than 0.1 per- cent of the fission products are present in the actinide waste stream, it may be possible to fabricate the ac ti ni de-bear i ng fuel assemblies in a glove box if the actinide concen- tration is not significantly higher than 0.5 wt% of the total fuel weight. This concentration limit may require the actinides to be spread over all the rods, resulting in an incremental cost of approximately $17/kg of fuel instead of $ 2 1 / kg for the base case. The enrichment cost would be $ 1 0/ k g of fuel instead of $14 in the base case, but the removal of Cf-252 and more fission products would require a cost of approximately $20/kg in- stead of $10 in the base case. The end result is a shifting of expenses but no real change in the total dol- lar figure (i.e., $47/ kg instead of $45/ kg for the base case). Recycling actinides in plutonium fuels in both LWRs and FBRs may ulti- mately be the best solution from a cost standpoint. In addition, since HTGR fuel will be fabricated remotely, the cost penalty projected in LWR fuel fabrication would not be ap- plicable if actinides were recycled in HTGRs. One of the early studies of plu- (24) tonium utilization v ; considered designing a fuel element which could 9.36 BNWL-1900 be used in both thermal and fast re- actors without chemically reprocess- ing. In light of current techno- logical problems of maintaining struc- tural integrity with zirconium clad LWR fuel and the nuclear waste man- agement considerations, perhaps this type of concept should be re-examined. The uncertainties in future occu- pational exposures allowed during fabrication of both LWR and FBR fuel will influence the base costs used to calculate incremental penalties. If these facilities must go to more re- mote operation, then the optimum strategy utilized for actinide recy- cle may change. It is quite possible that these uncertainties may actually lower the overall cost penalty asso- ciated with the recycle of actinides in a LWR. As a result, the cost penalty associated with the transmu- tation of actinides may actually be overstated by the values presented in Table 9.6. Consideration should also be given to evaluating the merit of having speci al -purpose reactors optimized for destroying actinides that are produced in commercial and government facilities. These reactors could be constructed and operated by the Federal government and the costs of this scheme added as a surcharge in the nuclear fuel cycle. 9.6 PUBLIC RESPONSE The actinide radionuclides repre- sent the greatest potential long-term risk from high-level waste. The po- tential risk to the public from these materials is thus present long after the benefits from nuclear power pro- duction have been received. The po- tential long-term risk is therefore directly related to the inventory of actinides, and transmutation can re- duce this inventory. Thus transmu- tation can result in lowering the potential risk to future generations. A pilot survey of the public response to alternative waste transmutation management concepts was made. Of the seven elements of risk (see Section 3 of Volume 1), the ability to have protective reaction, retriev- ability, and detectabi 1 i ty appear to be the most important factors. In the transmutation concept, the most toxic species of nuclear waste are kept in process. By keeping these materials in process, their location at any time is precisely known (i.e., can be easily detected and retrieved) Prior knowledge of accident proba- bility and consequences allows the development of adequate protective reaction measures. The material will certainly be in a stable configura- tion (e.g., clad fuel rods), and the utilization operations are nearly identical to those in current use in power reactors. The factors of dis- tance and population density should represent no more concern than the present ones for nuclear fuel cycle f aci 1 i ti es . 9.7 POLICY CONSIDERATIONS The transmutation concept is com- patible with existing policies and programs. Present federal policy^ ' (10 CFR 50, Appendix F) states that liquid "high-level radioactive waste" must be converted from aqueous to a dry solid 5 years after reprocessing 9.37 BNWL-1900 and the dry solid shipped to a fed- eral repository no later than 10 years after reprocessing. High-level waste is defined in this statement of policy as "those aqueous waste re- sulting from the operation of the first-cycle solvent extraction system, or equivalent, and the concentrated waste from subsequent extraction cycles, or equivalent, in a facility for reprocessing irradiated fuels." For the purposes of transmutation, this definition would have to be modified to the extent that actinides other than uranium and plutonium would not necessarily be classified as hi gh- 1 evel waste . The actinides in waste from other processing steps are classified as low-level waste because they are in dilute concentrations. Ultimately these materials may enter the high- level waste stream. The definitions of by-product, source, and special nuclear materials given in Parts 30, 40, and 70 of Title 10, Code of Fed- eral Regulations and the license re- quirements for owning and'handling these materials do not appear to pre- sent legal barriers to the transmuta- tion concept. and community. Actinide recycle in LWRs requires unique processing steps and some additional processing in some of the present fuel cycle steps. The additional processing is required because of the enrichment penalty as- sociated with the LWR recycle of actinides. The following discussions identify and quantify the additional processing requirements for the stan- dard fuel cycle components. The in- cremental environmental impacts of the additional processing steps are also discussed. 9.8.1 Incremental Increase in Fuel Cycle Processing In the transmutation base case, the U-235 enrichment required to ob- tain the same fuel cycle length had to be increased to 3.47 wt% U-235 instead of 3.3 wt% U-235. This analy- sis is based on a requirement for 1,825 MT/year of fuel (supplying sixty 1,000 MW(e) LWRs) processed at this higher enrichment. The addi- tional processing is shown in Table 9.8. Thus the uranium mine, mill, fluo- ride conversion, and enrichment 9.8 ENVIRONMENTAL CONSIDERATIONS Of the many transmutation concepts studied, actinide recycle in thermal fission power reactors (LWRs) is the only concept that utilizes an exist- ing operational device. Consequently, the analyses of incremental environ- mental impact have been limited to this concept. Environmental analyses consider the impact on the land, water, air TABLE 9.8 . Additional Uranium Re- quired to Supply 1 ,825 MT/ Year of Actinide Recycle Fuel (a) Additional Natural Uranium Additional Units of Separative Work 755 MT/Year 910 MT/Year a. Assuming diffusion tails composition of 0.30 wt% U-235 9.38 BNWL-1900 plants will experience higher through- put but no facility modification for actinide recycle. The remaining com- ponents of the fuel cycle, excluding transportation, do not experience any higher throughput although modifica- tions may be required to handle acti- nide recycle material. 9.8.2 Incremental Impact from In- creased Uranium Requirements The environmental impact from fuel cycle components has been quantified by the AEC.^ 25 ^ The following analy- sis is adopted from that study. Min- ing, milling, hexafluoride conversion and enrichment are considered. In every case, the impact is based on the requirement to produce 1,825 MT/ year of fuel to recycle actinides in a LWR. The incremental impact for all the additional processing is shown in Table 9.9. These numbers, although some appear large, represent an in- crease of about 8 percent in the pro- cessing requirements. Thus they do not significantly increase the pres- ent impact of the fuel cycle. 9.8.3 Incremental Impact of a Com- bined Reprocessing and Fuel Fabrication Plant for LWR Actinide Transmutation Fuel To minimize the amount of actinide transport required to carry out the transmutation concept, the fuel fabri- cation plant has been located adja- cent to the reprocessing plant. This represents a unique fuel cycle ar- rangement which does not currently exist and therefore has not been evaluated in any environmental state- ments. In addition, this preliminary study of transmutation has not been the subject of a detailed pathway analysis; thus dose values reported in this section are very preliminary estimates which must be verified at a later date using methods described in Section 3 of Volume 1. The incremen- tal change in dose received for pro- cessing transmutation fuel will be estimated by using a "Toxicity Index." As stated earlier in this report, it is recognized that this procedure gives limited insight on dose estimates . Two environmental assessments, one (25) for a typical reprocessing plant^ ; and one for a plutonium recycle fab- ( 26 ) rication plant v ' are used. The effect of plant construction and operation on the land, water and local community is essentially the same as the effect of existing fuel cycle facilities. Because the addi- tional facilities associated with fabrication and partitioning are easily sited in the exclusion radius around a reprocessing plant, the com- bined plant may actually result in better land utilization. The actinides produce more heat than conventional fuel; thus this ad- ditional heat would have to be dis- sipated in the environment. The entire annual inventory of 1,825 MT of actinide recycle fuel would gener- ate about 5 MW of thermal energy. Most of this heat would appear in process cooling water. In a properly designed system, there would be no significant environmental impact from that heat load. The process cooling water should contain no radionuclides. Present designs use an intermediate closed cooling loop with its own 9.39 BNWL-1900 TABLE 9.9 . Summary of Incremental Environmental Con- siderations for Uranium Mining, Milling, Fluoride Conversion and Enrichment for Actinide Recycle in LWRs (Based on 1,825 MT/Year of Fabricated Transmutation Fuel for a LWR) Hexafluoride Impact Mining Milling Conversion Enrichment Natural Resource Use Land (hectares) Temporarily Committed Undisturbed area Disturbed area Permanently committed Overburden moved (MT x 10" ) 0.2 Water (M 3 x IP" 3 ) 6.4 0.04 1.0 2.6 2.7 0.02 0.9 1.7 1.2 0.04 0.08 0.7 0.2 0.16 0.008 0.0 Discharged to air - 42.8 14.8 2712 Discharged to ground 81.2 Discharged to water bodies - - 156.8 325,500 Fossil Fuel Electrical Energy (MW-hr x 10) 0.4 0.47 2.3 2387 Equivalent coal (Mt x 10-3) 0.016 0.16 0.78 888 Natural Gas (m3 x 10-6) . 0.33 0.93 Effluents Chemical (MT) Gases S0 X o.62 6.4 30.9 3363 N0 X o.l6 2.7 10.3 8875 0.62 6.4 0.16 2.7 0.0016 0.23 0.004 0.06 Hydrocarbons 0.0016 0.23 0.66 87 CO 0.004 0.06 0.20 217 F" 0.12 4 Particulates 0.16 2.7 10.3 8875 Liquids Tailings Solutions 41,200 Ca++ 43 CI" 65 Na + 65 S0 4 = 43 Fe 3.0 N0 3 ' 22 Solids 15,860 41.2 Radiological (Curies) Gases (including airborne particulates) Rn-222 13-0 Ra-226 0.004 Th-230 0.004 U natural 0.006 0.014 0.015 Liquids U & daughters 0.4 0.4 0.15 Solids U & daughters 208 0.4 Thermal (MW-hr x 10" 3 ) 4 8 7595 9.40 BNWL-1900 cleanup system to transfer the pro- cess heat from the process equipment to the cooling water. Thus there are two isolation barriers separating the process equipment from the heat sink. The emissions to the air could change because of the additional pro- cessing required. Table 9.10 sum- marizes the results obtained using data from References (25) and (26) for present facilities. These must be adjusted to a combined reprocess- i ng-f abri cati on facility located at a remote reprocessing site. Ta- ble 9.11 summarizes the result for transmutation fuel reprocessed at the combined site. Preliminary estimates show that the releases from the fabri- cation part of the facility can be neglected in comparison to the acti- nide releases from the reprocessing plant. TABLE 9.10. Uranium Fuel Reprocessing' Fabrication Plants at In- dividual Sites - Gaseous Radioactive Effluents from Processing 1,825 MT/ Year Reprocessing Plant Releases Radionucl ide Organ of Interest Whole Body Annual Release, Ci Annual Dose, 1 ; mrem Kr-85 18 x 10 6 0.4 Skin 27.4 H-3 Whole Body 8 x 10 5 1.3 1-129 Thyroid 0.12 17.6 1-131 Thyroid 1.2 12.2 Other Fission Products Whole Body 51 <0.2 Transuranics Bone 0.21 0.31 Fabrication Plant Releases Plutonium Bone 4.0 x 10" 4 0.60 a. Estimated maximum exposure at plant boundary using currently licensed plants as a basis. TABLE 9.11 . Transmutation Fuel Reprocessing-Fabrication Plant Gaseous Radioactive Effluents from Processing 1 ,825 MT/Year Reprocessing Part of Facility Radionuclide Organ of Interest Whole Body Annual Release, Ci Ann ual Dose, 1 ' mrem Kr-85 18 x 10 6 0.4 Skin 27.4 H-3 Whole Body 8 x 10 5 1.3 1-129 Thyroid 0.12 17.6 1-131 Thyroid 1.2 12.2 Other Fission Products Whole Body 51 <0.2 Transuranics Bone 0.27 0.60 Fabrication Part of Facili ty Transuranics Bone 5 x 10" 5 0.000l' b ' a. Estimated maximum exposure at plant boundary using currently licensed plants as a basis. b. Reduced dose due to increased distance to site boundary for the combined facilities. The amount of krypton, tritium, iodine and volatile or semivolatile fission products emitted from the plant should not change significantly. The number of grams of transuranics released annually is expected to be the same for a transuranic plant as for a standard reprocessing plant. However, Claiborne^ ' estimates that toxicity index for airborne release has doubled for the composition of transuranics that have been recycled ten times. Thus, the exposure in Table 9.11 for transuranic processing was doubled. A comparison of Table 9.10 with 9.11 shows that a preliminary esti- mate of the maximum dose received by an individual at the plant boundary for the current fuel cycle are not significantly changed by transmuta- tion processing. 9.41 BNWL-1900 9.8.4 Environmental Impact from Waste Solidification Process i ng Five years after reprocessing, the liquid waste must be solidified. Based on a summary of solidifier op- erations, decontamination^ ' up to 10 are possible for nonvolatile (27) materials. ' Based on these values, the amount of transuranics emitted from the solidification process can be estimated. Table 9.12 summarizes the estimated emissions of radio- active waste for the LWR fuel and the transmutation fuel after 10 cycles. The emissions from the waste so- lidification step for transmutation were estimated assuming only 0.5 per- cent of the actinides appeared in the partitioned waste stream. The full reduction to 0.5 percent of the LWR levels is not realized because the isotopes build up from successive recycles. This buildup was estimated by Claiborne^ ' and was used to ad- just the LWR numbers to estimate the TABLE 9.12 Radiological Discharges from Waste Solidifica- tion (5 Years After Reactor Di scharge ) (1 ,825 MT/Year of Fuel ) Waste from LWR Fuel Waste from Transmutation Fuel Element Ci/yr q/y Neptunium 5.5 x 10 Uranium 3.3 x 10" Plutonium 9.3 x 10" A/nericium 3.1 x 10" Curium 3.8 x 10" Higher Isotopes 1.0 x 10" 9.7 x 2.6 x 5.5 x Ci/yr 1.8 x 10" 3.3 x 10 q/y 6.6 x 10 8.8 x 10" 1.3 x 10 2 x 10" actinide release from solidifying the waste from the transmutation process. The cumulative inventory of fission products is not significantly dif- ferent because the fuel was taken to the same goal exposure. Thus the in- cremental risk from solidification is adequately described by considering only the actinides. 9.8.5 Environmental Impact from Transporta ti on The major environmental impact of actinide recycle on transportation is associated with material processing. Although more uranium enrichment is required, the transportation aspects of uranium shipments have negligible environmental impact. The recycle of actinides represents shipments addi- tional to these, and their shipment results in an incremental impact. In addition, the reduced impact from transporting solidified waste with reduced actinide content is also i mportan t . Actinide recycle will increase the amount of material which has to be shipped between fuel cycle facilities The doses received along the routes could be significantly greater if no additional precautions were taken. One obvious way to minimize the transportation safety aspects of the problem is to have adjacent construc- tion of reprocessing and transmuta- tion fabrication plants. In this way no transportation of the transuranics would be necessary. The spent fuel casks are presently returned empty to the reactor. They could be used Decontamination factor is the ratio of material in the product feed to the material released to the environment. 9.42 BNWL-1900 to transport actinide recycle ele- ments. This would imply a two-cycle holdout period for actinides, which is not cost prohibitive. On a semi- annual refueling schedule a 1-year turnaround of actinides would still be possi bl e. Using these fuel cycle logistics it is estimated that the normal trans- port of 1,825 MT/year of transmuta- tion fuel elements to and from the reprocessor, a distance of 1,000 miles, would result in an annual whole body dose of 120 mrem/year to the population. ; This figure is double normal operation because twice as many shipments are required. The dose can be no greater than double be- cause the actinide recycle casks will have to be designed to the same ex- ternal radiation limits as present casks. The transmutation of actinides could reduce the impact of solid waste transportation in two ways. One reduction occurs because the vol- ume of waste solidified may be less. The other reduction results from the reduction in radiation, particularly neutron, which can be realized by extracting the actinides. Since the actinides make up less than 10 per- cent of the waste, the reduction in waste flow will not be considered. A reduction in exposure to people along the route from the reprocessing plant to the waste storage site may be realized. At the present time, if the waste from the processing of 1,825 MT/year of LWR uranium fuel is shipped 500 miles to a repository 10 years after reactor discharge, the exposure to the general populace along the route would be 60 mrem/ ( 28 ) year. ' The amount this could be reduced by actinide recycle is not known, but it could be significant. In this analysis no reduction is cons i dered . 9.9 SAFETY ASPECTS OF ACTINIDE RECYCLE IN LWRs The safety aspects of actinide re- cycle must consider the increased ac- cident potential from the additional uranium requirements and also the accidents from the resulting new pro- cesses. The potential for accidents in transportation steps must also be considered. In evaluating the potential for accidents, the results of higher ura- nium requirements are not significant First, the accident potentials from mining, milling, fluoride conversion, and enrichment are quite small. Sec- ondly, the additional uranium require- ment is only about 4 percent above the present demand. Thus the in- crease is relatively small. The po- tential hazards from the new process- ing steps are more critical. In particular, accidental releases in the reprocessing plant could be im- portant, and they are discussed separately. The safety aspects of the rest of the transmutation process- ing steps are included in the follow- ing discussion. 9.9.1 Effect of Actinide Recycle in LWRs on Accidents in Fuel Processing Plants The initial step in the reprocess- ing plant is fuel dissolution. This is followed by a separation of ura- nium and plutonium from the all fis- sion products, americium, curium and in some cases neptunium. This entire 9.43 BNWL-1900 stream is called the high-level waste stream. The wastes are then con- centrated and normally would be sent to liquid storage tanks. Additional treatment of the high-level waste stream is required to separate the other actinides. This may increase the risk of an accidental release in the reprocessing plant. A major concern with the addi- tional processing steps is the high level of radioactivity of the acti- nides. The alpha activity causes decomposition of solvents used in the chemical separation into gases which have explosive potential if they recombined. In addition, or- ganic solvents which increase the probability of a fire in the equip- ment will probably be required. The criticality accident at the head end of the reprocessing plant will probably be no more severe than the presently analyzed criticality event since the dose is from gases and semivolatile fission products. The actinides do not fall in that category. The next most frequently analyzed accident in reprocessing plant safety analysis reports is the explosion of a high-level waste con- centration tank. Most of the dose from this accident comes from ru- thenium. However, curium, if pres- ent in increased levels, could add to the hazard potential. The problem of a criticality in the actinide sepa- ration steps must also be analyzed. The severity of these potential ac- cidents is not known at the present time. 9.9.2 The Effect of Actinide Recycle in LWRs on Accidents in Other Fuel Cycle Components Following separation in the repro- cessing plant, the actinides would normally be transported to the fabri- cation plant. If the fabrication plant were in the same building, the risk from any additional transporta- tion step would be eliminated. In the fabrication plant, the haz- ard potential associated with recy- cling both plutonium and the higher actinides would be expected to in- crease. The release of plutonium dur- ing normal operation of an 1,825 MT/ year plutonium fabrication plant is ( 26 ) 300 yg/year. ' The release of ac- tinides from a similar sized actinide recycle plant would be about 30 ug/ year, but the potential hazard repre- sented by this release should be about doubled, due to the nature of material released. ' This conclu- sion was obtained from the observa- tion that the actinide inventory in an actinide recycle fuel fabrication plant will be one-tenth the actinide inventory of a similar sized pluto- nium fuel fabrication plant. It should be noted that it is assumed here that the actinide recycle plant is fabricating uranium fuel, not plu- toni urn fuel . A spent fuel cask may have to be used for transportation of the recy- cle assembly from the fabrication plant to the reactor. However, since there is no volatile material in an assembly, the hazards associated with 9.44 BNWL- 1 900 this transportation step are expected to be quite smal 1 . In the reactor, the increased risk associated with higher actinide re- cycle is not significantly greater than the risk associated with normal reactor operation. There is no pos- tulated accident that could produce significant quantities of particu- lates, and thus the introduction of nonvolatile actinides does not in- crease the risk. Following irradiation, the fuel must be shipped back to the repro- cessing plant. In this step there is no accident postulated which re- leases nonvolatile materials. Thus transporting the spent fuel should not be an increased risk consi derati on . The accident potential for the waste solidification step is quite small. The hazards, although small, would be reduced as a result of transmutati on . This analysis has identified only two areas where the risk of transmu- tation could be significantly in- creased. These are at the reprocess- ing plant and during fabrication. All other fuel cycle components have risks which are either reduced or not significantly different from the risks already associated with those components . At the present time, the conse- quences of accidents in reprocessing plants and plutonium fabrication plants are being investigated in a separate AEC study. When this inves- tigation is complete, it should be possible to quantify the increased accident potential for transmutation processing. Until that time, the risk will be evaluated using only the chronic release emissions. REFERENCES 1 . Code of Federal Regulations , Title 10, Part 50, Appendix F. 2. G. A. Bartholomew, "Spallation- Type Thermal Neutron Sources," Seminar on Intense Neutron Sources, C0NF-660925, September ber 19-23, 1966, Con ference Proce edings, TID-4500, p. 637. 3. M. Steinberg, G. Wotzak and B. Monowitz, Neutron Burning of Long Lived Fission Products for Waste Pi sposal , BNL-8558, Brookhaven National Laboratory, 1964. 4. F. L. Culler, Deputy Director, Oak Ridge National Laboratory, Letter to V. M. Staebler, USAEC, December 28, 1 971 . 5. Torso Ichimiya, Director of Institute of Physical and Chemi- cal Research, Japan, Private Communication to B. R. Leonard, Battelle, Pacific Northwest Laboratories, 1972. 6. M. Goldstein and E. Nolting, Proposal No. I BR- 72-2706 , Inter- national Business and Research, Inc. Proposal to USAEC January 24, 1972 revised February 11, 1972. 7. A. E. Bolon and D. R. Edwards, "Eliminating Radioactive Wastes by Underground Thermonuclear Ex- plosion," Research Proposal to USAEC, University of Missouri, May 1972. 8. N. F. Colby, Use of Isotopes To Reduce Neutron-Induced Radioac- tivity and Augment Thermal Quality of the Environment of a an Underground Explosion , M. S. Thesis, University of Missouri, May 1972. 9. Michael V. Gregory and Meyer Steinberg, A Nuclear Transforma- tion System for Disposal of Long- Lived Fission Product Waste in an Expanding Nuclear Power Economy , BNL-11915, Brookhaven National Laboratory, November, 1967. 9.45 BNWL-1900 10, 11 H. C. CI a i borne, f Transmutation of eutron- Induced High-Level Radio- active Waste , ORNL-TM-3964, Oak Ridge National Laboratory, Decem- ber 1972. A. S. Kubo, Technology Assessment of High-Level Nuclear Waste Management , ScD Thesis, Depart- ment of Nuclear Engineering, Massachusets Institute of Tech- nology, Apri 1 1 973) . A. S. Kubo and D. J. Rose, "On Disposal of Nuclear Waste," Sci- ence , vol. 182, no. 4118, pp. 1205- 1211 , December 21 , 1973. E. T. Merrill, ALTHAEA-A One- Dimensional Two-Group Diffusion Code with an Effective Four-Group Burnup , BNWL-462, Battelle, Paci- fic Northwest Laboratories, May 1971 . 14. M. J. Bell, ORIGEN-The ORNL Iso- 12. 13 15. tope Generation and Depletion Code , ORNL-4628, Oak Ridge National Laboratories, May 1973. W. C. Wolkenhauer, "The Controlled Thermonuclear Reactor as a Fission Product Burner," Trans . Am. Nucl . Soc. , vol. 15, no. 1, p. 92, 1972. 19. B. F. Gore and B. R. Leonard, Jr., "Transmutation in Quantity of 1 3 7cs in a Controlled Thermonu- clear Reactor," Trans . Am. Nucl . Soc . , vol. 17, pp. 52-53, November 1973. 20. B. F. Gore and B. R. Leonard, Jr., "Transmutation of Massive Load- ings of 137cs in the Blanket of a Controlled Thermonuclear Reac- tor," Nucl . Sci . Eng . , vol. 53, pp. 319-323, March 1974. 21 , 22 23 24 Nuclear Power 1972-2000 , WASH-1139 (72), December 1 , 1972. L. G. Merker and ' Compa ra tive Costs vol. 1, J. B. Burnham, D. E . Deoni gi , of Oxide Fuel Elements BNWL-273, Battelle, Pacific North- west Laboratories, July 1966. A. J. Wagner, "Power Environment, and Your Pocketbook," Public Utilities Fortnightly, vol. 89, no. 13, pp. 27-31 , 1972. H. Nail Fuel Re- R. J. Hennig and J. Progress Report on II: Analysis of Four Fast- Thermal Reactor Complexes , BNWL-254, Battelle, Pacific Northwest Laboratories, July Use, 1968 16. 17 W. C. Wolkenauer, "The Controlled Thermonuclear Reactor as a Fission Product Burner," BNWL-SA-4232 , Battelle, Pacific Northwest Labora- tories, June 1972. W. C. Wolkenhauer, B. R. Leonard, Jr., and B. F. Gore, Transmuta- tion of High-Level Radioactive Waste with a Controlled Thermonu - clear Reactor, BNWL-1772, Battelle, Pacific Northwest Labora- tories, September 1973. W. C. Wolkenauer, B. R. Leonard, Jr., and B. F. Gore, "Transmutation of High-Level Radioactive Wastes with a Controlled Thermonuclear Reactor," Trans . Am. Nucl . Soc. , vol. 17, p. 52, November 1973. 25 26 Environmental Survey of the Nu- clear Fuel Cycle , USAEC, November 1972. J. M. Selby et al . , Consider a- tions in the Assessmen t of the Consequences of Effluents from Mixed Oxide Fuel Fabric ation Plants, BNWL-1697, Revised 1974. 27. J. L. McElroy, Waste Solidifica - tion Program Summary Report , BNWL-1667, Battelle, Pacific Northwest Laboratories, 1972. 28. Environmental Survey of Transpor- tation of Radioactive Materials to and from Nuclear Power Plants , WASH-1238, USAEC, December 1972. BNWL-1900 APPENDIX SECTION 8: EXTRATERRESTRIAL DISPOSAL BNWL-1 900 APPENDIX 8. A NASA Executive Summary Feasibility of Space Disposal of Radioactive Nuclear Waste NASA TECHNICAL MEMORANDUM Csi ■ X NASA TM X 2911 %S»I FEASIBILITY OF SPACE DISPOSAL OF RADIOACTIVE NUCLEAR WASTE I - Executive Summary Lewis Research Center Cleveland, Ohio 44135 NATIONAL AERONAUTICS AND SPACE ADMINISTRATION • WASHINGTON, D. C. • DECEMBER 1973 1, Report No. NASA TMX-2911 2. Government Accession No. 3. Recipient's Catalog No. 4. Title and Subtitle FEASIBILITY OF SPACE DISPOSAL OF RADIOACTIVE NUCLEAR WASTE I - EXECUTIVE SUMMARY 5. Report Date December 1973 6. Performing Organization Code 7. Author(s) National Aeronautics and Space Administration 8. Performing Organization Report No. E-7679 9. Performing Organization Name and Address Lewis Research Center National Aeronautics and Space Administration Cleveland, Ohio 44135 10. Work Unit No. 770-18 11. Contract or Grant No. 12. Sponsoring Agency Name and Address National Aeronautics and Space Administration Washington, D.C. 20546 13. Type of Report and Period Covered Technical Memorandum 14. Sponsoring Agency Code 15. Supplementary Notes 16. Abstract This NASA study, performed at the request of the AEC, concludes that transporting radio- active waste (primarily long-lived isotopes) into space is feasible. Tentative solutions are presented for technical problems involving safe packaging. Launch systems (existing and planned), trajectories, potential hazards, and various destinations were evaluated. Solar system escape is possible and would have the advantage of ultimate removal of the radioactive waste from man's environment. Transportation costs would be low (comparable to less than a 5 percent increase in the cost of electricity) even though more than 100 Space Shuttle launches per year would be required by the year 2000. 17. Key Words (Suggested by Author(s)) Space Shuttle; Nuclear waste; Radioactive waste; Space tug; Waste disposal; Actinides 18. Distribution Statement Unclassified - unlimited 19. Security Classif. (of this report) Unclassified 20. Security Classif. (of this page) Unclassified 21. No. of Pages 20 22. Price* Domestic, $2.75 Foreign, $5.25 * For sale by the National Technical Information Service, Springfield, Virginia 22151 FOREWORD An exploratory study to assess the feasibility of sending radioactive waste materials generated by the nuclear power industry into space for disposal was conducted by the National Aeronautics and Space Administration (NASA) and is summarized in two volumes: I - EXECUTIVE SUMMARY and II - TECHNICAL SUMMARY. The study was performed at the request of the Atomic Energy Commission (AEC) as part of a review of various storage and disposal concepts for nuclear waste management. The study was performed by personnel from various NASA centers, NASA Headquarters, and the AEC. The various sections of the two volumes were written by members of the group and compiled by Robert E. Hyland of the NASA Lewis Research Center. The principal contributors and their respective areas of contribution are as follows: Robert E. Hyland Coordinator, package concept and reports NASA Lewis Research Center Robert Thompson Destinations, vehicles, and trajectories NASA Lewis Research Center Richard L. Puthoff Impact and postimpact conditions NASA Lewis Research Center Millard L. Wohl Shielding, impact, and fragmentation NASA Lewis Research Center Ruth N. Weltmann Nuclear safety NASA Lewis Research Center (Aerospace Safety Research and Data Institute) John Vorreiter Reentry shield NASA Ames Research Center Nathan Koenig Launch site and facilities NASA Kennedy Space Center Victor Bond Trajectories NASA Johnson Space Center Gus Babb Shuttle integration NASA Johnson Space Center Herbert Shaefer Nuclear safety, HQ monitor NASA Headquarters Thomas B. Kerr Nuclear safety NASA Headquarters Thaddeus J. Dobry Nuclear safety Atomic Energy Commission Robert W. Ramsey AEC /NASA coordinator Atomic Energy Commission 111 8.A.1 FEASIBILITY OF SPACE DISPOSAL OF RADIOACTIVE NUCLEAR WASTE I - EXECUTIVE SUMMARY National Aeronautics and Space Administration Lewis Research Center SUMMARY The concept of disposing of radioactive waste into space was studied and found to be feasible. Tentative solutions are presented for technical problems of safely packaging the separated long-lived actinide wasted. Disposal of these wastes is the primary con- cern because they will remain radioactive for extremely long times. The package design includes shielding to achieve reasonably low external levels of radiation. The logistics and potential hazards of launching these packages into either high Earth orbits or solar orbits or to escape the solar system have been evaluated. These destinations have been found to be the most promising. Although the solar system escape requires greater energy, it appears to be the most desirable for ultimate disposal. The total costs of a system for space disposal of radioactive waste are based on the rate of accumulation of fission products and uranium -free actinides in reprocessing plants serving the nuclear power industry and on the launch costs, the destinations, and the launch frequency. The number of waste packages to be launched per year depends on the degree of separation of the long-lived actinides. For example, a package containing about 200 kilograms of separated actinide wastes with about 0. 1 percent residual fission products could be ejected out of the solar system for a cost of about $150 000 per kilogram. Fifty to 100 Space Shuttle launches of such packages per year would be required in the 1990- 1995 time period to handle the actinide waste. To this cost must be added the estimated cost of separating and encapsulating the actinide waste. Although the space transportation cost would be several billion dollars per year, the cost prorated over the nuclear electrical capacity is less than 0. 1 cent per kilowatt-hour. A packaging design concept has been evolved that appears on a qualitative basis to provide protection against the radioactive waste in accident environments. The concept, however, does need a follow-up experimental program and safety assessment to estab- lish a system design. 8. A. 2 INTRODUCTION This report (part I) is a condensed summary of an exploratory study (part II) of the feasibility of radioactive waste disposal into space performed by the National Aeronautics and Space Administration (NASA) at the request of the Atomic Energy Commission (AEC). This study was conducted to provide a preliminary assessment of the safety of contain- ment and of launch capability and estimates of transportation costs. It is to be factored in with other studies on potential means for long-term management of high-level radio- active wastes. Battelle Pacific Northwest Laboratories coordinated these studies under contract to the AEC. RADIOACTIVE WASTE ACCUMULATION The electric power industry in the United States is projected to have an installed nuclear capacity that may reach 1000 gigawatts electric by the year 2000» The yearly production rate of nuclear wastes that accompany the increasing nuclear capacity in the U.S. is presented in figure 1. The nuclear wastes consist of fission products and acti- nides (i.e. radioactive elements above actinium, such as neptunium, plutonium, and curium) o ^ 60G 5» 400 £ 200 "a! E 100 2 80 2 60 o 40 15 20- 10 8 6 4 2h ■ Fission product waste Actinide waste 1970 1980 1990 2000 Year Figure 1. - Projected nuclear waste from U. S. powerplants. 8. A. 3 Integration of these rates indicates that by the year 2000 about 9000 metric tons of fission products and 1200 metric tons of actinides will have been accumulated. This assumes that no transmutation of actinides has taken place by further in-pile irradiation. The actinide inventory can be reduced to 300 metric tons by separation of essentially all uranium isotopes. This residual actinide inventory is the waste that is considered in the study. Transmutation of actinides, assuming neutron flux levels in typical pressurized water reactors, could reduce this inventory to about one -third if in-pile transmutation were considered feasible. Many of the actinide isotopes have half -lives measured in tens and hundreds of thousands of years. Representative fission products and actinides are described in table 1» These materials represent a long-term hazard to man and must be either stored or disposed of in an acceptable manner. For some of the isotopes with long half -lives, this could mean several hundred thousand years for storage. TABLE 1. - SOME RADIOACTIVE ISOTOPES WITH LONG DECAY TIMES Waste Isotope Half-life, yr Decay processes Fission products Tritium ( H) 12.3 Beta (electron) Strontium-90 27.7 Beta (electron) Technetium-99 2xl0 5 Beta (electron) Iodine- 129 1.6xl0 7 Beta (electron), gamma ray Cesium-137 30 Beta (electron), gamma ray Samarium- 151 87 Beta (electron), gamma ray Actinides Plutonium-239 2.4xl0 4 Alpha (He) particle, gamma ray Neptunium-237 2. lxlO 6 Americium-241 458 Americium-243 7.6xl0 3 Curium-244 18 \ } SPACE DESTINATIONS The potential space destinations considered were narrowed down to high Earth orbit, solar orbit, and solar system escape. They are illustrated in figure 2. 8. A. 4 First burn \ Y'" Earth orbit \ (a) High Earth orbit. Velocity increment from low earth orbit, AV, 4.11 km/sec; single shuttle launch to 370-km orbit; two burns to- 90 000- km circular orbit (above synchronous orbit); time between burns, ~ 20 hr. (b) Solar orbit to 0.9 AU. Velocity increment, AV, 4. 11 km/sec; single shuttle launch to 370-km or- bit; two burns to circular solar orbit (0.9 or 1. 1 AU); time between burns, ~6 months. ? fcD — — (c) Solar system escape. Velocity increment, AV, 8.75 km/sec; two shuttle launches to 370-km orbit (one shuttle carries payload and expendable tug, the other carries re- usable tug); two burns at perigee; time be- tween burns, ~8 hr. Figure 2. - Potential space destinations. HIGH EARTH ORBIT Placing waste packages in high Earth orbits (about midway between synchronous orbit and the lunar orbit) requires a relatively low increment in velocity (4. 1 -km/sec change in velocity from parking orbit). Daily launch opportunities exist for such flights. Retrieval of waste packages from such orbits is reasonable - Until the long-term integ- rity of the waste package can be guaranteed, such orbits can be considered as interim storage destinations for only hundreds to thousands of years- 8. A. 5 SOLAR ORBIT Solar orbits (nearly circular at ~0. 9 AU) can be achieved with a relatively low incre- ment of velocity (4. 1 km/sec) and also can take advantage of daily launch opportunities. Their disadvantage is that the circularization burn occurs approximately 1/2 year after injection into the transfer orbit, therby reducing the reliability of a successful circular- ization. A malfunction at that time could lead to a possible Earth encounter. Since the long-term stability of such orbits is uncertain, they are not recommended for permanent disposal at this time. SOLAR SYSTEM ESCAPE Although direct escape from the solar system requires a high increment in velocity (8. 75 km/sec), such disposal of radioactive waste from man's environment is permanent. Furthermore, the integrity of the package is required for a much shorter time period (years as compared with hundreds of centuries) since it will leave our solar system. The solar system escape launch appears to be the most desirable and was found to be economically and technically reasonable. OTHER DESTINATIONS CONSIDERED Sending the waste packages directly into the Sun is not possible with present launch vehicles. Indirect flight could be accomplished with present vehicles by using the more advanced planet swing-by trajectories. However, this is not practical because of limited launch opportunities. Lunar and planetary destinations were not considered because of the possibility of planet contamination and the very high increment in velocity required for soft landings. SPACE TRANSPORTATION VEHICLES The launch vehicles and space tugs considered were those that are available or are being planned and consist of expendable and reusable stages. They are shown in figure 3. The corresponding vehicle launch costs are shown in table 2. The Space Shuttle, in conjunction with either reusable or expendable space tugs, provides the lowest cost per kilogram of payload (total weight of waste package) delivered to the various destinations. Tables 3 and 4 summarize the costs for the various launch vehi- cles. Because the shuttle is a manned vehicle, its use considerably enhances the 8. A. 6 100 so 60 40 20 L_ Saturn V A A A wm Titan IIIE/Centaur Space Shuttle Figure 3. - Space transportation systems. A a Space tug/waste package CD-11570-31 TABLE 2. - SPACE TRANSPORTATION VEHI- CLE LAUNCH COST FOR RADIOACTIVE WASTE DISPOSAL MISSION Launch vehicle Launch cost. dollars Titan IIIE/Centaur 19.00xl0 6 Saturn V/Centaur 155.00 Space Shuttle: 10. 50 Reusable tug 1.75 Expendable tug 5.80 TABLE 3. - LAUNCH VEHICLE PERFORMANCE AND COST SUMMARY FOR HIGH EARTH ORBITS AND SOLAR ORBITS. [Velocity increment. AV. 4.11 km/sec. ] Launch vehicle Payload. Launch cost. kg dollars /kg Titan IHE/Centaur 3 860 4920 Saturn V 32 660 4590 Saturn V/Centaur 35 290 4390 Space Shuttle: Reusable tug (current size) 4 170 2940 Reusable tug (optimum size) 4 670 2620 Centaur (current size) 6 490 2460 Centaur (optimum size) 8 480 1920 8. A. 7 TABLE 4. - LAUNCH VEHICLE PERFORMANCE AND COST SUMMARY FOR DIRECT SOLAR ESCAPE MISSION. [Velocity increment, AV, 8. 75 km/sec] Launch vehicle Payload, kg Launch cost, dollars Cost, dollars/kg Saturn V/Centaur 7480 155X10 6 20 720 (2. 1, 1) Shuttle/tug configuration: 3 Without perigee propulsion With perigee propulsion 2270 3270 28.75xl0 6 28.75 12 660 8 790 (3, 1, 2) Shuttle/tug configuration: Without perigee propulsion With perigee propulsion 3040 4400 41.0X10 6 41.0 13 490 9 320 a Two shuttle launches, one expendable tug, one reusable tug. b Three shuttle launches, one expendable tug, two reusable tugs. reliability of the mission from launch to ignition of the tug engine following deployment. For the waste package mounted on a space tug within the manned shuttle orbiter, a dose level of 1 rem per hour at 1 meter from the surface of the package has been assumed* This value is reasonable and can be further attenuated by distance and by inter- vening structure in order to reduce the dose to the crew. The waste package will be subcooled prior to launch. Upon reaching orbit its decay heat will raise the package tem- perature., This heat will be dissipated by radiation to space when the cargo bay doors are opened. Reflectors will be provided in the cargo bay to direct the heat out through the bay door opening. A typical shuttle launch -to -landing sequence is shown in figure 4. The shuttle launch vehicle is assisted at lift-cff by two solid-fueled rocket motors. These are separated and dropped for recovery while the orbiter continues, fueled by the expendable external fuel tank. This external tank is jettisoned and deorbited by a small retrorocket. The orbit- er' s payload (waste package plus tug) is deployed from the bay of the orbiter. The orbiter later returns and lands at the prescribed landing site. Depending on the destination, the space tug with the waste package either awaits a second tug (solar escape) or initiates its firing sequence to place the package on its desired trajectory. A method of mounting the space tug with the waste package in, and deploying it from, the orbiter is shown in figure 5. If a malfunction should occur after deployment and before initiation of propulsion by the tug, the orbiter could retrieve either or both. If the malfunction were to occur in later stages of the mission, another tug, capable of retrieving the package, would be dispatched. 8. A. 8 Launch Solid-fueled rocket -motor (SRMI burnout and jettison Landing CD-11566-31 Figure 4. -Space Shuttle launch-to-landing sequence. 8. A. 9 (a) Mounted in cargo bay. CD-11568-31 (b) Readied for deployment. Figure 5. - Space Shuttle orbiter with nuclear waste package and tug. NUCLEAR WASTE PACKAGE DESIGN The radioactive wastes from nuclear powerplants can be processed to separate them into two waste streams: fission products, and actinides with residual amounts of fission products- Fission products in various concentrations were assumed to remain in the actinide waste because the cost of complete separation would be too great. A representative package design is shown in figure 6, The radioactive wastes are contained within a storage matrix which acts as a partial neutron and gamma shield as well as a heat-conducting medium. The actinide waste is in the form of small spheres approximately 3. 5 millimeters in diameter- The spheres are coated with a refractory metal and an oxidation -resistant material for retention of radioactive waste at high temperatures. The matrix, containing approximately 10 percent actinides by volume, is enclosed in a sphere of stainless steel to protect it against impact and fragmentation. This sphere also contains layers of neutron and gamma shielding material. The impact protection sphere is enclosed within an aerodynamically stable reentry body designed to 8. A. 10 1.5 to 3.0 m Radiation shielding Impact sphere Blunt-cone reentry body (heat shield) CS-65974 CD-11567-31 Figure 6. - Representative nuclear waste package. TABLE 5. - WEIGHT BREAKDOWN OF TYPICAL NUCLEAR WASTE PACKAGE [Solar system escape for actinides. ] Component Weight, kg Actinide waste Matrix containing waste Gamma shield Neutron shield Impact sphere Reentry body (heat shield) 200 625 1190 180 640 410 Total 3245 8. A. 11 survive reentry heating. This reentry body consists of two layers. The outer layer is a composite fiber of quartz woven into a mat with a silica binder that acts as a highly reflecting medium for steep-angle reentry protection. The inner layer is composed of 3D graphite to handle the convective heat load from shallow -angle reentries. A biological dose constraint of 1 rem at 1 meter from the surface of the waste pack- age was assumed for the configuration that was designed for solar escape. The package is thus weight optimized to contain about 1 kilogram of waste for every 30 kilograms of total package weight when actinide wastes contain 1 percent residual fission products. As the composition of fission products is reduced to 001 percent, the weight fraction of actinide waste increases to 1 kilogram in every 10 kilograms of package weight. These optimized weights are essentially independent of space destination A representative package weight breakdown is presented in table 5. For some of the heavier pay loads considered, the heat generated by the radioactive waste was a limiting factor in the design of the waste package. The waste package design concept presented in part II of this report would be applicable for disposal of other compositions of radioactive waste. TYPICAL DISPOSAL MISSION The sequence of events for a typical waste disposal mission to solar system escape is as follows: (1) Launch shuttle 1 to 370 -kilometer parking orbit. (2) Deploy reusable tug to rendezvous position. (3) Launch shuttle 2 to 370-kilometer parking orbit. (4) Deploy expendable tug and waste package to rendezvous with reusable tugo (5) Maneuver tugs to dock in tandem configuration. (6) Reusable tug fires to required AV, separates, and returns to shuttle 2. (7) Expendable tug fires and injects waste package into solar system escape trajectory. The major components involved in such a mission are shown in figure 7. 8. A. 12 Components -Waste -Shielding 'Impact sphere -Reentry body -Adapter Nuclear waste package Orbiter Liquid-propellant tank Space Shuttle Xl =l~/'~ y v ^'t^V. -11m Reusable space tug v . n 41m Expendable space tug CO-11569-31 Component Weight, kg Nuclear waste package: Waste (actinides plus 0. 1 percent fission products) Shielding (LiH, W, matrix) Impact sphere Reentry body (heat shield) Adapter 200 1995 640 410 120 Space Shuttle: Orbiter (dry weight) Liquid propellant and tank Solid rockets 68 000 737 000 1030 000 Reusable space tug: Propellant weight Burnout weight 23 900 2 900 Expendable space tug: Propellant weight Burnout weight 22 000 2 900 Figure 7. - Component weights for nuclear waste space disposal mission. Required for mission.- one shuttle carrying reusable space tug, and another shuttle carrying expendable space tug and nuclear waste package. LAUNCH FREQUENCY The frequency of Space Shuttle launches required is an important factor in considering the space destinations, the costs, and the launch facility requirements. For each radio- active waste composition and each disposal package design, the number of required annual shuttle launches was determined through the year 2010 for the three space destinations described. The high Earth orbits and the circular solar orbits require approximately the same number of annual flights and are plotted together in figure 8. This figure is for the extreme case of disposing of all fission products that have been ground stored for 10 years. 10* 8. & 8. A. 13 External dose rate from waste package, rem/hr 2010 Year Figure 8. - Number of Space Shuttle launches required per year for disposal of all fission products into solar orbit or high Earth orbit. Prior 10-year Earth storage. Fission products in actinide waste, percent 8. 1980 Solar ► system escape .001 Curium removed 1990 2000 2010 Year Figure 9. - Number of Space Shuttle launches required per year for disposal of only actinides into high Earth orbit or by solar system escape. Prior 10- year Earth storage. 8. A. 14 After 1990, more than one launch per day would be required. This launch frequency is not considered practical at this time. The launch frequencies required for space disposal of only the separated actinides are more reasonable, as shown in figure 9. Required launch rates vary from less than 10 to 350 per year through the year 2010 depending on the fission product composition of the actinides and the destination. With the launch facilities that are available and that could be made available, as many as 140 launches per year are possible. The estimated cost for additional equipment and facilities to handle this many launches per year is $230 million. This cost includes two new launch pads. SPACE TRANSPORTATION COSTS The launch costs for the shuttle and tugs, presented in table 2, are the overriding space transportation costs „ These costs, coupled with the packaging cost ($650/kg of actinides) and the expense of additional launch facilities (estimated at $70 000/flight or 140 flights/yr), determines the costs for transportation of radioactive waste to the space destinations considered. (The cost of separating the fission products from the actinides is not included here.) The waste disposal per mission and the space transpor- tation costs are presented in table 6„ To present these costs in perspective, they may be put in terms of the additional power cost to the consumer .(i. e. , space transportation josts per kW-hr of electrical power generated in producting the nuclear waste). The space transportation cost for the disposal of all the fission products is 1 to 5 cents per kilowatt- hour, For disposing of only the separated actinides, the cost is 0.01 to 0, 1 cent per kilowatt-hour. The cost depends on the space destination and on the composition of residual fission products contained within the actinides . The results of an optimization study that balanced estimated fission product separation costs against waste package transportation costs are shown in figure 10 and point to a fission product composition of less than 1 percent as desirable. Compared with the present cost of electricity, the space disposal of the separated actinide wastes represents less than a 5 percent increase in power costs to the consumer. Adding the estimated cost of separating fission products to the cost of transporting the waste to space yields the total cost. The optimum total cost, 0. 1 cent per kilowatt-hour, occurs for an actinide waste containing about 0. 1 percent fission products and for disposal beyond our solar system. The total annual costs for transporting actinides containing 0„ 1 percent fission products after a 10 -year temporary storage on Earth, as shown in figure 11, range from $30 million to $5 billion per year. 8. A. 15 TABLE 6. - TRANSPORTATION COSTS FOR DISPOSAL OF RADIOACTIVE WASTE Type of waste Destination Amount of Transportation for disposal waste disposed cost. of per mission. do liars /kg kg Fission products Earth orbit or solar orbit 189 88 000 Solar system escape 73 394 000 Actinides plus 1 percent Earth orbit or solar 288 57 000 fission products orbit Solar system escape 113 255 000 Actinides plus 0. 1 percent Earth orbit or solar 447 37 000 fission products orbit Solar system escape 200 151 000 Actinides plus 0.001 percent Earth orbit or solar 858 19 000 fission products orbit Solar system escape 308 94 000 Mission launch system: for high Earth or solar orbit, Space Shuttle with Centaur (optimum size); for solar system escape, two Space Shuttles, one reusable tug, and one expendable tug. Includes cost of packaging and additional launch facilities but not the separation cost. Cost — Total Space transportation Estimated separation 10 1 .1 .01 .001 Fission products in actinide waste, percent Figure 10. - Optimization of costs for space disposal of actinide waste by solar system escape. 8. A. 16 a ioic 10 v 10 e 10' Fission products in actinide waste, perc Solar system escape High Earth orbit or solar orbit 1980 1990 Year 2000 Figure 11. - Annual space transportation cost for space disposal of actinide waste. SAFETY The safety goal for nuclear waste disposal in space is to transport the radioactive waste to an acceptable destination in such a manner that potential radiation exposures and contamination are negligible. The accident conditions considered and the responses of the design waste package are summarized in table 7» In all cases the response of the waste package to the proposed accidents indicates that the release of radioactive waste would be prevented by the various protection shells designed into the total waste package The package response analysis was verified, where possible, by simulation experiments. However, much additional development and testing are required to confirm the design concept. TABLE 7. - POSSIBLE ACCIDENTS AND PACKAGE RESPONSES Type of accident Accident condition Package response Blast overpressure 150 atm No yielding to 175 atm Fragmentation Fragments up to 1070 m/sec No penetration to 1360 m/sec Fireball 2750° C, 20 sec No melting Residual fire 2400° C, 5 min Outer stainless-steel layer near melting Reentry heating 300 kW/cm 2 , 3 to 4 sec Sufficient thickness Impact on earth, water, 300 m/sec Some deformation, no release or concrete Postimpact Deep burial Outer vessel rupture due to pressure after about 5 days Deformed - no burial Integrity maintained 8. A. 17 With an appropriate package design and launch operation, the overall risks are expected to be low. Since the mission hardware and launch parameters were of a pre- liminary nature only, a risk assessment on a quantitative basis could not be performed. The key requirement for the overall safety of waste disposal missions is early recov- ery of the waste package in the event of an accident during any phase of the mission. For most accidents the early recovery could be handled satisfactorily. For some accidents, particularly an uncontrolled abort occurring in the later stages of a mission (i.e. , after deployment and prior to the tug achieving the required AV), recovery from space may be difficult if not impossible,. CONCLUSIONS GENERAL The results of this exploratory study indicate that space disposal of the long-lived radioactive actinides from nuclear .waste appears feasible from the viewpoint of both econ- omy and safety. The transportation costs for ejecting the actinides out of the solar system, for example, would represent less than 5 percent increase in the consumer bill for elec- tric power generated by nuclear powerplants. Such missions involve certain risks, how- ever small, which would have to be balanced against the benefits to be derived from re- moving the long-lived radioactive waste from man's environment and thus relieving future generations of the responsibility of protecting themselves against our radioactive waste. Quantitative evaluation of the risks requires more study, development, and testing. SPACE DESTINATIONS Of the destinations considered, three look promising: high Earth orbits (above synchronous orbit altitude), nearly circular solar orbits inside the Earth's orbit, and solar system escape. Only the last destination provides a permanent disposal of the nuclear waste. It is therefore the most promising destination, even though the cost- liest. Sending the waste directly into the Sun is not within the capabilities of present vehicles. Sending it into the Sun with acceleration assists from planetary swing-by is not practical- 8. A. 18 SPACE TRANSPORTATION VEHICLE The currently planned Space Shuttle, in conjunction with space tugs, provides a substantially lower cost per kilogram of waste delivered to the space destinations than any of the current expendable launch vehicles. Because the shuttle is manned and has considerable maneuvering capability, the overall safety aspects of such a transportation system could be superior to those of expendable launch vehicle systems. WASTE PACKAGE DESIGN CONCEPT The nuclear waste package design allows sufficient radioactive waste per package for economic disposal and should prevent release of radioactive waste under the accident conditions reviewed., Further study could optimize the design to increase the waste con- tent and to better define its limitations. SAFETY No quantitative risk assessment was possible because the mission hardware and the mission parameters are preliminary. Only a qualitative evaluation was performed. This evaluation indicated the design could prevent release of radioactive waste under conditions imposed in accident environments. With appropriate system design and launch opera- tions, the risks involved are expected to be relatively low, COSTS The transportation costs for space disposal of radioactive actinides would represent an increase in the consumer's electric costs of approximately 5 percent. To this trans- portation cost must be added the cost for separating the actinide waste and the fission product waste. Preliminary data from a study conducted by Battelle Pacific Northwest Laboratories for the Atomic Energy Commission indicate that the separation costs will be of the same order or less than the costs of transportation out of the solar system. Both the space transportation cost and the launch frequency are feasible and practical for the disposal of separated actinide waste. However, the space disposal of all fission product waste is neither economically nor practically feasible at this time because the large quantities would require an excessive launch rate. Lewis Research Center, National Aeronautics and Space Administration, Cleveland, Ohio, October 25, 1973, 770-18. NASA-Langley, 1973 31 E-7679 BNWL-1900 APPENDIX 8.B Study of Extraterrestrial Disposal of Radioactive Wastes Part II NASA TECHNICAL NASA TM X-68147 MEMORANDUM i— i oo vO I X < on < STUDY OF EXTRATERRESTRIAL DISPOSAL OF RADIOACTIVE WASTES PART II by R. E. Hyland, M. L. Wohl, R. L. Thompson, and P. M. Finnegan Lewis Research Center Cleveland, Ohio October 1972 8.B.1 STUDY OF EXTRATERRESTRIAL DISPOSAL OF RADIOACTIVE WASTES Part II Preliminary Feasibility Screening Study of Extraterrestrial Disposal of Radioactive Wastes in Concentrations, Matrix Materials and Containers Designed for Storage on Earth by R. E, Hyland, M. L. Wohl, R. L, Thompson, and P. M. Finnegan Lewis Research Center ? 1. INTRODUCTION H AEC has initiated a program aimed at providing near-term and long- term solutions to the problems associated with the handling and manage- ment of radioactive wastes. Battelle- Northwest has been requested to conduct studies and to make evaluations of all currently envisioned long- term waste management methods. The objective of these efforts is to identify feasible long-term waste management systems and their components; identify the research and development necessary for their establishment; and estimate the sched- ule and costs associated with selected systems, In addition, these studies will be used as the basis for providing a discussion of alterna- tives in the statement of environmental impact required for authorization of a Federal waste repository project. The concepts to be studied include. (1) The application of alternate geologic storage techniques. (2) An international off-shore repository. (3) Storage in the seabed. (4) Use of the permanent ice caps. 8.B.2 (5) A ten-mile deep hole. (6) A deep cavity generated by a nuclear device. (7) Extraterrestrial transport. (8) Transmutation. (9) Other methods as yet unidentified, NASA has been requested by the AEC to study the feasibility of ex- traterrestrial transportation of radioactive wastes. More specifically, NASA has been requested to study the extraterrestrial transport of at least three types of radioactive waste materials: lo Radioactive wastes in concentrations, matrix materials, and containers currently designed for storage on Earth. 2. Actinide wastes with 0= 1 and 1 percent contamination by other radioactive wastes. 3. The third type to be defined later in the study. The general approach in each of these studies will be similar . The studies will be divided into -several phases. The first phase is a pre- liminary feasibility screening study. This phase will establish the maximum amount of the particular radioactive waste that could be trans- ported to space per launch. It will also establish the minimum cost of disposing of this particular waste in space and estimate the number of launches required per year. The effect of integration of the package with a vehicle and accident conditions will not be treated in this phase . The waste disposal container will be designed considering primarily normal operation. The primary emphasis will be on heat transfer, radiation shielding, and criticality. If this preliminary feasibility screening study indicates that the cost is reasonable, then a phase II feasibility study will be conducted which will include integration with the launch vehicle and consideration of all the safety aspects of the study. If the phase I study indicates that this maximum waste payload and minimum cost system may not be feasible, then the phase II study will be postponed until other more promising space disposal systems are considered, 8.B.3 If the method appears feasible after the phase II feasibility study, then the third phase of the study would be conducted. This phase would identify the research and development necessary to demonstrate feasi- bility and estimate the schedule and costs associated with the develop- ment and operation of the system. This report describes the results of the phase I preliminary feasi- bility screening study for the first class of radioactive wastes, that is, the radioactive wastes in concentrations matrix materials, and con- tainers currently designed for storage on Earth 2. DESCRIPTION OF WASTE PRODUCTS, MATRIX MATERIALS AND CONTAINERS 2 1 Description of Waste Products The radioactive wastes in this report are the fission products that remain after processing as indicated in references 1 and 2 iollowed by a ten- year hold in temporary storage. This processing separates the fission products from the unfissioned fuel, structural materials, and the actinide class of radioactive wastes. However, the iission products after processing still contain small amounts of these materials Their effects are considered negligible in this report Figure 2-1 shows how the activity (curies/gm) and the thermal power (watts/gm) of the fission products vary with time after discharge from the reactor. Both have decayed by about a factor of ten by the end of the ten- year hold in temporary storage. At that time about half the heat is generated by three isotopes, 90y, 137^-,, and its daughter 137 B Table 2-1 shows the characteristics of fission products from light- water reactors (LWR) and from liquid metal fast-breeder reactors (LMFBR)„ For the purposes of this study the important characteristics are the amount of fission products per MWe day and the thermal power and activity per gram of fission products. LMFBR waste products have less thermal power and activity per gram because the isotopic distribu- 8.B.4 tion of the fission products produced by fast and thermal reactors are somewhat different. In addition, the amount of fission products produced per MWe day is less for the LMFBR because the expected efficiency of the plant is expected to be higher, 40 percent compared to 33 percent for the LWR„ For reference, the LMFBR produces 2. 64 gm/MWe day com- pared to 3. 23 gm/MWe day for the LWR, 2,2 Description of Matrix Materials Four types of solid-matrix materials designed for storage on Earth are described in Table 2-2 (ref. 3). The four types are spray melt, pot calcine, phosphate glass, and fluid- bed calcine, The spray melt was selected for this study. It was selected because its activity (curies per unit volume) is high, yielding a more compact disposal package. In addition (1) it has a tough structure compared to crumbly, brittle, and granular structure of the other materials permitting it to remain intact in case of impact, (2) it has a high maximum stable temperature, 1170 K compared to 1170, 770, and 870 K for the other materials, allow- ing it to operate at higher temperatures thus permitting larger diameters and higher payloads, and (3) it has a comparatively high- thermal con- ductivity, 1. 8 watts/m(K) compared to 0. 5 and 1. 8 for the other mater- ials, permitting high heat removal rates or large diameters without exceeding center line temperatures. Spray melt appears to be the best of the four candidate matrixes presently available but it has several drawbacks from the space disposal viewpoint. It is desirable to have a material with a higher thermal con- ductivity, higher maximum stable temperature, and higher values of curies per unit volume. Any future studies should consider other matrix materials when they become available. 8.B.5 2, 3 Description of Fission Product Storage Containers The containers designed for Earth storage of fission products in the above matrix materials are cylinders. The cylinders are made of a non- reactive material which would be stainless steel for pot calcine and either stainless steel or mild steel for the spray melt, phosphate glass, and fluid bed calcine. Diameters of 6 to 24 inches and lengths of 8 to 10 feet are being considered (ref. 3). The diameter for Earth storage is selected primarily to assure the maximum centerline temperature of the matrix material is below the maximum stable temperature The centerline tem- perature will be higher for the larger diameters The lengths are selec- ted on the basis of requirements such as loading, handling, and mainte- nance, In the design of a cylindrical container for space disposal the diam- eters will also be selected to assure that the temperature of the matrix material does not exceed the maximum stable temperature, which is 1170 K for spray melt. The diameter of the vessels will, however, be permitted to exceed 24 inches if this is advantageous. The length of the cylinder will be determined by the maximum payload of the launch vehicle or the space tug, whichever is limiting. Table 3-1 shows the maximum allowable payload weights for the candidate combinations of payload, tug, and launch vehicle. The information in this table will be discussed in the next section, Figure 2-2 is a schematic drawing of the cylindrical containers con- figured for space disposal. The fission products are contained in the in- ner stainless steel cylinder. This cylinder is surrounded by a depleted uranium gamma shield, The cylinder with the gamma shield is then in- serted in an outer stainless steel radioactivity containment vessel. Both the inner and outer vessels are welded and helium leak checked for tight- ness. For the purposes of this study all materials are assumed to be in intimate contact so that the temperature drop across the interfaces can be neglected. This is an optimistic assumption and the effect will have to be checked if the concept warrants a phase II feasibility study. 8.B.6 The uranium shield was selected because it gave the lowest shield weight However, uranium changes phase (a to /3) when its tempera- ture goes above 930 K and the phase change causes the material to ex- pand prohibitively. The maximum normal operating temperature is about 600 K. The feasibility of using depleted uranium for the gamma shield would be reconsidered in the phase II study because abnormal and emergency conditions may cause the shield temperature to exceed the phase change temperature. 3, PAYLOADS, COSTS, AND DESTINATIONS FOR CANDIDATE SPACE VEHICLES This subject is discussed in detail in reference 4 "Space Trans- portation Considerations for Disposal of Radioactive Wastes" by J. Ramler, R, Thompson, and S. Stevenson, This section summarizes some of the pertinent information in this report and discusses the rea- sons for selecting the shuttle as the launch vehicle and Earth escape as the destination for this study. The candidate destinations starting with the highest AV require- ments and lowest payloads are: direct solar impact, direct solar es- cape, solar impact via Jupiter, solar escape via Jupiter, solar orbit, solar orbit via Venus, solar orbit via Mars, Earth escape and Earth orbit. The candidate expendable vehicles starting with the highest pay- load capability are: Saturn V/Centaur, Saturn V. and Titan III E/Cen- taur There is one shuttle design but two types of shuttle launches. One is a single shuttle launch which carries both the payload and the tug. The tug transports the payload from low Earth orbit to its final destination. The other requires two or more shuttle launches. One carries the payload. The other shuttle or shuttles carry one tug each. The payload and tug (or tugs) rendezvous and are assembled in low Earth orbit. The tugs may be either expendable or reusable. 8.B.7 Table 3-1 shows the payloads and costs for the candidate destina- tions and candidate launch vehicle and tug combinations This table shows that direct solar impact is not possible with today's vehicles. It also shows that the shuttle is the most economical launch system The shuttle was therefore selected as the launch vehicle. Earth escape was selected as the destination for this phase. High Earth orbit permits carrying about the same payload but was not selected because this destination was considered to be too near the Earth . This was an arbitrary decision and will be re-evaluated with other promising destinations in later phases of the study when other aspects, in addition to payload, are considered in the selection of the destination The trajectories used to calculate the above payloads were not dog- legged Dog- legged trajectories may be required to avoid potential impact on land after abort during ascent. Dog- legging the trajectory requires more fuel and, if the shuttle is fully loaded, the payload de- creases. When the trajectories are dog- legged, the payload for the single shuttle launch may be reduced more than that for the dual launch as follows: In the single launch case, the shuttle is fully loaded and thus the final payload decreases for missions with a dog- leg In the dual case the tug- carrying shuttle is fully loaded but the waste carrying shuttle is only partially loaded, 34 000 pounds, compared to a maximum allowable of 62 000 pounds. The payload is limited to 34 000 pounds be- cause the maximum the tug can put into Earth escape orbit is 31 000 pounds (the 3000 pound difference is due to structures that remain in the shuttle). Thus, 28 000 pounds of additional fuel could be carried for the dog- leg maneuver without reducing the payload, The second shuttle with the tug may not have to be dog- legged since land impact of the tug i s less h azardous. If the shuttle with tug is not dog- legged, the Earth escape is a solar orbit obtained by one burn from Earth or- bit, The important characteristic of this orbit is that it intersects the Earth's orbit and introduces the possibility of Earth impact. Solar or- bit refers to orbits about the Sun which are either inside or outside the Earth orbit with negligible probability of impacting the Earth, 8.B.8 maximum payload for the dual launch mode will not be much lower than 31 000 pounds. This is the main reason for looking at the dual shuttle launch mode of operation. Without such a potential advantage the single shuttle would be preferred because it is a less complicated launch mode. The cost comparison was made using existing expendable vehicle designs . If space disposal appears feasible, then the development of an expendable vehicle should also be considered. Cost savings may result due to mass production and potentially higher payload capacity. 4. ANALYSIS The phase I analysis has two main parts. First, determination of the maximum amount of fission products that could be carried in the shuttle-orbiter-tug vehicle to earth escape. Second, calculation of the launch cost per pound, per curie, per MWe day, and per kw-hr electric to establish the effect of launch cost on the electric generating cost. The number of launches required in 1985 and 2000 are also estimated, If the effect on electric cost makes the system potentially not feasible, then the phase II feasibility study would be postponed until phase I stud- ies on potentially more promising waste-matrix- container systems have been completed. If the results of the phase I study indicate this waste-matrix- container combination to be potentially feasible, then a phase II feasi- bility study would be conducted which would include design for in- shuttle cooling and for off-normal, emergency and accident conditions and would consider shuttle safety environment, abort, re-entry, impact, and heating after impact. The following sections describe the design criteria, procedure, and assumptions for the phase I analysis 4. 1 Design Criteria For this phase of the study three classes of criteria are considered: radiation dose levels, matrix material temperature limits, and shuttle 8.B.9 payload limits. In the double shuttle launch where the pay load and tug are placed in orbit on separate launches, the maximum payload is deter- mined not by the shuttle but by the tug. The tug limit is based on what payload it can take to a destination or put on a trajectory. Radiation dose levels. - Dose levels in three situations were con- sidered: Shuttle crew Shuttle instrument dose for unmanned shuttle After accident public exposure 2. 5 mrem/hr in the crew compart- ment which corresponds to 10 rem/hr at 3 meters from the center of the container 7 10 rad to the nearest instrumen- tation which corresponds to 500 rem/ hr at 3 meters from the center of the container 1 rem/hr at 3 meters from the center of the container Temperature limits . - Calculated spray melt matrix material tem- perature shall not exceed 1100 K which is 70 K below the maximum stable temperature for spray melt. Shuttle payload limits for Earth escape. - ingle shuttle launch Payload 17 000 lbs Structure on payload 500 Structure in shuttle 3 000 Tug 44 500 Total 65 000 'wo shuttle launch Shuttle 1 Tug 59 000 Structure in shuttle _3_000 Total 62 000 8.B.10 Shuttle 2 Payload 30 000 Structure on payload 1 000 Structure in shuttle 3 000 Total 34 000 lbs 4. 2 Procedure and Assumptions for Waste Container Design The steps for a phase I design of a container and shield which meets the normal operation dose, temperature, and payload criteria are listed below: 1. Select a set of container diameters in the range- proposed for Earth storage. Diameters of 6, 12, 18, 24, and 28 inches were selected. 2. Calculate the radiation and heat source for a selected activity concentration and container diameter. 3. Calculate gamma shield thickness for one of the dose criteria assuming uranium metal" as the shield material. 4. Calculate the surface temperature of the container in orbit assum- ing an emissivity of 0. 8. 5. Calculate the temperature in the center of the matrix material assuming no temperature drop across the material interfaces . 6. Calculate the weight per unit length of payload where payload is the fission product, matrix material, containment cylinder, and shield. 7. Calculate the weight of the end shields and containment vessel end caps and subtract from the allowable payload to get the weight of the center section of the payload. 8. Calculate the length of the cylinder which makes the payload weight equal the payload criteria. 9. Calculate the weight of fission products per launch. 10. Perform above calculations for three dose rate constraints, three activity concentrations and for single and double shuttle launches. 8.B.11 4. 3 Procedures and Assumptions for Calculation of Space Transport Cost The cost of space disposal of radioactive wastes can be divided into several categories. (1) Temporary storage on Earth. (2) Separation, concentration, and preparation of wastes in matrix materials, (3) Design and fabrication of the space disposal container system and assembly of wastes and matrix material into the container. (4) Shipment of wastes to the launch site. (5) Space transportation cost. This report is concerned with only one of these costs - the space transportation cost. The space transportation costs begin when the pay- load is delivered to the launch site. The major space transportation costs end when the payload gets to its destination. The costs, however, may not go to zero at this time. There may be additional monitoring costs depending on the disposal destination. The other (all of the above) costs will have to be determined before a complete economic analysis can be made, The purpose of the present analysis is to determine the relation of the space transportation cost to the cost of generating electricity. Specifically, the space transporta- tion cost will be compared with the bus-bar cost of electricity which is assumed to be 8 mills/kw-hr. The factors that affect the space transport cost to the electric con- sumer are: (1) Launch cost including shuttle and tug, (2) Destination and gross payload. (3) Ratio of radioactive waste to gross payload. (4) Interest on funds collected and set aside for space disposal at the end of Earth storage time. (5) Radioactive waste Earth storage time. 8.B.12 The steps for a phase I economic analysis are: lo Determination of the gross payload for the candidate destina- tion which is Earth escape in this study. Gross payload is defined as the weight of the waste container system delivered to the destination and includes all structures and auxiliary systems fixed to the con- tainer 2o Determination of the net waste container payload by substract- ing the weights of the structures and auxiliary systems fixed to the waste container from gross payload. 3„ Determination of the amount of fission products that can be car- ried in a container whose weight including the shielding equals the net payload. 4. Determination of the launch cost including shuttle and tug, 5. Determination of the cost per curie of fission products trans- ported to space. 6 Determination of the discounted space transportation cost per curie disposed, that is, determine the amount of money per curie that could have been put in a trust fund for space transportation. This as- sumes that the consumer was charged for space transportation when he used the electricity and that the money was put in a trust fund and com- pounded at current interest rates. 7. Determine the amount of electricity (MWe days) that was gener- ated per curie disposed. 8. Determine the cost of space transportation of wastes in units of mils per kw-hr of electricity. 9. Compare the cost of space transportation from step 8 (mills/ kw-hr) to the bus-bar generating cost of 8 mills per kw-hr. 5. RESULTS AND DISCUSSION This section has four main parts. Part one establishes the net payload and launch cost as a function of the destination and launch ve- hicle. The next part describes the design of the waste payload package 8.B.13 and discusses the effect of the design parameter on the dimensions of the container, the container temperature and thermal power, and the amount of fission products per launch. The third part describes the launch costs in mills per kilowatt hour of electricity and discusses the effect of container design parameters, earth storage time, and interest rates on the mill/kw-hr cost. The fourth part estimates the required number of shuttle flights per year to 2010 AD and discusses the effect of the destination and the design parameters on the number of launches - 5. 1 Gross Payload and Launch Cost For this phase I feasibility screening study Earth escape was selected as the disposal destination and the shuttle was selected as the launch vehicle. Gross payload is defined as the waste container plus the structure attached to it. At this time single and dual shuttle operations appear to have equal feasibility and are both considered In single shuttle operation both the waste payload and tug are carried in the same shuttle. In dual shuttle operation one shuttle carries the waste payload and the other carries the tug. The following table summarizes the pay- load and cost data for Earth escape. Vehicle Single shuttle Dual shuttle launch Gross payload, wt. (lb) Launch cost, $ Cost per pound of gross payload /lb 17 500 31 000 11 M 21 M 628 677 Payload and Cost Data for Earth Escape The costs per pound of gross payload are $628 and $677 and are es- sentially the same within the accuracy of this study. The selection of single shuttle or dual shuttle operation will be made at a later time and will depend on additional considerations, for example, ratios of waste 8.B.14 weight to gross payload weight (this tends to increase with increasing gross payload), effect of dog- legging the trajectory (this tends to re- duce the single shuttle payload more), the complexity of the dual oper- ation compared to single shuttle operation and safety considerations. For reference, Table 3-1 shows the payloads and costs for the other candidate launch vehicles and destinations, The candidate des- tinations starting with the highest AV requirement (lowest payload) are: direct solar impact, direct solar escape, solar impact via Jupiter, solar escape via Jupiter, solar orbit, solar orbit via Venus, solar orbit via Mars, Earth escape, and Earth orbit. The candidate expendable vehicles starting with the highest payload capability are: Saturn V/ Centaur, Saturn V, and Titan III E/Centaur. When selecting the candidate destination and launch vehicle, other factors besides payload and launch cost must be considered. For ex- ample, the Jupiter, Venus, and Mars fly-by missions require less AV but more accurate instrumentation and control than the more direct mis- sions. In addition, Jupiter, Venus, and Mars are in the proper positions for launch for only about one month every 12, 19, and 25 months, re- spectively, Thus the fly-by missions would require all the launches for the 12 to 25 month period be made in about one month This would re- quire several launches per day during about a 1/2 hour launch window,, More detailed discussion of these aspects can be found in reference 4, Earth escape, Earth orbit, and solar orbit result in the highest payloads per vehicle but each has drawbacks that must be investigated,, In the case of Earth escape the possibility of re-encounter with the Earth at some future time must be made negligibly small for thousands of years due to the long-life of the waste materials. In the case of Earth orbit the possibility of interference with other space activities must be studied and made acceptable. Solar orbit reduces these problems but requires additional burns later in the mission as does Earth orbit. Solar system escape would eliminate these problems but would be costly. These types of considerations are discussed by Ramler, Thompson, and Stevenson in reference 4. More detailed analysis of this type, integrated 8.B.15 with safety and economic analysis is required before a destination can be firmly selected. 5.2 Waste Payload Design The physical features of the waste, matrix materials, and containers are described in Section 2, Table 2-2, and figures 2-1 and 2-2. The design procedures and assumptions for the container and shield were described in sections 4. 1 and 4. 2. The results of the parametric anal- ysis of these designs are presented in figures 5-1 through 5-8 and are discussed below. There are three main categories of design criteria: (1) Radiation dose rates of 1, 10, and 500 rem/hr at three meters from the container center line. (2) Centerline temperature less than 1170 K which is the maximum stable temperature for the spray melt matrix. (3) Gross payload weight: 17 500 pounds for single shuttle and 31 000 pounds for dual shuttle operation. Additional independent parameters are: (1) Radioactivity concentration in the matrix to a maximum of 10 curies/cCo (2) Earth storage time. The dependent parameters are: (1) Diameter of the waste plus matrix material. (2) Diameter of the outer containment vessel, (3) Length of the container. (4) Payload thermal power. (5) Containment system outer surface temperature. (6) Amount of radioactive waste in the container. Diameter of waste matrix . - This diameter was determined the requirement that the matrix centerline temperature not exceed the max- imum stable temperature of the spray melt matrix material which is 1170 K. Figure 5- 1(a) and (b) show the design point diameter for the 8.B.16 matrix material to be about 28 inches for an activity concentration of 10 curies/cc and a radiation dose of 10 rem/hr and 500 rem/hr. The matrix diameter is the same for both dose rates due to two compensating effects. First, the lower dose rate requires a thicker shield and, for the same matrix material diameter, the temperature drop through the thicker shield is higher. Second, the thicker shield makes the outer container diameter larger and this reduces the surface heat flux and the surface temperature. These two effects essentially cancel each other and the 28 inch matrix material diameter satisfies the maximum centerline tem- perature requirement independent of dose in the range considered, which was 1 rem/hr to 500 rem/hr . Diameter of outer cont ainment vessel . -This diameter was deter- mined by the gamma shield thickness for the three dose rates considered and a 1 inch thick impact shield as outer shell. The shield material was depleted uranium and its normal operating temperature was about 600 K. The uranium shield thickness, for the design dose rates of 1, 10, and 500 rem/hr were 4.5, 3. 5, and 2 inches, respectively. These thick- nesses were determined by the comparison method, using data from ref- erence 5 For all doses the matrix material diameter was 28 inches and the inner containment vessel thickness was 1/2 inch. Figure 5-2 shows the outer diameter as a function of dose rate. The outer diameter is 40, 38, and 35 inches for dose rates of 1, 10, and 500 rem/hr, o Length of c ontainer. - Figure 5-3 shows the length of the container as a function of the dose rate and for single and dual shuttle payloads of 17 000 and 30 000, respectively. The lengths for a 17 000 pound payload and for dose rates of 1, 10, and 500 rem/hr are 2.59, 3,33, and 5,21 feet, respectively. The length for a 30 000 pound payload and for dose rates of 1, 10, and 500, rem/hrs are 5.09, 6.35, and 9.63 feet, respectively. Thermal power, - Since the matrix material diameter is a constant and independent of the dose rate, the thermal power generated per foot 2 The length includes the active matrix plus the end shielding and container thicknesses. 8.B.17 of cylinder is also a constant of 6 kw/ft. The length of the container is determined by the allowable payload and the radiation dose as dis- cussed in the previous section. Figure 5-4 shows the thermal power as a function of dose rate for single and double shuttle launches , The thermal power for a single shuttle payload of 17 000 pounds and dose rates of 1, 10, and 500 rem/hr are 9,6, 15.1, and 27. 8 kilowatts, re- spectively. The thermal powers for a dual shuttle payload of 30 000 pounds and for the same dose rates are 24,6, 33,2, and 54.3 kilowatts. Containm ent vessel outer surface tem pe ratur e. - The power per foot of capsule is a constant as discussed in previous sections. The surface temperature is then a function of the dose rate which defines the container diameter and radiating area per foot. Figure 5-5 shows the surface temperature as a function of dose rate. The surface tem- perature for dose rates of 1, 10, and 500 rem/hr, 3 meters from the container centerline was 480, 500, and 525 K, respectively. Amount of radioactive waste in the container . - The weight and amount of radioactivity per foot of capsule is the same for all capsules because the activity concentration is constant at 10 curies/cc and the waste in matrix diameter is constant at 28 inches. The amount of fis- sion products per container is a function of the maximum allowable pay- load and the dose criteria. Figures 5-6 show the amount of fission products in the container in curies and the packaging weight ratio for single and dual shuttle launches and for radiation dose rates from 1 to 500 rem/hr at 3 meters from the container centerline. The number of Megacuries for dose rates of 1, 10, and 500 rem/hr at 3 meters are 1,91, 3.02, and 5,56 for single shuttle payloads of 17 000 pounds. The number of Megacuries for a dual shuttle payload of 30 000 pounds for the same dose rates are 4.90, 6.62, and 10,9. 5.3 Space Transportation Cost The factors that affect the space transportation cost and the proce- dures and assumptions for calculating that cost are described in 8.B.18 Section 4. 3 The purpose of the analysis is to estimate the space transportation cost to the electric power consumer and to compare this cost to the electric cost, which is 8 mills per kw-hr at the bus-bar and 24 mi'lls per kw-hr average to the residential consumer. The effect on the transportation cost of each of the main parameters in the cost analysis will be determined. The parameters, the baseline values for the parameters, and range of variation of the parameters are listed in table 5-1. The effect of a parameter will be determined by varying the parameter, keeping the other parameters fixed at the baseline value. The results are presented in table 5-2 and the effect of the following parameters on the space transportation cost are dis- cussed below: (1) Destination (2) Dose rate (3) Earth storage time (4) Space disposal fund interest rate (5) Activity concentration Effect of destination on cost . - The gross payload and the cost per pound of gross payload are presented in table 3-1 for the candidate des- tinations and the required vehicles. Gross payload is defined as the waste container and the structure attached permanently to it. The ratio of radioactive waste to gross payload depends on the character of the waste and the design of the container. The cost per pound of gross pay- load for the candidate destinations is listed below: Earth escape, $/lb of gross payload . High Earth orbit ...... Solar orbit via Mars or Venus Solar orbit ...... Solar escape via Jupiter Solar impact via Jupiter Direct solar escape . . Direct solar impact . . . 628 . 628 . 794 . 800 3500 4700 4420 Payload is zero with existing vehicles 8.B.19 The effect of the destination on space transportation cost in terms of mills per kW hr electric is presented in table 5-2 for several destina- tions. The cost for Earth escape, solar orbit, and solar escape is 4, 5, and 28 mills/kw-hr, Effect of d ose rate, - The effect of varying the dose rate from 1 rem/hr to 500 rem/hr was determined for the Earth escape destina- tion and is shown in figure 5-7, The cost per pound of waste delivered for the 10 rem/hr dose is 65 percent of the 1 rem/hr cost. The cost for the 500 rem/hr dose is 33 percent of the 1 rem/hr cost. Effect of Eart h storage time . - The effect of Earth storage times of 10, 20, and 40 years was determined and is shown in figure 5-8, The cost to the electric customer goes down as the storage time increases primarily due to the increased interest accumulated on the disposal fund. The cost of storing the material is small compared to the interest on the fund and is neglected in this analysis. The activity of the waste de- creases with time which also tends to reduce the cost of disposal (less shielding, more waste payload). However, unless the nonradioactive decayed materials are removed from the waste to keep the curies per cc near the original level, the shield weight savings will be small. Time affects the cost in another and more significant way. Interest on the funds set aside for waste disposal increase much faster than the fission products decay. The time for 10 year old fission products to decay by half is about 30 years. The money doubling time is about 10 years at a seven percent interest rate. The effect on cost due to fission product decay during storage is neglected. Only the effect of interest on money in the disposal fund is considered. The cost per pound of waste delivered for the 20 year storage case was about half the 10 year cost and the 40 year storage time was about 1/8 of the 10 year cost. At a 7 percent in- terest rate the charge to the customer is reduced by half for each ten year storage time. Therefore, a storage time can be found for each destination which will make the initial charge to the electric consumer acceptable, 8.B.20 Effect of space disposal interest rate . - The interest on the space disposal fund can affect either the space transportation cost or the re- quired storage time. The effect of interest rates of 5, 7, and 10 per- cent were determined and are shown in figure 5-8, The cost at a 7 per- cent interest rate for materials stored ten years in about 1. 3 times that at 10 percent and about 80 percent of that at 5 percent. To get the same transportation cost at 5 and 10 percent as for 7 percent with a storage time of 10 years requires a storage time of about 14 years at 5 percent and 7 years at 10 percent. Effect of activity concentration . - Increasing the concentration of the radioactive waste has a strong effect on the transportation cost but the amount the concentration can be increased is limited. Doubling the concentration would reduce the cost by about 45 percent. In general, the gain is not this great because the diameter must be reduced to keep the center line temperature below the maximum stable temperature. The base case in this study had a concentration of 10 curies per cc and the full density fission products have a density about 26 curies/cc after 10 year storage. Thus going to full density fission products would re- duce the costs by possibly more than half. The costs could be further reduced by removing the gamma emitters and/or the high thermal energy emitters and launching the harmful wastes that remain. The cost of launching separated radioactive wastes will be considered in later re- ports, 5. 4 Number of Shuttle Launches per Year The production rate of ten- year old fission products as a function of years from 1970 to 2000 is shown in figure 5-9. The amount of fission products that can be carried per launch is a function of (1) Destination (2) Dose rate (3) Fission product concentration 8.B.21 Figure 5-10 shows the number of launches per year to 2010 AD for Earth escape, single or double launch mode, Earth storage for ten years, and three dose levels. In 1985 the required number of launches for dose levels of 1, 10, and 500 rem/hr at 3 meters from the package center are 300, 210, and 115, respectively. The effect of the other parameters on the number of launches per year is shown in table 5-2. The payload to Earth orbit is about equal to the payload to Earth escape and the number of launches is also similar. The payload to solar escape is about 10 percent of the payload to Earth escape and the number of launches would be increased accordingly. Increasing the Earth storage time to 40 years decreases the total activity by 50 percent. The number of launches is decreased accordingly if the fission product concentration is assumed to be maintained at 10 curies/cc thus keeping the curies per launch constant (i.e. , more grams of fission products for the same dose level with longer storage time). An additional decrease in launch fre- quency could be obtained by removing the decayed isotopes after a long storage time. It appears that launching of all fission products at an early time period results in a higher cost and high launch frequency. Both can be avoided by holding for a longer time followed by separation. 6. CONCLUSIONS For this report all of the fission products (Le. , no separation) were considered for space disposal after being stored in Earth- storage facilities for 10 years. The fission products were assumed to be mixed in a solidified matrix material and contained in cylinders. These cylin- ders were sized based on the temperature limits on the matrix material and shielded to reduce the radiation dose rate to levels ranging from 1 to 500 rem/hr at 3 meters from the center of the package. In this report the impact of accidents on safety was not considered, and thus the con- clusions obtained pertain only to the package as designed for normal operations. This implies minimum cost and maximum quantity of fission 8.B.22 products per payloacL The payloads were based on results of a pre- vious study (ref. 4) in which the shuttle was selected as the lowest- cost vehicle- The destinations chosen for the report for comparison were Earth escape, Solar orbit, and Solar escape. The following conclusions were obtained from the results presented. 1. Matrix material such as spray melt can be used without exceeding temperature limits on matrix, but materials with higher thermal con- ductivity would be more desirable. Diameters of 28 inches or less were acceptable but not optimum based on fission products per package or cost. 2o The cost in terms of mills per kw hr electric, of space disposal of fission products (after 10 year temporary storage) in matrix mater- ials and containers currently designed for Earth disposal and shielded (1 rem/hr) is 4, 5, and 28 mills per kw hr for Earth escape, solar or- bit, and solar escape, respectively. This compares to 8 mills per kw hr bus-bar cost and about 24 mills per kw hr average consumer cost. 3, A major factor effecting cost was the Earth storage time. As- suming 7 percent interest on the funds set aside for space disposal, the cost to the electric consumer of space disposal is reduced by a factor of 2 for each 10 years of storage time. If the fission products are stored for 40 years prior to launch then the cost to the electric consumer are 0.5, 0.6, and 3,5 for Earth escape, solar orbit, and solar escape, re- spectively. There is, therefore, for each destination, a storage time that will make the initial charge to the electric consumer acceptable. Based on a normal operating condition design for solar escape, a stor- age time of more than sixty years is required to make the space disposal charge less than 10 percent of the bus-bar electric cost. 4o Large changes in dose rate are required to significantly affect the cost. Increasing the dose at 3 meters from the center of the package from 1 to 500 rem/hr results in factor of 3 reduction in cost. 5. The number of shuttle launches would exceed a launch per day within 5 years after the program was initiated if the material was launched as prepared for Earth storage and held for 10 years without further processing., 8.B.23 Inasmuch as fission products will decay with time, both the space transportation cost and number of launches can be reduced considerably by increasing hold time. Large reductions in launch costs might pos- sibly be achieved if the fission products are separated and only, say, the actinides are launched into space. The actinides, in particular, present a special hazard if they are permanently stored on the Earth because they have such very long half- lives. The extent to which this principal hazard can be reduced at low launch cost warrants further study. REFERENCES 1. Anon, : Siting of Fuel Reprocessing Plants and Waste Management Facilities. Rep. ORNL-4451, Oak Ridge National Lab. , July 1970. 2. Pittman, Frank K. : Plan for the Management of AEC-Generated Radioactive Wastes. Rep. WASH-1202, USAEC, Jan. 1972. 3. Schneider, K. J.: Solidification and Disposal of High- Level Radio- active Wastes in the United States. Reactor Tech. , vol. 13, no. 4, Winter 1970-71, pp. 387-415. 4. Ramler, J„ ; Thompson, R. L. ; and Stevenson, S. : Study of Extra- terrestrial Disposal of Radioactive Wastes. Part T. Proposed NASA Technical Memorandum. 5. Jaeger, R. G. , ed. : Engineering Compendium on Radiation Shield- ing. Vol. III. Springer- Verlag, 1970, pp. 21-23. 6. Anon.: Potential Nuclear Power Growth Patterns. Rep. 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Shielded with depleted uranium for lOrera/hr at 3 meters from the center of the cylinder. 1400 1200 1000 Centerline temperature 800 600 400 200 Maximum St able Te mperature ( P p t calc^n Radioactivity of matrix curies / cm 5 6 10 14 18 22 Waste Container Diameter, inches 26 30 8.B.32 Fig. 5-lb Temperature at Center of Cylindrical Containers Containing Fission Product Waste Material Stored for 10 Years. Shielded with depleted uranium for 500rera/hr at 3 meters from the center of the cylinder. 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For our purposes, the yield of particles for the amount of vapor which has to be processed is low. 6.6 HIGH-ENERGY DISCHARGE METHODS 6.6.1 Arcs Since in a high-intensity arc the major fraction of the arc input energy is dissipated at the anode, it is possible to incorporate material to be subdivided into the anode, either by itself or with a small amount of conductive material added. The material then issues as part of the (29) anode tail-flame at temperatures approaching 7,000°K. The resultant high rate of cooling causes condensation to occur, with the production of very fine particles. By varying the rate of cooling of the condensing vapors, particle size can be controlled. Particles from 0.005 to 0.2 um have been produced by this technique. A number of variations employing 8.D.19 the basic concept are described in the literature. One such technique is discussed in Reference 30. 6.6.2 Plasma Torches In many instances, the plasma torch is not used as a method for pro- ducing small particles; rather, it is used for the spheroidization of a (31) material available in powder form. In this method of operation, materials, often refractory, which are close to the desired shape are put in a high-temperature plasma. As the particles melt, surface ten- sion produces spheres of the same volume as the original particles. The molten droplets are then usually sprayed into water, where they harden. (32) (33) In other applications, spheres are made from wire or metal rod. Particles less than 20 um in diameter can be made. 6.6.3 Exploding Wires In this method, a wire of the metal required in powder form is ex- (34) ploded instantaneously by discharging high current through it. Par- ticles of the metal or its oxide, in the case of explosion in the atmos- phere, are formed in the 0.02-ura-diameter range. The method is a useful one from the point of view of many research needs, but the noncontinuous nature of its operation makes it an unlikely choice for any kind of pro- duction application. 6.7 THERMAL DECOMPOSITION Decomposition methods have been used to prepare metal oxide particles by spraying solution droplets of the metal salt into an oxidizing flame, which changes the salt into the oxide. Particle size is controlled by the size and salt concentration of the droplets and by the temperature and velocity of the flame. Studies on the thermal oxidation of plutonium have (35) indicated formation of an oxide aerosol in the size range 0.1 to 10 um. 6.8 KUCLEATION METHOD In this method, the vapor from a suitable source is mixed with the vapor of the material for which it is desired to form the aerosol. Atoms of the former substance serve as nucleation centers and the aerosol results. 8.D.20 The mixture is then diluted with air to minimize coagulation. The LaMer (Of. 'i'J) aerosol generator employs this principle. * 6.9 CONVENTIONAL LIQUID ATOMIZERS Under this heading are included all methods in which tiny droplets result when a film or sheet of liquid encounters an appreciable velocity difference between itself and a new environment. The shearing forces produced cause the film to collapse into filaments and droplets that travel away from the sheet's main body. The velocity difference can be produced by several means: A. Centrifugal spray nozzles. In this method, pressure energy Is converted into a high-velocity swirling film, and a liquid mist ( 38} results. Extensive studies have been conducted on this method. B. Rotating disc atomizers. There are a number of variations on (39 40") this method. v * ' The principle is to emit the liquid film from the edge of a spinning disc or bowl and thus create the condition of a large velocity difference. The droplets formed using the above methods are typically in the 100- pm size range. In addition, energy utilization efficiency is rarely more than a few percent, and uniform droplet size cannot be attained without an auxiliary method for removing the over- and undersize. Theoretical papers are available on the general topic of drop formation from rapidly moving liquid sheets. 6.10 TWO-FLUID OR PNEUMATIC NOZZLE ATOMIZERS In this type of device, a slow-moving liquid stream is injected into a gas stream that is moving at near-sonic velocity. Examples of studies conducted using this method are found in References 42 and 43. The two- fluid method can produce droplets typically in the 100-ura size range, but the power consumed per unit mass of material atomized Is quite high. 6.11 ULTRASONIC ATOMIZER A high-frequency (about 20 kHz) transducer provides ultrasonic vibra- tions that are focused at the surface of the liquid from below. A liquid spout (called an ultrasonic fountain) then arises, and a fog is emitted 8.D.21 from its base. Droplets are formed with diameters under 5 ym. More details can be found in Reference 44. 6.12 UNSTABLE LIQUID JET Under the action of surface tension forces, a free column of liquid will spontaneously disintegrate into droplets. All the droplets will be of about the same size, with a diameter of about twice the jet diameter. It is assumed that the surrounding gas in such a case does not influence the disintegration. The droplets formed are not exactly of the same size and have not exactly the same speed; thus, they collide and coalesce. This coalescence can be considerably delayed by making the size, spacing, and speed of the droplets more uniform through velocity modulation of the jet. (45) The theoretical basis for the method is the work of Raleigh. According to his analysis, a cylindrical jet of inviscid liquid is un- stable with respect to an axially symmetric perturbance which has a wave- length X greater than tt times the unperturbed jet diameter. Indeed, the optimum wavelength to diameter radio (X/D) for droplet formation is 4.5. In such a case, one droplet per wavelength results. The size and frequency with which the droplets are produced depend upon the flow rate of the liquid through the capillary, the capillary diameter, its resonant fre- quency, and the amplitude of the oscillation of the capillary tip. The method is most suitable for producing beams of droplets of near uniform size at a relatively high production rate. Various schemes have been employed to obtain instabilities in a (46 47) liquid jet. Typically, some means of electromagnetic disturbance ' have been employed as well as piezoelectric transducers. ' Strom has used a membrane, vibrated by means of electrostrictive elements. In addition, modulation via the pressure of the liquid inside the capillary has been employed by a number of workers. ' A variation on this method is presently being used at Gulf Energy & Environmental Systems Company (GEES) to fabricate fuel particles for use in the high-temperature gas-cooled reactors. 8.D.22 6.13 ELECTROSTATIC ATOMIZER Such a device produces droplets when an electric stress produced by the applied electric field exceeds the liquid's surface tension force. The electric stresses draw the liquid to a filament and disperse it into charged droplets. By itself, a column of liquid issuing from a capillary will disinte- grate into droplets under the action of surface tension forces. For a vertical capillary, for example, the droplet at the exit of the capillary will grow In size until its weight overcomes the restraining forces of surface tension along the edge of the capillary tip. At this point, the drop falls from the capillary, and the process repeats itself. When an electric pocential Is applied between the capillary and a ground plate cen- tered beneath the capillary, the liquid becomes charged and strong downward forces on the droplet result. The effect of the applied voltage will be to reduce the size of the falling drops and to increase their frequency of formation. The size will continue to decrease with increasing voltage until the diameter of the falling "drops is roughly twice the diameter of the cap- illary. At this point, the meniscus at the capillary tip will become elec- trolydrodynamically unstable, and haromnic electrical spraying of liquid drops from the meniscus will commence. The transition from the dripping mode to the spraying mode will be marked by a sharp decrease in the size of meniscus and the diameter of the emitted drops, and a sharp increase in the frequency of droplet emission. References 53 to 56 are descriptions of work based on the use of the principle. Specific use of this method is made in the paint industry where paint droplets are electrically deposited on the object to be painted. More uniform coverage and less overspray results. 7. CHOICE OF A METHOD TO PRODUCE SMALL PARTICLES FOR WASTE EJECTION The criteria for choice of a method for producing small particles of radioactive material for waste ejection are: 1. A method in which the waste material (radioactive waste and matrix material) is transferred as a solid mass into orbit, as opposed to 8.D.23 a method in which a powdered material is launched into orbit. The safety reasons are obvious. 2. A method in which the particles are formed in a continuous beam ready for acceleration purposes. This eliminates the mechanical problems associated with making the particles and then transferring them in powder form to the charging and acceleration stages. 3. A method capable of an ejection rate which is commensurate with the projected radioactive production rate of 100 tons/year from 100 power plants of 1000-KW capacity each. The dilution caused by the use of a matrix must also be considered. A. A method in which nearly all the same size particles are produced. This would eliminate the need for reprocessing and the accelerator design would consequently be simpler. 5. A method in which the particles produced would be of the smallest size commensurate with steps 1, 2, 3, and A. The smaller the par- ticles, the lower the acceleration voltage required for ejection and, thus, the less demand on accelerator engineering. In addition, the device used to produce the particles should be the least complex possible from the point of view of telemetry needs, the num- ber of moving parts kept to a minimum because of lifetime problems, and the weight kept as low as possible to save on launch costs. By a process of elimination, only two of the methods listed in Section 6 appear to meet the above requirements. They are the electrostatic atom- izer and unstable liquid jet methods. Our system concept is based on the use of either of these methods or a combination of both. 8. SYSTEM CONCEPT The envisioned system consists of two main sections, the accelerator section and the waste container-particle generation section. The accelerator section would be placed in orbit and would not be recoverable. Its main function would be to supply the acceleration poten- tial required to eject the material. In addition to the necessary elec- trodes and collimators, it would have a means of generating the high voltages The emphasis should be on simplicity and reliability. 8.D.24 The waste container, besides holding and shielding the radioactive waste, would also possess the required equipment for generating the par- ticles. In this manner, the most complex, .and thus less reliable, portion of the entire system could be checked out on the ground before launch and, in the event of failure in orbit, it could still be recovered. Consequently, there would be no need to engineer a system possessing extremely high reli- ability and long lifetime, and the cost of the system would be reduced con- siderably. This portion of the system would dock with the orbiting accel- erator section. 8.1 WASTE PROCESSING If the electrostatic atoraization and unstable liquid jet methods are to be used, the waste must be in the liquid state; yet for safety reasons, the waste material must be solid during the launch. Consequently, some means of melting the material and containing it once in orbit has to be devised. One possibility would be to take advantage of the internal heat generated by the radioactive waste itself. The waste container in such a case could possibly be made of small enough diameter so that the ratio of surface area to volume was small enough to allow the entire mass within the vessel to melt. External cooling fins could then be attached to in- crease the radiative heat transfer and consequently keep the mass solid on the ground and during launch. Once in orbit and linked up with the accel- erator section, the fins could be retracted by some mechanical means and the mass permitted to melt. The use of a solar concentrator to supply the necessary heating should also be considered. Such a unit should be integral with the accelerator and remain in orbit. In addition, it may be possible to use part of the power from the system on board the accelerator section to supply the required heat energy. 8.2 ACCELERATOR The accelerator design itself has yet to be looked at in detail. From the discussions in Section 4, it would appear that for a 1-um-diameter par- ticle of density 3 g/cc, it might be necessary to generate an acceleration potential as high as 20 MeV. Achieving these potential levels is not a limitation in itself. The limitation appears rather to come from the prac- ticle problems associated with size and weight. From this point of view, 8.D.25 it behooves one to explore all the possible ways of reducing the required voltage level. This can be done, effectively, by reducing the particle size and/or increasing the charge-to-mass ratio. One of the most important parameters determining particle size in either of the above methods is capillary size. It is possible to purchase, (57) nowadays, collimated hole structure ' fabricated from stainless steel which possesses line-of-sight capillaries down to 1-um diameter. In addition, porous plugs of sintered metals of extremely small pore size are available. Various methods for increasing the charge-to-mass ratio of particles have already been discussed In Section 6. One should compare the cost associated with using these suggested methods to that associated with simply building the accelerator larger. 8.3 PARTICLE CHARGING The most convenient method of particle charging would be that of employ- ing electric field charging as in the electrostatic atomizer (Section 7) . The fact that the high field strengths are already available for acceler- ation purposes makes this a desirable charging method. In the event that the above charging method cannot be employed, the particles could be charged by electrode contact or by a crossed ion or elec- tron beam before entering the acceleration region. The additional equipment required, however, is relatively complex and would possess a short operating lifetime. For these reasons, it would be best made a part of the recover- able waste container package. 8.4 POWER CONSIDERATIONS The production rate of radioactive waste has been estimated at 3 g/sec, and if we assume the ratio of matrix material to actual radioactive waste material to be 1:1, the waste ejection rate from the accelerator is 6 g/sec. Further, the mass, m, per particle assuming spheres is 2 m «= A/3 nr p Thus, for p = 3 g/cc and r = 0.5 urn, i.e., 5 x 10 cm m c 1.6 x 10 g 12 Thus, a waste ejection rate of 6 g/sec is equivalent to 3.8 x 10 " particles/sec 8.D.26 The charge on each particle, q, is given by q » 4 -nr e E where, if we take E = 2.5 x 10 V/m, we obtain r( ,« X0 1 " coulombs This is a current of 0.25 amperes. Thus, the power required to maintain a potential difference of 20 million volts with this charged particle ejection rate is 5 MW. In addition, it will be necessary to eject charge of the opposite sign to maintain charge neutrality of the satellite. This does not involve the use of voltages greater than a few kilovolts, however, so the power needed for this function is in the kilowatt range. In such a case, however, the electrons would best be expected to be trapped In the magnetic field of the earth, since otherwise they will attempt to return to the satellite and not alleviate the electrostatic charging problem. 8.5 PARTICLE SOURCE CONSIDERATIONS 12 In Section 8.4, the particle ejection rate was found to be 3.8 x 10 particles/sec, assuming 1-um particles of density 3 g/cc. In addition, typical drop frequencies have been found to be, for a good yield of particles of the correct size, about 1500 particles/sec in the GEES drop-forming device for fuel particle production. Thus, from these data it would appear that some 2.5 billion orifices would be required. Collimated hole structures are typically sold off-the-shelf with 750,000 holes of 1 urn diameter within an 2 (57) area of 0.0014 ft . Using these values, therefore, the capillary area which would be required to handle the above waste ejection rate would be 2 some 5 ft , which is a circular cross section of 2.5-ft diameter. This area may have to be increased to provide more space between the capillaries, but even so, it is not unreasonable to envision the use of collimated hole structures to obtain the required number of orifices. In addition, it is likely that higher droplet formation rates can be obtained through research and development efforts. 8.D.27 Concerning the handling of molten oxides, this does not appear to be a particular problem. Small spherical particles containing the oxides and carbides of zirconium, thorium, and uranium are typical made in the GEES apparatus. In addition, Reference 51 describes a glass-drop-forming device. Collimated hole structures are presently manufactured from stainless steel but the potential fabrication of structures from refractory metals is indicated. (57) 9. CONCLUSIONS This precursory study suggests the feasibility of using an electro- static particle accelerator for the ejection of radioactive waste material from earth orbit into outer space. The major findings of this study are summarized as follows. The main advantage appears to be in the saving in energy, and conse- quently cost, which results compared to using chemical rockets. Using a projected waste production rate of 100 tons/year from 100 power plants of 1000-MW capacity, experimental values for the electric field strength, and an assumed particle size of 1 um diameter, the calculated power outlay is about 5 MW to escape the solar system. Of the various ejection schemes considered, the one which appears best from the point of view of simplicity of design of accelerator, least cost, and high reliability is that involving ejection frcra earth orbit out of the ecliptic plane. An order-. (± of-magnitude calculation indicates that the expected radioactive contamina- tion of celestial bodies would be negligible. Accelerator design requirements were obtained by seeking pertinent information on particle charging and acceleration reported in the litera- ture. Such Information is available from studies which are attempting to simulate micrometeoritic impact by the acceleration of ym-size particles to hypervelocities. From the experimental values of electric field strength available, the potentials necessary for acceleration of various size par- ticles are calculated. Indications are that, for an assumed density of waste material (radioactive plus matrix material) of 3 g/cc, a potential of 8.D.28 20 million volts would be required to accelerate a 1-um-diameter particle to solar escape speed. Although not a limitation in itself, from a practical point of view it would be best to reduce this potential to a lower value. A number of suggestions are provided in the text. An extensive literature search has been made seeking methods of pro- ducing small particles. Of all the methods surveyed, two in particular appear to be appropriate for this problem. They are the electrostatic atomization and unstable liquid jet methods. The former method is the most appealing since the particle charging is accomplished most conveniently by means of the high potential which is already required for the acceleration. The unstable liquid jet method, on the other hand, is more appealing from the point of view of generating uniform-size drops, but may require a sep- arate charging scheme. A combination of the two methods may perhaps be the best method. A number of problems concerning waste processing must be solved before either method can be employed. The system has been envisioned to be in two main sections, the accel- erator section and the waste-container/particle-generation section. The accelerator section remains in orbit and is not recoverable. The waste- container/particle-generation section is recoverable, thereby making it" possible to segregate the particle production equipment, which is the least reliable, from the more massive acceleration section. This scheme allows launch-pad checkout of the most sophisticated portion of the system, and recovery if a malfunction occurs in orbit. The waste-container/particle- generation section is docked with the accelerator section once in orbit, and no separate transfer of materials between vehicles is necessary. In essence, the waste container becomes the ion source for the accelerator stage. 10. FUTURE WORK The amount of money remaining in the original allotment is $6000. In this section, a program plan is presented for the expenditure of these remaining funds. The contractor will use its best efforts to: 1. Carefully survey the present-day proposed matrix materials and investigate if any of these are suitable for use with the 8.D.29 accelerator method proposed in this study. Particular attention is to be given to the ease with which such material can be made molten during the particle generation step and the viscous prop- erties required in the liquid state for passage through small capillaries. 2. Investigate the materials problems which arise due to containment of the proposed molten matrix material. Present-day containment methods will be explored for suitability. 3. Investigate the various means by which the matrix can be "melted" once in orbit. A. Investigate the problems which could arise when a large bundle of capillaries are used instead of a single capillary for formation of the liquid droplets. 5. Consider an accelerator design based on the best practical esti- mates of particle size and surface field strength for the particu- lar matrix material considered. Include a crude cost and weight estimate. 6. Consider the power source needs for acceleration in more detail. This includes examining the potential use of on-board solar and thermionic converters and microwave beam transmission of power from the surface of the earth. 7. Investigate future research- and development goals which include a. Suggested laboratory studies needed for development of the small particle production technique. b. Suggested studies needed for designing and testing the par- ticle charging and ejection method. c. Suggested research and development needed for the construction of a moderate-sized prototype. d. Suggested studies required for the simulated space environment tests of the above prototype including lifetime testing of components. 8. Provide a crude cost and time schedule of the total research and development required prior to fabrication of the actual large- scale unit. 8.D.30 REFERENCES 1. R. E. Hyland, et al ., "Study of Extraterrestrial Disposal of Radio- active VJastes," Part II, NASA TM X-681A7, October 1972. 2. H. Shelton, C. D. Henricks, R. F. 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Journal , p. 80, March 1961. 39. E. N. Brown, "Rotating Bowl for the Production of Uniform Drops," Rev . Sci. Instr . 32 , p. 914, 1961. 40. S. I. Cheng, J. Cordero, "Droplet Formation from a Liquid Film over a Rotating Cylinder," AIAA Journal , 1_, p. 2597, 1963. 41. R. P. Fraser, et al . , "Drop Formation from Rapidly Moving Liquid Sheets," A.I. Ch. E. Journal , £, p. 672, 1962. 42. P. Gooderum, D. Bushnell, J. Huffman, "Mean Droplet Size for Cross- Stream Water Injection into a Mach 8 Air Flow," J. Spacecraft , 4^ p. 534, 1967. 43. J. Chen, V. Kevorkian, "Mass Production of 300 Micron Water Droplets by Air-Water Two-Phase Nozzles," Indust. & Eng. Chem . , ]_, p. 586, 1968. 8.D.33 44. W. Gauthier, "Operational Characteristics of the Ultrasonic Nebulizer," Proceedings of the First Conference on Clinical Applications of the Ultrasonic Nebulizer, Chicago, Illinois, January 22, 1966. 45. J. W. S. Raleigh, "On the Instability of Jets," Proceedings London Mathematical Society, X, p. 4, 1879. 46. B. J. Mason, 0. W. Jayaratine, J. D. Woods, "An Improved Vibrating Capillary Device for Producing Uniform Water Droplets of 15 to 500 urn Radius," J. Sci. Instrum ., 40_, p. 247, 1963. 47. T. Erin, C. D. Hendricks, "Uniform Charged Solid Particle Production," Rev. Sci. Instrum . 39 , p. 1264, 1968. 48. J. M. Schneider, C. D. Hendricks, "Source of Uniform-Sized Liquid Drop- lets," Rev. Sci. Instrum ., 35 , p. 1349, 1964. 49. N. R. Lindblad, J. M. Schneider, "Production of Uniform-Sized Liquid Droplets," J. Sci. Instrum ., 42, p. 635, 1964. 50. L. Strom, "The Generation of Monodisperse Aerosols by Means of a Dis- integrated Jet of Liquid," Rev. Sci. Instrum ., 40, p. 778, 1969. 51. H. P. Beerman, "Small Glass Drop Forming Device," Amer. Ceramic Soc . Bull ., 37, p. 272, 1958. 52. C. D. Hendricks, J. B. Y. Tsui, "Production of Uniform Droplets by Means of an Ion Drag Pump," Rev. Sci. Instrum ., 39 , p. 1088, 1968. 53. C. D. Hendricks, "Charged Droplet Experiments," J. Colloid Sci ., 17 , p. 249, 1962. 54. R. L. Peskin, R. J.' Raco, "Drop Size from a Liquid Jet in a Longi- tudinal Electric Field," AIAA Journal , 2_, p. 781, 1964. 55. J. J. Hogan, C. D. Hendricks, "Investigation of the Charge-to-Mass Ratio of Electrically Sprayed Liquid Particles," AIAA Journal , 3^» p. 296, 1965. 56. B. Raghupathy, S. B. Sample, "A New Apparatus for the Production of Uniform Liquid Drops," Rev. Sci. Instrum ., 41 , p. 645, 1970. 57. Available from Brunswick Corporation, 1 Brunswick Plaza, Skokie, Illinois 60076. BNWL-1900 APPENDIX 8.E Terrestrial and Stellar Contamination from Space Disposal of Nuclear Wastes 8.E.1 NASA-AMES, MS: 202-8 Moffett Field, CA 94035 April 13, 1973 MEMORANDUM for Chief, Space Applications Branch From : Lawrence C. Evans, Ph.D. Subject: Terrestrial and Stellar Contamination from Space Disposal of Nuclear Wastes As we have discussed, objections have been raised concerning the use of space as a sink for radioactive waste because of the possibility of these wastes contaminating other terrestrial or stellar environments. In the case of solar disposal, it is feared that the waste will be swept out from the sun by the solar wind, and when re-encountering the Earth will contaminate the terrestrial environment. In the case of disposal by means of solar system ejection, since the spacecraft will eventually encounter another stellar system, there is the possibility that the nuclear waste will contaminate that system. It is therefore appropriate to investigate several aspects of this problem in order to evaluate the validity of such objections. The results of such an analysis are summarized here, and the documented results are available elsewhere. The primary conclusion is that such contamination, both of the terrestrial environ- ment and of other stellar systems, is virtually impossible. The analysis leading to this conclusion with respect to solar disposal deals with the heliocentric vaporization distance of the waste, the ionization state of the atoms of waste, the effects of the juxta-solar radiation environ- ment, the interplanetary propagation of the ions, and their access to the terrestrial environment. In the case of disposal through solar system ejection, the main considerations are the probability of encounter- ing another stellar system, the distance traveled before such an encounter, and the time required for this travel. I will briefly summarize the solar disposal analysis, and then deal with stellar disposal. Before the solar plasma and magnetic field can interact with the material being injected into the sun, the material must be vaporized and ionized. In order to deal with the effects of the juxta-solar environment, we must therefore estimate the heliocentric distance at which vaporization takes place, and we must also estimate the state of ionization in the atoms. The vaporization should occur within one or two solar radii above the surface of the photosphere, and all of the atoms of the nuclear waste products will be very highly ionized. For instance, on the average, Strontium atoms will be stripped of approximately 17 of their electrons, 8.E.2 while Cesium atoms will lose approximately 24 of their electrons. The ionization states which the atoms attain near the sun will be the ionization states which they will maintain throughout the interplanetary propagation. Once the waste material is vaporized and ionized, it begins to be convected out from the sun with the solar wind plasma. In the meantime, however, these ions are emersed in the most intense radiation environment in the solar system, and it is critical to determine their survivability in such an environment. We can show that the ions will be within about 3 solar radii of the center of the sun for about 1 day. I have examined the survivability of the waste ions in terms of the probability of nuclear fission, and made the simplifying, if somewhat unrealistic, assumption that if a nucleus of waste material were to undergo fission, then the products would not be considered undesirable waste and their •introduction into the terrestrial environment would not be considered contamination. I have only been concerned with fission induced by three types of incident radiation: solar gamma rays, neutrons, and protons. In the case of' c photo-induced fission, the fission cross- sections for most of the*. radioactive waste materials are low enough that this is not a significant effect. Lifetimes for the waste materials are probably on the order of 10 19 seconds for photo-induced fission; compared with Sun-Earth travel time on the order of 5 x 10 5 seconds. The situation with particle-induced fission is, however, very much different. For proton-induced fission, it is particularly significant where the waste material is targeted on the Sun with respect to sources of charged particle radiation. If the waste materials are targeted into a quiet portion of the Sun, one might expect lifetimes on the order of 1.5 x 10 17 seconds, whereas if the wastes reach the Sun at the same time and place as a moderately large solar flare, one would expect lifetimes on the order of 10" 7 seconds. Although it is very difficult to anticipate and target for a solar flare event, it is possible to target for active regions on the Sun where the probability of flare events is much higher and where the ambient proton flux is also higher. Average lifetimes in these regions are perhaps 5 x 10 8 seconds. Neutrons will also induce fission, and, since they are uncharged particles, the fission cross-sections at low energies is much higher than those for incident protons. As a consequence, the probability of fission is very large even in the case of the waste materials being targeted for a quiet region of the Sun. In such a quiet region one would expect average lifetimes on the order of 10" 3 seconds, whereas if the spacecraft were to reach an active region of the Sun, one might expect lifetimes on the order of 10" 1 * seconds. If the waste material encountered a flare, the lifetimes would be proportionately shorter, reach- ing 10~ n seconds for large flares. Thus even if the waste were to be targeted to a quiet region of the Sun, we would expect the amount of waste material to be decreased by a factor of 1 million in 0.017 second. The problem of the propagation of the surviving waste ions from the vicinity of the Sun outward into and through interplanetary space is dominated by the fact that the ions have very low kinetic energies. As 8.E.3 a consequence, the spatial aspects of the motion of the v/aste ions in interplanetary space is determined by the configuration of the interplane- tary magnetic field, and it is therefore to the stochastic and ergodic aspects of this field which we must address ourselves. A consideration of these aspects leads to the conclusion that even those waste nuclei which do survive the juxta-solar environment will be spread out into a wery large diffuse cloud in interplanetary space. The fraction of these particles which would impinge upon the Earth is dependent upon the solar-longitude at which they were targeted, and this fraction varies from a maximum of 10" 5 to a minimum of 10" 100 . For that fraction of the waste particles which escapes the Sun and subsequently impinges on the Earth, we finally come to the question of the contamination of the terrestrial environment. In order to determine the degree of contamination, we must determine the fraction of those particles impinging on the Earth which actually enter the atmosphere and where this contamination will occur. Since the v/aste atoms are highly ionized when they reach the vicinity. of the Earth, their access to the terrestrial environment is determined by their interaction with the Earth's magnetic fie'jd. These particles are thus constrained to enter the magnetosphere through the low energy particle access regions far down the geomagnetic tail, and as a consequence, only a wery small fraction (approximately 10" 5 ) of those particles near the Earth in interplanetary space will actually impinge upon the upper atmosphere. It can also be shown that this contamination will occur only near the geomagnetic poles. Turning now from solar impact to an analysis of the contamination involved in disposing of nuclear v/aste by targeting them to escape the solar system, we ask the specific question of v/hether these wastes pose any realistic contamination threat to other stellar systems. Interstellar distances are so vast that one would expect that the travel times involved in encountering other stellar systems would be very large, and it is indeed so in this case. These distances are so large, in fact, that it would take several hundred decay lifetimes for the nuclear waste to reach any stellar systems. As a consequence, the amount of radio- active material would be reduced by a matter of at least 10 100 . This essentially means that the nuclear waste cannot survive the long trip times involved in going from solar system to some other stellar system. The net result of this analysis is, as you can see, the conclusion that disposing of nuclear waste in space by targeting them either to impact the Sun or to leave the solar system represents a final and complete disposal of these wastes and that wastes disposed of in this manner pose virtually no threat of contamination to either the terrestrial environment or any stellar environment. The details of this analysis are, as I say, contained in the more extensive report which is presently being typed and will soon be available for your perusal. Lawrence C. Evans, Ph.D. BLS LCEvans:bh 4/13/73:5891 BNWL-1900 APPENDIX 8.F Cost Estimate Data 8.F.1 BNWL-1 900 TABLE F. 1 . Transuranic Encapsulation Capital and Operating Cost Summary(a) $/kg of Transuranic Operating Cost Direct Labor $ 65 Direct Material 340 Indirect Manufacturing Expense 65 Materials and Supplies 50 Utilities and Services 50 Overhead 130 Total Operating Cost $ 690 Capital Cost Building and Equipment Cost at 18% Per Year $40,000,000 total cost; $7,200,000 per year * 2,600 kg/yr 2,800 Sub Total 3,490 Contingency, 30% 1 ,010 Total Encapsulation Cost 4,500 Level ized encapsulation cost charged at time of reprocessing (per Volume 1) 3,700 Levelized heat shield cost^ 1,000 Total Levelized Cost for Encapsulation and Heat Shield 4,700 Based on a plant encapsulating 2,600 kg/yr of transur- anics or the transuranics from about 3,650 MT/yr of LWR uranium fuel . Based on heat shield cost estimate of $220,000 for a unit, regardless of size (per personal conversation with John Vorreiter, NASA Ames, June 1973). 8.F.2 BNWL-1900 TABLE F.2 . Direct Labor and Equipment Estimates, Waste Transuranic Encapsulation for Space Disposal Basis: Plant to handle waste from two 1825 MT/Year Reprocessing Plants, 2600 kg of transuranic capacity per year, 180 operating days/year ■ 15 kg/day, 70% operating efficiency, 6 hr/ shift, throughput required - 4 kg/hr. Operation Throughput per Machine kq/hr No. of Machines Requi red Per Machine $ Total Equipment, $ No. of People Required Particle Preparation 1. Purification 5 1 20,000 20,000 1 2. Powder Consolidation 10 1 5,000 5,000 0.5 3. Crushing 10 1 5,000 5,000 0.5 4. Sizing 10 1 10,000 10,000 0.5 5. Sintering 10 1 25,000 25,000 0.5 6. Allowance for Recycle — -- -- 65,000 2 Coating 7. Coating Material Treatment 8. Magnesium Coating 9. 1st Tungsten Coating 10. Sacrificial Material Removal 11. Decontaminate 12. Second Phase Tungsten Coating 13. Aluminum Oxide Coating Assembly 14. LiH Treatment 15. Al Treatment 16. LiH-Al-Act. Oxide Mixing 17. Compacting 18. Clad/Prep 19. Assembly 20. Joining TOTAL 10 5,000 5,000 10 10 10 10 10 10 10 10 10,000 10,000 1 20,000 20,000 1 10,000 10,000 1 5,000 5,000 1 100,000 100,000 1 20,000 20,000 1 30,000 30,000 1 200,000 200,000 $ 550,000 $ 675,000 0.5 10 1 20,000 20,000 0.5 2 2 20,000 40,000 2 5 1 20,000 20,000 1 5 1 5,000 5,000 1 2 2 20,000 40,000 2 0.5 0.5 0.5 0.5 0.5 0.5 J 17 Note: Using an average manpower salary of $10,000/yr, total labor cost is $170,000/yr or $65/kg of transuranics. 8.F.3 BNWL-1900 TABLE F.3 . Direct Material Estimates, Waste Transuranic Encapsulation for Space Disposal Pounds of Material Per Pound . of Transuranic' 3 ' Magnesium 0.5 Tungsten Hexafluoride 2 Aluminum Oxide 0.2 Lithium Hydride 2.6 Aluminum 1.7 Tungsten 6.3 Stainless Steel 5.2 Total, $/lb of Transuranic Cost, $/kg of Transuranic Cost $/lb of Material Cost, $/lt>, of Transuranic 0.39 0.20 10.00 20.00 0.14 0.03 8.05 16.10 0.408 0.81 10.00 63.00 10.00 52.00 $152.14 $340.00 a. See Table F.4 for approximate amount of materials used. 8.F.4 BNWL-1900 TABLE F.4. Approximate Weight Distribution of Capsule of Transuranic Waste for Space Disposal Material Approximate Weight, kg/package( a ) 191 Approximate Unit Weight, kg of Material Per kg of Transuranic Transuranic 1.0 Matrix Aluminum LiH 625 1.7 1.7 Lithium Hydride 178 0.9 Tungsten Shieldi ing 1,190 6.3 Stainless Steel Shell 640 5.2 Other, Re-entry Shield, Fission Products, Oxygen, Uranium 446 2.3 Total 3,270 Based on solar escape with capsule containing 0.1% of the original fission products (see Table 8.5). BNWL-1900 APPENDIX SECTION 9: TRANSMUTATION PROCESSING APPENDIX 9. A TRANSMUTATION BY ACCELERATORS Pa^e SUMMARY 9.A.1 DETAILED RESULTS 9. A. 2 A. Charged Particle-Nuclear Reactions 9. A. 2 B. Beta Decay Acceleration by Coulomb Excitation 9. A. 5 C. Photon Transmutation Processes 9. A. 16 - Electron Bremsstrahlung 9. A. 16 - Stimulated Gamma Emission 9. A. 20 D. Spallation Accelerators 9. A. 25 - The Spallation Accelerator 9. A. 25 - Feasibility 9. A. 30 - Other Proposals 9. A. 38 REFERENCES 9.A.40 APPENDIX 9. A TRANSMUTATION BY ACCELERATORS SUMMARY The BeV proton-induced spallation device is the only accelerator transmutation concept showing promise. Other accelerator schemes can be ruled out on the basis of the critereon of energy balance. Conclu- sions reached are: (1) The inventory of intermediate half-life radioactive waste materials (Sr-90, Cs-137 and Kr-85) can be reduced signifi- cantly only by wery high neutron flux levels, on the order of 10 -10 neutrons/(cm )(sec). Materials problems in the transmutation target, including its clad, and other parts of the system can be anticipated to be severe. (2) The energy invested per neutron produced when BeV protons strike heavy target material is the order of 50 MeV/neutron. The transmutation of Sr-90 and Cs-137 with perfect neutron utilization would thus use about 1.0 and 2.5 MeV of the energy per fission that produced them. Adding other fission products for transmutation or operation at 50% efficiency then uses at least 10% of the recovered energy from fission to transmute the fission products produced by that fission. (3) The spallation intense neutron source will produce large amounts of short-term radioactivity. The approximately 20 neutrons produced per incident BeV proton will come from on the order of 5 to 10 parent nuclei, almost all of the daughters being left in radioactive precursor states. (4) The inventory reduction is significant if long-life fission products 1-129 or Tc-99 are exposed to a high thermal neutron flux. Since these materials have extremely long half-lives for natural decay (1-129 half-life is 16 million years), transmutation appears as a possible alternate to other means of handling these materials, particularly the higher hazard 1-129. 9. A. 2 (5) Preliminary calculations indicate that if fission product nuclei are used directly as the target for the accelerated protons, then the transmutation might be technically feasible. The capital cost of the many accelerators which would be required, however, would be prohibitive. In order to have any practical possibility, very large advances are required in high-energy accelerator technology. It is therefore likely that accelerators would find application only for eliminating isotopes of special concern, such as 1-129. DETAILED RESULTS A. Charged Particle - Nuclear Reactions It can be demonstrated from fundamental considerations that direct nuclear reactions of charged particles from accelerators with a few tens of MeV energy are not particularly attractive for radioactive waste burnup. Charged particles from fusion reactions and photonuclear reactions will be considered independently, as will BeV proton induced spallation processes. Proton penetration at intermediate Z nuclei requires energy on the order of tens of MeV, with more energy required for heavier particle penetration. The larger energy requirement for penetration by nucleons with higher charge than protons and their larger Coulomb energy losses in material targets make them less likely candidates for transmutation of wastes. Consider a charged particle slowing in a medium containing transmutation targets. The number of nuclear reactions N per projectile in slowing from some initial energy E. to a reaction threshold energy E. is given by I E t N, o E. l t "t dE (Al) 9. A. 3 where N. is number density of transmutation target nuclei and a. is the cross section per nucleus. The expression dE/ds is the particle's energy loss per unit path length. An effective cross section can be defined by writing N - N t J t ■>#,- (A2) E n m The energy loss per unit path can be re-expressed in terms of a density-independent tabulated function of material and energy where M is the mass per unit area of material traversed and p is mass density of the material. The average of (dE/dM)~ for reasonable energy intervals can be taken as some multiple of a (dE/dM) at ten MeV in aluminum. ' N t E i " E t N = a T W 10 MeV,Al The effective cross section must be some number b of barns, while the energy range E. - E. must be some number c of tens of MeV. The ratio N./p will be no isotopic target, for which case MeV. The ratio N./p will be no higher than in the case of a pure — =-r (A5) Here N is Avogadro's number and A is atomic mass number. Accordingly, Equation (A4) can be written , , 6.023 x 10 23 Y in -24 Y 10 abcdx go xlO x ^ 4xio abed x 2 x 10" 3 (A6) 9. A. 4 where d is a factor to account for departure of target atomic mass from 90. Since units were chosen to make abed of order -3 unity, it can be concluded that only on the order of 10 of the charged particles experiences a nuclear reaction. Since the energy cost per charged particle is greater than c x 10 MeV, the energy input per reaction is greater than I = abd x 500 ° MeV / reaction - ( A7 ) This energy per radioactive waste-eliminating reaction should be compared with 200 MeV/fission thermal energy available from the fission process. Even a multiple reaction (waste eliminating (p,n) reaction followed by waste eliminating (n,y) reaction, etc.) does not help much. Use of energy from those waste-eliminating reactions which are exothermic does not aid materially since they tend to release less than 10 MeV per reaction. Energy deposited by "the charged particles in slowing down can be recovered only by thermodynamic means since they are quickly thermalized. Thermodynamic efficiency for the utilization of this energy would probably be no greater than 0.4 as determined by materials considerations and available reservoirs. Utilizing the thermal energy supplied to the target, an energy increment aE per reaction must still be supplied, satisfying E - (1 - n) | (A8) > (1 " ' 4) ' aR x 500 ° MeV/reaction (A9) > ^d x 3000 M eV/reaction. (A10) Since the number of fission products expected to transmute per _2 fission is of the order of 10 to unity, the energy expenditure required with accelerator beams of charged particles of a few tens of MeV energy by direct interaction is excessive. 9. A. 5 To make waste transmutation by beams of charged particles from accelerators more nearly feasible, two improvements are needed. First, the probability of a reaction must be increased. Second, the number of waste nuclei eliminated per reaction must be increased. If the energy of a proton is increased above 150 MeV, the mean free path for nuclear interaction becomes shorter than the range. This is principally because the energy loss per unit path length decreases with energy at high energy. Proton interactions at high energy in high Z targets result in the emission of multiple energetic cascade particles which in turn evaporate many neutrons. B. Beta Decay Acceleration by Coulomb Excitation Beta decay from certain excited states of some nuclides proceeds more rapidly than beta decay from the ground state or in other cases more rapidly from the ground state than from certain metastable states. If this is the case for a radioactive fission product, it may be possible to reduce the effective lifetime of a nuclide by inducing a transition to a more rapidly decaying state. The FP isotope Kr-85 is representative of some of the circum- stances desired. Beta decay from the 9/2+ ground state proceeds with a 10.76 year half-life. The 1/2- excited state at 0.305 MeV decays with a 4.4 hour half-life, 77% by beta decay to the 3/2- state of Rb-85 and 23% by gamma decay to the 9/2 + Kr-85 ground state. Figure A.l displays an energy level diagram for Kr-85. The Kr-85 case also illustrates the difficulty in attempting to accelerate beta decay. The 1/2- - 9/2+ transition is slow because of the 4 unit change in angular momentum required; hence, the B~ decay to the 3/2- state of Rb-85 competes favorably with gamma emission. To excite Kr-85 to the 1/2- state also requires a change of 4 units of angular momentum. Large angular momentum changes can occur in Coulomb excitation, and the cross section falls off less rapidly for higher angular momentum changes than do decay rates in gamma decay. 9. A. 6 4.4 hr. 1/2 - °- 305 Q 10.76 yr. 9/2+ t 77% Q fl = .67 MeV 8 0.514 Rb-85 FIGURE A.I . Energy Level Diagram for Kr-85 9. A. 7 Therefore, necessary, but not necessarily sufficient, conditions on cross sections may exist for Coulomb excitation as a possible excitation method for radioactive waste trans- mutation. The recoverable energy per fission energy and the cumulative yields of objectionable fission products sets an upper limit (dE/dN) to the energy expended per fission max product nucleus transmuted. If it is decided to expend no more than a fraction of the recovered energy per fission in treating f fission products per fission, then (4=-) = (Fn/f) x 190 MeV/fission (All) max where n is the thermodynamic efficiency of the reactor. A charged particle passing through fission products of number density n will have a probability per unit path length dN/ds of eliminating a fission product given by ^=n„ (A12) where a is the cross section for a reaction leading to elimination of the fission product. The particle meanwhile will be losing, energy at a rate dE dE ds " p dM (A13) where dE/ds is energy lost per unit path length, dE/dM is energy lost ^er unit mass/area, and p is the density. The energy expended per fission product eliminated will be dE - dE/ds _ pdE/dM A dE (A14) dN dN/ds no N o dM o where A is atomic number of the fission product and N is Avogadro's number. Requiring dE ,dEv (A15) d¥ Wmax requires 9. A. 8 A dE /#dEx > N dM / Wmax (A16) This requirement on the cross section for transmutation elimination must be satisfied for at least part of the charged particle track if the elimination is to be economically feasible. To exemplify the above criteria, consider Kr-85 with f = 0.013 (0.013 nuclei per fission). Assume (optimistically) willingness to expend 5% of the recovered fission energy to eliminate this isotope. Then an upper limit on energy expended per Kr-85 nucleus transmuted "is: (4jj-) = (Fn/f)xl90 MeV max = (.05x.4/.013)xl90 MeV (A17) = 292 MeV/ Kr-85 nucleus. Assuming a 40% thermodynamic efficiency for the power plant and using tabulated and calculated values for the stopping power of Kr-85, the cross sections required at various energies are obtained for economically competitive transmutation by protons as shown in Table A.l. To estimate whether the cross section magnitudes needed for economical waste transmutation occur physically, the semiclassical quantum theory of the Coulomb excitation process is examined. (Semi- classical means here that the incident ion is considered to move in the 9. A. 9 TABLE 9.A.1 . Cross Section Needed For Economical Transmutation of Kr-85 by Protons Assumption: 5% maximum of recovered fission energy to treat 0.013 Kr-85 nuclei per fission. Proton Energy (MeV) Required q(Barns) 5. 20. 7.5 15. 10. 12. 12.5 10. 15. 8.8 20. 7.0 50. 3.3 100. 1.9 200 1 .0 300 0.73 400 0.57 9. A. 10 classical Rutherford trajectory). The cross section for Coulomb excitation of a 2 A order multipole transition in the semi-classical approximation is given by Z.e 2 °EA = (flljjO a" 2A+2 B(EA) f Q U) (A18) where Z.e = incident ion charge u = initial relative velocity fi = Planck's constant/2-rT 2 2 a = Z.Z VMiT l e ' = half distance of closest approach in head on collision 7 7 2 r a_w ^ _j aE u hu 2E B(Ea) = reduced electric multipole transition probability of multi polarity A £ I | 2 M i M f u M(EA,ij) = nuclear electric multipole operator of order A r A Y AM (e,*) p (r)dx Here p(r) is the nuclear charge density operator. 9. A. 11 The functions fr (?) are approximately constant for C<<1 (although quite different for different a) but fall off rapidly for £>1. Coulomb excitation cross sections tend to be too small to observe for £>1 The condition 5 < 1 requires that the energy of the incident ion be sufficiently large to excite a given excited state. To Coulomb excite the 0.305 MeV level in Kr-85 requires protons of energy of 1 MeV or greater. To estimate the magnitude of the cross sections for the high order multipole transitions needed for beta decay acceleration by Coulomb (2) excitation, following J. D. Newton x ' and using a "single particle model" estimate of the reduced electric multiple transition probability: B, n (EA) = fi R 2A (A19) sp 4tt 1 /3 where R = r Q A - is the nuclear radius. This model gives as the Coulomb cross section of electric multipole order 2 A : Ea Uu ) 4tt a ( ? } f EA EA=l,-TT h ■'(^WO (£) of Adler, et al. The Table stops at E4 because Adler, et al . , gives fr x (€) values only to X = 4. From Table 9. A. 2 it appears that the minimal required cross sections of Table 9.A.1 will not be obtained for objectionable radioactive iso- topes. An open question is whether the "single particle model" reduced transition probabilities B(EA) represent an adequate representation for a realistic assessment. The B(E1) values found experimentally are smaller by factors 10 3 to 10 5 than the B (El) used here, but B(E2) values are frequently enhanced by factors of up to 100 over the single particle values. Enhancements of higher order pole moments probably also occur, but E3 transitions are the highest order Coulomb excitations known to have been identified. Second order quadrupole transitions E2,E2 occur and may be applicable in obtaining some desired states, but seem inapplicable to Kr-85 for two reasons: (1) the needed parity change would not be given, and (2) an intermediate state accessible by E2 transition is not known (though it may exist). Coulomb excitation may occur simultaneously to a number of energy levels higher than the beta decaying metastable level with subsequent cascading that leads to the metastable state. Assume that five such levels exist, each having the "maximum plausible" cross section of 3 barns each. Then if all levels cascaded without loss to the known beta decaying metastable state and 11% of the excitations result in beta decay, the effective cross section for elimination would be 11.5 barns, which is close to the estimated "local minimum" cross section requirement for 10 MeV protons given here. The actual cross sections should be higher still, since compensation for 9. A. 13 TABLE 9. A. 2 . Coulomb Cross Section Estimates For A Z=36, A=85 Nucleus Using Single Particle Model Reduced Transition Probabilities Proton Energy °Ei( m " il " ,ibarns ) a E2 (millibarns) ° E3 (millibarns) ° E4 (mil libarns) 20 MeV 3.1 0.57 0.45 0.65 10 MeV 4.4 0.28 0.06 0.02 9. A. 14 accelerator inefficiency and the almost total loss of the last MeV of proton energy in stopping in the target material is needed. That such a fortuitous combination of physical properties exists is unlikely but cannot be ruled out. This discussion has focused attention on numerical examples of beta decay acceleration by Coulomb excitation on Kr-85 only because it meets some of the physics requirements. It is in fact a poor candidate for transmutation, since its 10.8 year half-life and noble gas nature suggest retrievable storage. Sr-90 does not look like a good candidate for beta decay acceleration by Coulomb excitation, because its approximately 11 known excited states between 0.83 and 5.23 MeV gamma decay rapidly. The only hope for this concept for Sr-90 would be that as yet undiscovered metastable states at energy much less than 0.83 MeV exist. It is not known whether this can be ruled out from present experimental information, although the existence of such states is certainly doubtful. Cs-137 is also lacking in identified low energy fast beta decaying levels accessible by Coulomb excitation; the excitations result only in rapid gamma decay back to the ground state. 1-129 has a wealth of low lying energy levels, including a 0.0178 MeV 5/2+ level which should be readily excitable by a Coulomb excitation E2 transition. Unfortunately, the fairly rapid Ml (magnetic dipole) trans- ition returns it to the ground state with a 15 nanosecond half-life and negligible beta decay. No beta decay has been observed from any of the low-lying excited levels of Tc-99. 9. A. 15 It is concluded that beta decay acceleration by Coulomb excitation is probably not applicable to most of the problem fission products, although surveillance of nuclear physics developments for these materials may be warranted. It does not appear likely that any of the other fission products are amenable to beta decay acceleration Coulomb excitation. However, materials like Ho-166 whose lowest energy state decays more rapidly than a metastable state, as shown in Fig. A. 2, may merit further study. 7- 26.9 hr. Ho-166 009 1.2 x 10 yr, \v Er-166 FIGURE A. 2 . Ho-166 Levels and Decay 9. A. 16 C. Photon Transmutation Processes Electron Bremsstrahlung - As an alternate approach to fission waste burnup by. transmutation, one can consider bombarding targets contain- ing objectionable waste with electron beams in the tens to low hundreds of MeV range. The electrons produce a "shower" of photons from Bremsstrahlung, from annihilation of electron-positron pairs, and from other processes. Some of the photons in the "giant" resonance energy region will undergo nuclear interactions such as (y,n), (y,p), (y,2n), or (y,np). The feasibility criteria for such a system are the reaction yields; i.e., number of reactions per incident electron, and the energy supplied per reaction. The parameters for optimization of this concept are influenced by the following considerations: (1) Initial electron energy should be sufficiently high that radiation energy losses in matter are of the same order as or higher than ionization losses. Above some critical energy E radiation losses dominate. Sample critical energies are 6.9, 24, and 103 MeV for Pb, Fe, and C, respectively. (2) The energy spectrum of the bremsstrahlung should overlap the "giant resonance" region of the target. The photon energy of the center of the giant resonance is given approxi- mately by E res = 40 A "°' 2 MeV (A21) where A is the nuclear mass number. 9. A. 17 (3) The bremsstrahlung spectrum has as an upper bound the energy of the decelerating electron, with most of the radiation being emitted at much lower energy. These three considerations taken together dictate that the electron beam have at least tens of MeV initial energy. (4) Barber and George v ' suggest that an approximate yield of neutrons from thick targets bombarded by electrons can be calculated from Y(Z,E ) - K (/c(Z,k)dk) I(Z ^' R(Z) ^ ^ (A22) where t (Z) is the number of grams/cm in one radiation length, A is the atomic weight of the target material, E the incident electron energy, and k is the photon energy at the peak of the giant dipole resonance. The integral dk (A23) /o(Z,k) is the photon energy integral for producing a neutron by any reaction The expression R(Z) I(Z) + R(Z) (A24) is a ratio of the electron energy loss by radiation to that by both ionization and radiation. The constant K was determined to give agreement with their experimental yield in lead with E = 34 MeV. Taking ratios of ionization and radiation energy losses at an electron (4) energy of 30 MeV, Barber and George generate anticipated yields of neutrons per incident 34 MeV electron shown in Table A. 3. 9. A. 18 From the data in Table A. 3 it is clear that neutron yields from the electron bremsstrahlung-photonuclear reaction process increase with increasing nuclear size. Hence, any system for radio- active waste transmutation using 34 MeV electrons is likely to have a photoneutron yield less than that of uranium, which Barber and (4) George v ' estimate as Y(92, 34 MeV) = 2.59 x 10~ 3 neutrons/electron (A25) (5 61 A number of experiments ' Indicate a roughly linear increase in yield with incident electron energy above a few tens of MeV. Hence, yield per electron can be increased by increasing electron energy, but energy expended per reaction could not be reduced significantly. Energy per reaction E/N will be of the order |=- 34_MeV = ■ L31 x 1Q 4 MeV/reaction (A26) N 2.59 x 10"- 3 Since this is s/ery large compared with 200 MeV/fission, the concept does not meet the energy balance criterion. 9. A. 19 -o o .— +->^-^ cu -a •r-T)r- >- CD O) N -r- "O -r- >, O) .— 4-> CO -Q u E Q-«3- a» i- o Q- O Q-r— x 2: X LU O) X «* o 00 r— r- sO CO CM CX> (£> O LO r- (\J M a> E O s- 4- O) o +-> r— O cu x X LO CO ID CVJ 3 CsJ > E a; u 2: r~ | CO E en +-> fO > cu cc 2: "3- «3- CX> t-«. vD O CO CO CO sO CO a> LO u c cu £= ■0 fO c S- cu Q. to 2: CU c Q O T3 1 S- E M -»-> ra O T3 CU s- CU 1— cu +j LU 1 — <*s 1— 1— > •1 — ZJ CU F u 2: — ■ cu o CO n3 CM E a cu CsJ E u LO LT) CM CM CsJ «3- "=3" CM CVJ sO 00 LO LO CO ■ CTt 00 cu CO CVJ CM LO r~- CO CT> 00 LO CO CO ■o cu cu 000 O CM o o o o lo co r^ CO 00 CM 1— CO co «sj- r>« 4-> cu E cu (_> CJ> LU 9. A. 20 Stimulated Gamma Emission - The phenomenon of stimulated emission appears attractive on first thought as a radioactive waste manage- ment technique. A material matrix is visualized including a waste isotope emitting its characteristic radiation, with probability of emission by a given nucleus increasing with increasing intensity. The result would be the elimination of the radiation hazard on a short time scale rather than at the natural decay rate. The two major problems which tend to discourage hopes for this concept are: t Stimulated emission is possible only for neutral Boson emission according to best available theory. For practical purposes, this means photon emission. There is no stimulated emission phenomenon in beta decay. • The low attenuation material medium needed for stimuated gamma decay may be unattainable. Applicability only to gamma decay, limits interest in stimulated emission for radioactive waste to a few slow gamma emitters. Metastable Ho-166 and metastable Sn-121 illustrate a situation which might be considered for stimulated emission. If a transition from the 1200 year half-life metastable state of Ho-166 to the 26.9 hour beta-decaying ground state could be induced, (see Figure A. 2) the retention time for this component of radioactive waste could be reduced. Similarly, it may be desirable to stimulate a gamma transition from the 76 year half-life 11/2- state of Sn-121 to the 27 hour 3/2+ state. Consider the requirements of stimulated gamma transitions in an extended medium. If gamma stimulation cannot be achieved in an infinite medium, then it clearly cannot be achieved in a finite one in which diffraction and mirror losses occur. The condition for stimulated 9. A. 21 emission in an extended medium is that the stimulated emission gain per unit length exceed the attenuation loss. The intensity change per unit length dl/ds will be dl / v hv B..(v) I(v) g. (h?i\ -hi = LL (n, - - 1 n.) - BI(v) • < A27) ds c i 9j J The first term on the right hand side represents stimulated emission gain, while the second represents attenuation. The coefficient B^(v) is the Einstein induced emission coefficient, related to the spontaneous transition probability A...(v) by B..(v) =^-A (v). (A28) 1J 8Tihv 3 1J The spontaneous transition probability A.j(v) is related to the transition probability A., from state i to state j by a line shape factor: A..(v) -^U. ^ (A29) U V (v-v o ) 2 + (Av) 2 . The preceding line shape is appropriate in the case of a homogeneously broadened line, i.e, one whose width is determined by the finite life- times of the upper and lower states. For the materials of interest for stimulated gamma emission, this would probably occur only at cyrogenic temperatures in nearly perfect crystals. For such a homogeneously broadened line, the line width is related to the total decay constants of states i and j by where Av = (A. + A.)/2tt (A30) A. = z' A.. 1 k ik 9. A. 22 is the decay rate constant from state i and the sum is over states to which decay occurs. The criterion for stimulated emission at line center becomes -^- — -^ ( n *!i n .) > B (A31) 4uv 2 A. + A. q. J This expression shows the devastating effect on the requirements for stimulated emission which stem from alternate decay modes of the upper excited level. This can be seen from the alternate form: r2 ♦*,- 9 U (n. -II n,)>6 (A32) 4ttv 2 1 + (t./tj) ^ q. J where a., is the fraction of the decay from state i that goes to state j a t. is the lifetime of state i. nd The attenuation coefficient B describes all processes which take a gamma photon out of the beam other than nuclear resonant absorption by the isotope of interest. Here any scattering process that changes the direction of the photon or the phase of its wave function effectively removes it from the coherent beam, just as do all absorption processes. A transition from the Ho-166 metas table state to the Ho-166 ground state would produce a 7 keV gamma. For this gamma, the photoelectric effect is the dominant attenuation mechanism. Using graph values for attenuation of 10 keV gammas in lead and an extrapolation formula given by Evans, ' we estimate the attenuation in holmium as 9. A. 23 3 = 390 cm -1 For the desired holmium transition we estimate Cf *JA ( n - !in) = w*.. (1 -^-/) x 1.23 cm" 1 (A33) 4 ttv 2 1 + (t./t.) n g j J 1J 9 J n i where w is the ratio of density of the holmium in the material medium to the density of elemental holmium. Since each of the symbolic factors on the right side of the preceding equation is less than unity, the criterion for stimulated emission clearly cannot be met. The factor tj).. is in fact much less than unity because most of the decay is by beta emission rather than the highly forbidden gamma transition we seek. Clearly, stimulated emission for radioactive waste management requires some orders of magnitude reduction in attenuation of gammas in the material medium. Such a reduction in attenuation in specific crystallographic directions has been observed (Borrmann effect) and has (8) been proposed v ' for stimulated gamma emission. An electromagnetic wave of appropriate wavelength incident on certain crystals at an angle appropriate for Bragg scattering will split into a pair of transmitted waves (forward diffracted) in addition to the wave diffracted out of the crystal. If the transmitted waves interfere to give electric field minima at the planes of the atoms, attenuation is dramatically reduced. Difficulties with applying this anamolous transmission to stimulated gamma emission are: 9. A. 24 (1) Having electric field minima at the atom planes may (9) reduce stimulated emission as much as it reduces attenuation. ' (2)' A highly perfect crystal would have to be grown in the presence of disruptive radiations. Stimulated emission for radioactive waste elimination would require considerable development work to be applicable only to a limited set of radioactive nuclides. It may well be totally impossible. The possibility of stimulated gamma emission as a radioactive waste trans- mutation process depends on technical developments which cannot be anticipated and appear to be unlikely. 9. A. 25 D. Spallation Accelerators High energy (_> 1,000 MeV), large current proton accelerators have been proposed which would provide the most intense, continuously operating source of neutrons yet attained. The largest spallation accelerator which has been seriously studied was the Intense Neutron Generator (ING) which was proposed by the Canadians^ ' (Atomic Energy of Canada, Ltd.). The conceptual design of the ING envisioned accelerating a 65 mA beam of protons to an energy of 1 GeV (1,000 MeV). The beam was to produce high energy neutrons primarily through the spallation process in a Pb-Bi target. The high energy neutrons were to be moderated in a surrounding D ? medium to provide a thermal neutron flux of at least 10 n-cm~ -sec" . Thermal flux values of this magnitude are required for the transmutation of low neutron cross section FP isotopes. In 1967, Steinberg, who had for some years been studying neutron transmutation of fission products, " ' proposed the usage of a higher powered version of the proposed ING for this (13) purpose. In 1971 an 0RNL group v ; also reported brief conclusions regarding Steinberg's proposal. A Japanese atomic energy teanr ' has also studied the use of a spallation accelerator for the transmutation of fission products. In addition, several studies have been made of the use of spallation accelerators to transmute fertile to fissile material, a concept which is closely related in technical feasibility to the FP transmutation concept. The Spallation Accelerator - In order to discuss the technical feasi- bility of transmutation using a spallation accelerator it is necessary to under- stand the physical processes which take place in the spallation target and environs so that feasibility factors can be clearly identified. 9. A. 26 The range of a 1 GeV proton in a heavy element is the order of 30-60 cm. In the spallation reaction of 1 GeV protons about 0.5 pi mesons with an average energy of about 100 MeV are formed independent of target material. In a thick target these pions rapidly undergo nuclear interaction or decay within a short distance of where they are formed. A few percent (M>%) of the incident protons result in a component of (15) yery high energy neutrons. ' These neutrons are referred to as the cascade neutron component and are characterized by an evaporation energy spectrum with a Maxwellian temperature of the order of 250 MeV for a 1 GeV (15) proton team. ' The highly excited residual nuclei formed in the primary cascade reactions can decay by the evaporation of one or more nucleons which may be followed by fission and further neutron emission. The cascade nucleus may also fragment into two or more excited fragments followed by neutron evaporation. The neutrons evaporated from highly excited nuclei are characterized by an energy spectrum with a characteristic temperature of 2.5-3 MeV. The low energy part of this neutron spectrum is degraded rapidly to a few tenths of an MeV by neutron inelastic scattering in the target material . The dominant spallation reactions in a Pb-Bi target lead to a variety of undesirable radioactive nuclei with Z(proton) and N(neutron) values slightly less than those of the struck nucleus. ' These include: 9. A. 27 Nucleus T l/2 Bi-208 3. 68x1 5 yr Bi-207 38 yr Bi-206 6.24 d Pb-205 1.4xl0 7 yr Pb-203 52 h Tl-204 3.78 yr Tl-202 12.2 d Very short half life products are not shown on the above table. The sum of all of the nuclei of this type formed in the spallation reaction is 0.5 to <1 per incident proton. The fission plus fragmentation cross section of Pb and Bi is not well known. At a proton energy of 340 MeV this cross section is about 0.2b which is about 13% of the total nonelastic cross section. The cross section for fission is expected to peak near 750 MeV. According to theoretical predictions the fission cross section will never exceed half of the non- elastic cross section. Thus near 1 GeV an estimate of 0.3+0.15 would seem to be reasonable for the fraction of nonelastic processes leading to fission and fragmentation. Within a few percent, all incident protons undergo a nonelastic event in the target. 9. A. 28 The kinetic energies of ultimate fission and fragmentation nuclei and of fission neutrons represents the exoergic reaction energy liberated in the target and environs which is gained in the spallation reaction. Kinetic energies of fragments has been measured to be 111 MeV per fission for 435 MeV protons on bismuth. By comparison with the variation of fission excitation energies with incident proton energy on other nuclei it appears that the fragment kinetic energy should be the order of 200 MeV per fission on Pb or Bi near 1 GeV. The additional energy liberated in the target due to fission fragments can then be estimated as (0.3+0.15) x (200+50) = 60+35 MeV/p It is known from experiment that there are about 20 neutrons produced per 1 GeV proton incident on a thick lead target. Of this number, perhaps half represent exoergic fission neutrons kinetic energy. Assuming an average energy of 2 MeV for these neutrons would result in an energy augmentation of about 20 MeV. Combined with the fission fragment kinetic energy we then expect an energy libration of some 1080 MeV per 1000 MeV proton. This energy augmentation is considerably less than has been estimated for 900 MeV protons incident on a U-235 target. In this case (15) a value of 1295 MeV per 900 MeV proton has been estimated. ' The fission cross section for U-238 is, however, considerably higher than for Pb or Bi and also U-238 fissions occur from secondary neutron reactions. 9. A. 29 The product nuclei resulting from 1 GeV proton fission and fragmentation events have a distribution with mass number which is quite different in character from that observed for slow-neutron induced fission. For protons with energies approaching 0.5 GeV the distribution of fragment masses is roughly symmetric about A = 95 with a near constant yield for some +20 mass units. At a proton energy of 3 GeV, however, the distribution of fragments is constant with A to within a factor of several. Thus, a whole host of radioactive decay products are produced in the spallation reaction, many of which are not encountered in neutron induced fission. In the ING concept the Pb-Bi target of a few cms radius was to be sur- rounded by a beryllium sleeve to achieve further multiplication of the source neutrons by n,2n reactions on beryllium. The projected source strength including neutron multiplication in Be^ ' was 1x10 n-sec" . It is important to note that this is a production of 1-1/4 moles of neutrons per day and that this figure is absolute maximum quantity of FP nuclei that could be transmuted without further neutron multiplication. The spallation target with Be sleeve of the ING was to have been surrounded by a D ? moderator tank to provide a maximum thermal neutron flux of 10 -2 -1 n-cm -sec as a source for neutron physics experiments. This flux value depends upon no absorption of neutrons other than in the D^O moderator. The presence of any absorbing FP target material would lower this value. 9. A. 30 Feasibility - The feasibility criteria are applied to determine the merit of using a spallation accelerator to transmute fission products. Energy Balance The required electrical energy, E, input to the spallation accelerator to produce N net neutrons available for fission product transmutation can be expressed as: E E o /e a ~ n th (E th ) (A34) N " (N'-F)e b where: E = the energy of the proton accelerator e = proton beam energy/proton on target input accelerator power/proton on target E.. = thermal energy deposited in target and environs n th = e ^^ lclenc y °f conversion of E.. to electrical energy M' = neutrons produced per incident proton F = radioactive nuclei produced per incident proton z. = fraction of neutrons absorbed in FP A minimum energy/FP can be estimated by assuming the most optimistic condi- tions including the maximum conversion of thermal energy in the target to electrical energy. For the minimum value of E/N it is assumed that E rt = 1000 MeV o E th = 1080 MeV n th ■ 0.40 e a = ] = e b N' = 24.6, F = 9. A. 31 Then the minimum value E 1000 - 0.4 (1080) > 24.6 = 23 MeV/FP (A35) A reasonable upper bound to E/N can be obtained from more pessimistic assumptions. For the first assumption it can readily be assumed that no thermodynamic conversion of target heat to electricity is possible. There are a number of possible reasons why this might be a realistic assumption. The ING was designed to utilize a flowing liquid lead-bismuth eutectic target without a window for the proton beam. The liquid target temperature is then limited in order that the eutectic vapor pressure not exceed the 10" torr vacuum requirement for beam transport. Besides being thermodynamically limited in temperature, materials damage problems could preclude the use of an effective heat transfer system near the target. The 100% efficiency assumed for the conversion of electrical energy to proton beam energy is clearly idealistic. Davidenko^ ' has suggested that a value of 50% would be a reasonably optimistic assumption. He also suggests a neutron utilization of e. = 0.8. The number of radioactive nuclei produced per proton results from primary spallation nuclei and fission-fragmentation products. With 0.3 fissions/proton and 3 fragments fission plus 0.5-1 heavy nuclei/proton it would appear that 1.6 radioactive nuclei/proton in the target region might be a reasonable assumption. Thus an upper estimate is obtained as E 1000/0.5 - 1in M W/CD /noc x N 1 (24.6-1.6)0.8 = 110 MeV/FP (A36) 9. A. 32 This value is not considered as an absolute bound but rather a practical bound. If this performance could not be attained the accelerator probably would not be built. Thus the energy required is estimated to be 23 F p eV < | < 110 MeV/FP (A37) This required electrical energy must be less than the electrical energy pro- duced in a fission power reactor per candidate transmutation fission product. This energy balance is strictly true only as long as electrical power is being obtained from fission reactors. The electrical energy produced per FP can be obtained from the production rates given on the Chart of the Nuclides. These values are estimated" for a 1000 MW(e)-yr, slightly enriched uranium, 25,000 MWD/T fuel exposure with 85% duty cycle. The values given for candidate FP Cs-137 41.6 Kg Sr-90 18 Kg Tc-99 27.5 Kg correspond to the following values of electrical energy produced per FP nucleus: Cs-137 900 MeV/FP Sr-90 1370 MeV/FP Cs-137+Sr-90 550 MeV/FP Tc-99 985 MeV/FP Cs-137+Sr-90+Tc-99 350 MeV/FP 9- A. 33 Thus, the energy input to the ING required to transmute only these candidate FP seems to be clearly less than the electrical energy produced in their formation. Quantity of FP Transmuted The energy balance derived above can also be used to estimate the quan- tity of fission product which can be transmuted in this conceptual system. An (13) ORNL study group proposed v ' that 1000 MW of power to spallation accelerators would be required to handle the Sr-90 produced in 9000 MWe of LMFBR's. This conclusion is only in approximate agreement with the values derived here for LWR's. For the pessimistic assumption, i.e., 110 MeV/FP, and the transmutation of Cs-137 + Sr-90 + Tc-99 the ratio of spallation power to fission electrical power is one-third. A more optimistic assumption would be about half of that. From an economic viewpoint, however, a power ratio of 1/6 or 1/9 may still be prohibitive. The ING proposal was for an accelerator of only 65 MW proton beam power. Thus fifteen ING's would be required for 6-9 LWR's and the 1967 projected costs of an ING were (18) $182 M without conversion of thermal energy to electrical power. ' Thus projected large improvements in beam power in a single accelerator would appear to be necessary for economic feasibility. Specific Transmutation Rate It has been shown that the energy balance for transmutation of Cs-137, Sr-90 and Tc-99 in an ING spallation accelerator could be favorable. In order for a transmutation scheme to be attractive the induced transmutation rate must 9. A. 34 significantly exceed the radioactive decay rate. These values are compared on 1 f\ ? l Table 9. A. 4 for the projected maximum thermal flux values of 10 n-cm~ -s~ of the ING. The reduction in half live for Cs-137 is seen to L»e decreased not \/ery significantly even for the infinite dilute thermal flux value. 15 -2 -1 In addition, if the effective flux is reduced to the order of 10 n-cm -s due to neutron self shielding in the FP sample or other factors the transmutation rate of Sr-90 becomes equally less attractive. Projected thermal flux values several times larger than 10 n-cm~ -sec" would be required to make the spallation accelerator genuinely attractive for transmutation of Sr-90 and Cs-137 FP isotopes. Radioactive Waste Balance In the general discussion of the spallation accelerator some of the radio- active waste produced in the spallation of 1 GeV protons on Pb-Bi was identified. These included the production of nearby nuclei of the type: Nucleus T l/2 Bi-208 3. 68x1 5 yr Bi-207 38 yr Bi-206 6.24 d Pb-205 1.4x10 yr Pb-203 52 hr Tl-204 3.78 yr Tl-202 12.2 d 9. A. 35 TABLE 9. A. 4 . Time For 99% Elimination in Thermal Fluxes With Frequent Rapid Daughter Product Removal Time for 99% Elimination, yrs Thermal Flux -2 -1 (n-cm -sec ) Thermal Capture (barns) Decay Only ft = 10 15 * = 10 16 Sr-90 1 191 83. 14. Tc-99 22 1.40xl0 6 6.6 0.66 Cs-137 0.11 199 170. 80. Assumed cross sections in barns: Sr-90 - 1 Tc-99 - 22 Cs-137 - .11 9. A. 36 The production of nuclei of this type was estimated to sum to 0.5-1 per incident proton with only a small probability for the formation of a stable nucleus. The fission plus fragmentation of the cascade nucleus was estimated to occur 0.3 times per incident proton at 1 GeV. Each such event produces about three product nuclei on the average, the probability of formation of a fragment nucleus of mass A being independent of A within a factor of several. Some specific radioactive fragments which have been identified are shown on Table 9. A. 5 along with its yield, if known. Since about one nucleus of this type will be formed per incident proton it is apparent that the short and moderate half life activity in the target environs will be orders of magnitude larger than the disintegration rates of Sr-90 and Cs-137. This observation would seem to rule out any consideration of the use of spallation accelerators for the transmutation of radioactive waste with half-life values less than a few years. If 24.6 source neutrons are produced per proton and the utilization for fission product absorptions is 80% as previously assumed then 5 neutrons/proton are absorbed in Be, D, and structural material. These absorptions result in Be-9 + n - 1.6xl0 6 y Be-10 H-2 + n + 12.33y H-3 In addition, tritium will be produced in appreciable quantity by fast neutron reactions f- 8ms f- Be-4(n,a) b He -+ \i(n,a)T 9. A. 37 TABLE 9. A. 5 . Radioactive Product Nuclei Identified in the Spallation of 1 GeV Protons Lead (16) Nucleus Half-Life and Decay Production Cross Section (mb) F-18 11.1s 0.04 Na-24 15 h 0.4 Mg-28 21 h -> 2.25 m Al-28 0.08 P-32 14.3 d 0.09 Sr-91 9.48 h -> 58.6 d Y-91 Mo-99 66 h ■> 2xl0 5 y Tc-99 2.5 Cd-115m 44.6 d 0.5 Cd-115 53.5 h 0.3 Ba-128 2.42 d •* 3.8 m Cs-128 0.5 Ba-129 2.1 h ■* 32.3 h Cs-129 0.3 Ba-131 11.7 d - 9.69 d Cs-131 0.9 Ba-133m 38.9 h ■> 10.4 g Ba-133 9. A. 38 and Be-9(n,T)Li-7 Thus it appears that some 6.6 radioactive nuclei with a wide range of half-lives may be former per 19.6 fission product transmutation. Although detailed calculations do not exist to support these rather uncertain values, it does not appear that the projected reduction of about 3 to 1 in radioactive nuclei is worth detailed consideration. (12} Other Proposals - Steinberg v ' has proposed the transmutation of fis- sion products in a spallation accelerator of about ten times the beam power of the ING. This would improve the accelerator capital investment and the (13) maximum thermal flux value. As the ORNL study group pointed out v this type of accelerator is not feasible within the limits of current technology. In addition, increasing beam voltage or current has no effect on the considerations given the ING with regard to energy balance or radioactivity produced per proton. (14) Dr. Torao Ichimiya v ' has proposed the use of Cs-137 target material for a ten BeV proton beam. This evades the problem of the low thermal neutron capture cross section of Cs-137 by obtaining additional transmutation from high energy n,2n reactions. The transmutation rate for this procedure cannot be deouced from physics information at hand, but Ichimiya and associates assume 100% utilization of secondaries to estimate 85 nuclei transmuted per 9. A. 39 incident proton. At 100% accelerator efficiency, this would require almost 6 MeV per reactor fission to transmute the Cs-137 alone. Observed is: (1) this is energetically expensive, (2) the physics calculation is very uncertain, and (3) the radioactive waste production is certain to be large. It is concluded that the spallation device is ineffective as a short term alternative in reducing the inventory in curies of radioactive materials. Its ultimate merit lies in eliminating materials whose hazard persists longer than that of human historical records. The objective is to trade the long term (10 year) hazards of the heavy element wastes for the shorter term (10 year) hazards of fission products. 9. A. 40 REFERENCES 1. L. C. North'cliffe and R. F. Schilling, "Range and Stopping Power for Heavy Ions," Nuclear Data Tables , vol. 7, no. 3-4, p. 233, 1970. 2. J. D. Newton, "Coulomb Excitation," Nuclear Structure and Electromagnetic Interactions , Scottish Universities, Summer School, 1964, N. McDonald, ed., Plenum Press, N.Y., 1965. 3. K. Alder, A. Bohr, T. Huus, B. Mottelson and A. Winther, Reviews of Modern Physics , vol. 28, no. 432, 1956. 4. W. C. Barber and W. D. George, "Neutron Yields from Targets Bombarded by Electrons," Physical Review , vol. 116, no. 9, p. 1551, 1959. 5. R. G. Alsmiller, Jr., and H. S. Moran, "Photoneutron Production from 34- and 100-MeV Electrons in Thick Uranium Targets," Nuclear Instruments and Methods, vol. 51, p. 339-340, 1967. 6. G. C. Baldwin, "Neutron Production by Electron Bombardment of Uranium," Physical Review , vol. 104, no. 6, p. 1652, 1956. 7. Robley D. Evans, "Gamma Rays," American Institute of Physics Handbook , Dwight E. Gray, Ed., McGraw-Hill, NY, 1963. 8. Walt Vali, Private Communication. 9. J. H. Teshune and G. C. Baldwin, Physics Review Letters , vol. 14, no. 15, p. 589, 1965. 10. G. A. Bartholomew, "Spallation-Type Thermal Neutron Sources," Seminar on Intense Neutron Sources," p. 637 of Conference Proceedings, TID-4500, C0NF-660925, September 19-23, 1966. 11. M. Steinberg, G. Wotzak, and B. Manowitz, Neutron Burning of Long Lived Fission Products for Waste Disposal , BNL-8558, Brookhaven National Laboratory, 1964. 12. Michael V. Gregory and Meyer Steinberg, A Nuclear Transformation System for Disposal of Long-Lived Fission Product Waste in an Expanding Nuclear Power Economy , BNL-11915, Brookhaven National Laboratory, November 1967. 13. F. L. Culler, Deputy Director, Oak Ridge National Laboratory, Letter to V. M. Staebler, USAEC, December 28, 1971. 9. A. 41 14. Torao Ichimiya, Director of Institute of Physical and Chemical Research, Japan, Private Communication, 1972. 15. R. R. Fullwood, J. D. Cramer, R. A. Haarman, R. P. Forest, Jr., and R. G. Schrandt, Neutron Production by Medium-Energy Protons on Heavy Metal Targets , LA-4798, January 1972. 16. E. K. Hyde, "The Nuclear Properties of the Heavy Elements III, Fission Phenomena," Prentice-Hall, Englewood Cliffs, NJ, 1964. 17. V. A. Davidenko, "On Electronuclear Breeding," Sov. At. En. , vol. 20, p. 866, (Engl.) September 1970. 18. Table 3.1, AECL-2750, 1967. APPENDIX 9.B TRANSMUTATION BY FISSION AND THERMONUCLEAR EXPLOSIVE DEVICES Page SUMMARY 9.B.1 DETAILED RESULTS 9.B.2 REFERENCES 9.B.13 9.B.1 APPENDIX 9.B TRANSMUTATION BY FISSION AND THERMONUCLEAR EXPLOSIVE DEVICES SUMMARY Because thermonuclear explosive devices are known to produce large yields of neutrons, the use of these neutrons for the transmutation of (1-3) radioactive waste has been proposed . In a preliminary evaluation, PNL concluded that the concept did not appear to be technically feasible. A group of reviewers at Los Alamos Scientific Laboratory (LASL), however, reviewed the PNL study and strongly disagreed with its conclusions and concluded that the concept may be very attractive. The details of the LASL review are contained in Appendix 9.E. Because of the LASL conclusions, PNL has conducted a more extensive evaluation of this concept and this evaluation is reported in the Detailed Results of this Section. The PNL evaluation does not support the conclusions reached in the LASL review. The PNL evaluation differs from the LASL primarily in the following respects: • The LASL estimation of the yield of fission products Tc-99, 1-129, and Cs-137 is too low by a factor of about three. • The LASL assumption that successive neutron captures in product isotopes does not occur does not appear to PNL to be physically justified. Successive neutron captures lowers the efficiency of utilization of source neutrons. • Successive neutron captures further lowers the efficiency of the use of source neutrons due to the presence of other iodine and cesium isotopes. 9.B.2 The main conclusions of PNL of the feasibility of the concept are: • The concept can be considered only for radioactive waste with half- lives much greater than the 12.3 year tritium because of the significant production of tritium residue in the device. Hence, the transmutation of short-lived radioactive waste is not technically feasible. • The estimated cost of transmutation of elemental fission product technicium, iodine and cesium is almost twice the cost of the electricity produced in their production. Hence, this concept is not technically feasible. • If isotopically separated 1-129 and Cs-135 could be obtained, the tech- nical feasibility would be within the range of uncertainty of the present analysis. More rigorous calculations of the neutron transmutation process would then be required to establish technical feasibility. • The estimated cost of transmutation of Np, Am, and Cm waste is estimated to be less than 20% of the electrical cost and is, thus, apparently technically feasible. This concept is, however, judged to be much less attractive than that of actinide recycle in fission reactors. DETAILED RESULTS The thermonuclear explosive device is a potential candidate for neu- tron-induced transmutation of high-level radioactive waste in an under- ground explosion because it is a source of large numbers of available neutrons. Details of many aspects of thermonuclear devices are not avail- able for public use because of classification for weapons purposes. How- ever, the general features of such devices which are necessary to assess their 9.B.3 possible uses for waste transmutation appear to be publicly available. (4) These features are given, for example, in a book by Teller v ' and in many published papers given at the January 1970 Symposium on Engineering (5) with Nuclear Explosives^ . The important features of the device can be categorized as cost, radioactive waste produced and radioactive waste consumed. The characteristics of these features are summarized' in the following discussions. The thermonuclear explosive device consists of a small amount of fissionable material and the bulk of the material is that which is used to develop the (DT) fusion reaction. The quantity of fissile material in the explosive device apparently does not depend greatly on the explosive yield of the device. Most of the cost of the device is apparently due to the manufacturing process, rather than material costs. The costs of fabrication, emplacement, arming and firing the device appear to be in the range of $0.5 million to $0.6 million for devices in the range of 100 kiloton to 400 kiloton' 6 ' 7 ' although Cohen, et al/ 8 ^ used a value of $1 million for a smaller yield device. The cost of drilling and casing emplacement holes has been taken ' to be $33 per meter for holes 150 centimeters in diameter for depths up to 1524 meters, but Cohen, (8) et al , used a cost of $1 million for an emplacement hole 610 meters. (9) The LASL review group proposed that a 100 kiloton device at a depth of 1500 meters was appropriate for the transmutation concept. These characteristics were apparently chosen in order that geographically adjacent shots be seismically decoupled. The LASL proposal estimated a total cost per shot of $1 million assuming that the emplacement hole could be used several times. Heckman^ ' also proposed reuse of a hole 9.B.4 but it is apparently not a proven technology. Heckman also proposed a cost of $200,000 for a neutron target of uranium and thorium. The LASL proposed cost apparently does not include the costs of preparing the high-level radioactive waste targets. For the purposes of this analysis, a cost of $1.5 ± 0.5 million will be assumed to cover target fabrication costs and their uncertainty and the uncertain possibility of reuse of the emplacement hole and its costs. This estimated cost figure implies that the transmutation explosions are done on a routine basis and at the Nevada Test Site since estimates of experimental shots at other locations are of the order of $4 million. ' The radioactive waste created in a thermonuclear explosion results from fission products from the fission component of the device, from tritium which is created but not burned in the thermonuclear explosion, and from neutron-induced reactions in the device and in the surrounding earth medium. Estimates of these waste components are shown in Table 9.B.1 in units of moles of product. Also shown are the moles of neutrons created which are, in principle, available for transmutation. These yield estimates are primarily obtained from those of Evans and Kruger . The neutron leakage to the surrounding media was calculated by Lessler for a 100-kiloton device using a 150-centimeter diameter hole and boric acid shielding to inhibit neutron loss to the surrounding media. The values shown on Table 9.B.1 allow the following general conclusions. • The fission product yield is a negligible fraction of the neutrons available for transmutation except for devices of yery small explosive yield. The use of small yield devices is also ruled out on cost balance grounds. 9.B.5 • The radioactivity resulting from neutron leakage to the surrounding media can be made a negligible fraction of the neutrons available for transmutation. • Since about 1 gram mole of 12.3 year tritium is produced per 4 moles of neutrons available for transmutation, the concept of thermonuclear explosives for transmutation is feasible only for radioactive waste with half-lives greater than some 50 years. TABLE 9. B.I . Yield and Radioactive Products of Thermonuclear Explosives Yield (gram moles) Neutrons 2.5 /kt Fission Product T l/2 = 15 " 50 y r °- 15 /device T l/2 > 50 yr °* 15 / device Unburned Tritium (T ]/2 = 12.3 year) 0.67 /kt Radioactive waste in surrounding media <0.001/kt To assess the feasibility of transmutation by thermonuclear explosives, it is necessary to consider the types and quantity of long-lived radioactive waste which are candidates for transmutation. Among the fission products these appear to be Tc-99(T 1/2 = 2.1 x 10 5 years), I-129(T, /2 = 1.6 x 10 7 years) and Cs-135(T, /2 = ^ x ^ years). The base case is taken to be the product of a 1000 MW(e) PWR for fuel processed 90 days after discharge. The calculated yields of this base case for these isotopes and their chemical elements in an equilibrium fuel cycle are shown on Table 9.B.2. The impact of stable and short-lived isotopes of the element is seen to be most important for cesium. In this case, the number of atoms which would have to be transmuted is increased 9.B.6 by almost an order of magnitud2, and, because of stable Cs-133, even allowing 100 years of radioactive decay reduces the amount of material to be transmuted by less than a factor of two. TABLE 9.B.2 . Long-Lived Radioactive Waste Produced by Reference 1000 MW(e) PWR Yield (gram moles) at Cooling Time Element Isotope 1 yr. 100 yrs. Iodine Total 72 72 127 11 11 129 61 61 Technetium 99 290 290 Cesium Total 678 373 133 260 260 134 34.6 135 82.2 82.2 137 302.8 30.8 Total Fission Product 1040 735 Neptunium 237 70 (a) Americium Total 23 (a) Curium Total 5 (a) Total Np, Am, Cm 98 a. Not calculated 9.B.7 For neutron- induced transmutation of the fission-product nuclei, the assumption was made by the LASL review group that (successive) neutron capture events can be made the dominant process by proper design of the device and target configuration. There are not detailed calculations to support this assumption and it is necessary to point out that the requirements are different for the different fission product elements. For cesium, the multiple capture rate must be sufficiently high that essentially all of the Cs-133 and Cs-134 are transmuted beyond Cs-135. For 1-129, the multiple capture rate in 127 and 129 must be adjusted so that no significant amount of mass 135 is produced. Whether either of these restrictions can be achieved in the face of competing n,2n reactions needs more definitive analysis than has yet been made. The need for multiple neutron capture in the chemical fission product elements iodine and cesium greatly reduces the number of thermonuclear neutrons which can effectively cause a transmutation of the desired isotope. Ignoring the complexity of competing n,2n reactions, if each isotope has the same value of effective capture cross section, a, and sees the same effective neutron flux, , the distribution of atomic masses, (12) Ao + n, after adding n neutrons to the initial Ao nucleus is v ' ( Ao + n) = Ao Ly- e (Bl) where r = a <}> is the capture rate. Values of r in the range of 1 to 25 are clearly available from the device. For an r value of 5, 8.4% of the initial Cs-133 atoms would transmute to Cs-135 (or 1-127 to 1-129) and the average number of available neutrons used per transmutation is about 4.3. For an r value of 7 the percentage of Ao + 2 transmutations is 2.2% and 9.B.8 the average number of neutrons used per transmutation is about 6.3. Thus, for a reasonable reduction in the radioactive waste only about 20% of the available source neutrons can be used for transmutation. It is also clear from the rate equation (Bl) that the neutron flux cannot decrease very much through the transmutation target. Reduction of the flux (r) from a value of 7 to a value of 5 increased the production of the undesirable Ao + 2 isotope to a marginally feasible level. In order to keep the flux level to the required value in the transmutation target, the order of 1/2 of the source neutrons must escape to the device shield to be absorbed. Coupled with the estimate of the order of 5 neutrons absorbed for an effective transmutation, the overall efficiency of useage of source neutrons per transmutation appears to be no better than some 10% for FP iodine and cesium. Also shown on Table 9.B.2 is the estimated production of the long-lived actinides neptunium, americium and curium. The transmutation concept for these elements is the reduction to fission product waste by successive captures and fission which occurs primarily at odd mass plus a neutron value. The effective reduction of neptunium to fission product differs from the iodine and cesium cases in that the average number of source neutrons required to cause a fission is probably larger than those required for I and Cs transmutation. However, the fission events in the actinide targets will cause source neutron multiplication so that the overall efficiency for transmutation may not- be significantly different. The efficiencies of utilization of source neutrons proposed here are quite different than those assumed by the LASL review group. The LASL group 9.B.9 assumed that only single neutron captures could be achieved with an overall utilization of 80 percent. Such a process cannot be achieved with eq.(Bl) where each isotope has an equal neutron capture cross section. In order to achieve the LASL assumption, the neutron capture cross section of the Ao + 1 isotope would have to be much less than that of the Ao isotope. The opposite situation exists for Cs-133 and almost certainly for 1-127. The technical feasibility of transmutation by thermonuclear explosives requires a comparison of the cost of transmutation to the cost of the elec- tricity produced in the production of the radioactive waste. For this com- parison, the reference 1000 MW(e) PWR produces an assumed $35 million worth of electricity per year based on 5 mills/kW-hr and a plant factor of 0.8. The neutron yield of a 100-kiloton device is 250 gram moles at an estimated cost of $1.5 million. At the 10% efficiency estimated for transmutation, the cost of transmutation is estimated to be of the order of 25 moles ~ $ 6 0>°00 P er 9 ram mole of waste. (B2) The costs of transmutation of the various waste components of the reference reactor are then estimated as on Table 9.B.3 and expressed as a percentage of the value of the electricity produced in their production. Also shown is the annual number of explosions required to keep abreast of the production of one 1000 MW(e) PWR and the annual number of explosions required for a 150,000 MW(e) fission power reactor economy which is postulated by about 1980. 9.B.10 TABLE 9.B.3 . Estimated Costs of Transmutation by Thermonuclear Explosives Element Annual Cost per PWR $ Million Percent of Electri Power Costs cal Annual No. of 100 kt Explosive Shots per 1000 MW(e) per 150,000 MW(e) I + Cs 45 129% 26 3900 Tc 18 50% 11 1650 Np+Am+Cm 5.9 17% 3.5 525 From the estimated cost values given in Table 9.B.3, we conclude that transmutation of the elemental fission products iodine, cesium and technicium by thermonuclear explosives is not technically feasible. If it were possible to obtain isotopically separated 1-129 and Cs-135 at not too great a cost, then the question of technical feasibility would fall within the uncertainty of the. present analysis. The transmutation of the actinide waste, Np, Am, and Cm does appear to be technically feasible for the criteria of quantity, rate, and energy balance. The concept does not appear attractive, however, based on the large number of continuing explosions which would be required to keep abreast of even a static fission power reactor economy. The use of thermonuclear explosives to transmute the actinides to fission product is judged to be a much less desirable process than that of recycle in fission reactors. The PNL analysis of estimated costs for transmutation of fission products by thermonuclear explosives differs greatly from that of the LASL review group who concluded the concept "may be very promising". The differences in the assumptions of the two analyses are shown on Table 9.B.4. The differences in estimation of cost per explosion plus the amount of fission product Tc-99, 1-129 and Cs-135 amount to a factor of about five. The lar- 9.B.11 gest part of this difference is due to the fact that the LASL yield was erro- neously estimated for a 1000 MW thermal reactor rather than the reference 1000 MW(e) PWR for which the PNL yields are based on ORIGEN calculations. The largest difference in the estimations is due to the LASL assumption of only single neutron captures. According to the PNL analysis, an average of about four successive captures is a necessary consequence of efficient removal of only a single isotope. This conclusion lowers the efficiency further due to the presence of other fission product iodine and cesium isotopes. TABLE 9.B.4 . Comparison of PNL and LASL Estimations of Transmutation of Fission Product by Thermonuclear Explosive Devices Element LASL PNL Comment Cost/Explosion Transmutations Explosion Moles of Fis- sion Product PWR $1 Million $1.5 million 200 gram moles 130 gram moles 25 gram moles 1040 gram moles The PNL estimate contains costs of target fabrication apparently not included in the LASL estimate, LASL assumed no multiple neutron capture. The PNL analysis indi- , cates that multiple neutron cap- tures are necessary for efficient transmutation of even separated isotopes. LASL neglected the 1-127 and the fission product Cs other than Cs-135. This assumption involves the assumption of single neutron captures which the present analysis indicates is not feasible. The LASL calculation of fission prod- uct yield was also erroneously based on 1000 MW(th) rather than 1000 MW(e). 9.B.12 Dr. Edward Teller, Associate Director of Physics of Lawrence Livermore (13) Laboratory, was also questioned^ ; by PNL on the results of any studies by LLL on the thermonuclear explosives transmutation concept. Dr. Teller's (14) reply v ' did not indicate that the concept had received technical evaluation at LLL but he characterized the concept as "complicated and expensive" and less attractive than burial in salt domes or in chimneys of silicate rock created by nuclear explosives. 9.B.13 REFERENCES 1. M. Goldstein and E. Molting, Proposal No. IBR-72-2706, International Business and Research, Inc., Proposal to USAEC dated January 24, 1972, Revised February 11, 1972. 2. A. E. Bolton and D. R. Edwards, Eliminating Radioactive Wastes by Underground Thermonuclear Explosions , Research Proposal to USAEC, University of Missouri, May 1972. 3. N. F. Colby, Use of Isotopes to Reduce Neutron-Induced Radioactivity and Augment Thermal Quality of the Environment of an Underground Explosion , M.S. Thesis, University of Missouri, May 1972. 4. E. Teller, et al, The Constructive Use of Nuclear Explosives , McGraw- Hill, Inc., N.Y., 1968. 5. "Symposium on Engineering with Nuclear Explosives", January 14-16, 1970, Las Vegas, NV, C0NF-700101, 2 vols., May 1970. 6. Ray B. Evans and Paul Kruger, "The Use of a Rubble Chimney for Denitrification of Irrigation Return Waters", pp. 1222-1245, vol. II, C0NF-700101 (Ref. 5), May 1970. 7. Richard A. Heckman, "Thermonuclear Neutron Sources--A New Isotope Production Technology", pp. 1295-1305, vol. II, C0NF-700101 (Ref. 5), May 1970. 8. J. J. Cohen, A. E. Lewis and R. L. Braun, Use of a Deep Nuclear Chimney for the In-Situ Incorporation of Nuclear Fuel-Reprocessing Waste in Molten Silicate Rock , UCRL-51044, May 4, 1971. 9. D. Graham Foster, Jr., et al, "Review of the PNL Study on Transmutation Processing of High-Level Waste", LA-UR-74-74, Jan. 14, 1974, (Appendix E of this report). 10. A. E. Lewis, "Chemical Mining of Primary Copper Ores by Use of Nuclear Technology", pp. 909-917, vol. II, C0NF-700101 (Ref. 5), May 1970. 11. Richard M. Lessler, "Reduction of Radioactivity Produced by Nuclear Explosives", pp. 1563-1568, vol. II, C0NF-700101 (Ref. 5), May 1970. 12. Samuel F. Eccles, "Production of Heavy Nuclides in Nuclear Devices", pp. 1269-1282, vol. II, C0NF-700101 (Ref. 5), May 1970. 13. "Disposal of Radioactive Waste by Thermonuclear Explosion", Letter from A. M. Piatt, Battel le-Northwest, to Dr. Edward Teller, Lawrence Livermore Laboratory, dated October 17, 1972. 14. Letter from Dr. Edward Teller, Lawrence Livermore Laboratory, to A. M. Piatt, Battel le-Northwest, dated December 5, 1972. APPENDIX 9.C TRANSMUTATION BY FISSION REACTORS SUMMARY DETAILED RESULTS A. Fission Products B. Actinides - Review of Claiborne's Work - Kudo's Analysis - PNL Studies REFERENCES f >age 9. X.1 9, ,C.l 9 .C.l 9. .C.4 9 .C.5 9 .C.8 9 .C.ll 9 .C.22 9.C.1 BNWL-1900 APPENDIX 9.C TRANSMUTATION BY FISSION REACTORS SUMMARY Fission reactors produce neutrons which, in principle, can be used to trans- mute waste. The earliest proposal for the use of fission reactor neutrons to transmute waste was made by Steinberg et al. ' In this proposal, the fission products krypton-85, stronti um-90, and cesium-137 were considered candidates for transmutation in a special purpose high flux reactor. Studies conducted ( 2 ) at Oak Ridge National Laboratory (ORNL) reported by Claiborne v ' considered neutron-induced transmutation of fission products and actinide elements. Clai- borne's work on transmutation of actinides in commercial LWRs is the most ex- tensive study of the subject to date. Calculations of actinide recycle in a commercial PWR were made at PNL to compare results with Claiborne's study and extend the analysis to include other strategies. In reviewing the work of others and the PNL studies, it is concluded: 1. Transmutation of fission products in fission reactors is not technically feasible because it fails to meet the waste inventory and transmutation rate criteria. 2. Transmutation of actinides in fission reactors is technically feasible. Reductions of approximately two orders of magnitude in the long-term hazard potential of actinides in high-level waste can be realized by recycling actin- ides in fission reactors. The transmutation can effectively be accommodated in either uranium fuels (UO-) or plutonium fuels (U02-Pu0 2 ), and perhaps even in specially fabricated target rods selectively located in the reactors. Though existing calcul ati onal studies have been made only for a commercial pressurized water reactor (PWR), perhaps even better transmutation results are expected for other fission reactors such as High Temperature Gas Cooled Reac- tors (HTGRs), and Fast Breeder Reactors (FBRs). DETAILED RESULTS Fission reactors can, at least in principle, be applied in fission product transmutation and actinide transmutation. The technical feasibility of each is described below. A. Fission Products Steinberg, Wotzak, and Manowitz, made the initial study* 1 ' of fission prod- uct transmutation in fission reactors. Their study was concerned with the possibility of transmutation of the fission products Kr-85, Sr-90 and Cs-137. In their analysis, they considered the competition for neutrons of other fis- sion product isotopes of the same FP chemical elements in order to determine if isotopic separation was required. In this analysis, they used thermal 9.C.2 BNWL-1900 neutron cross sections. Hence, their study implicitly assumed that the trans- mutation reactor was a thermal reactor. They concluded that isotopic separa- tion was necessary for Kr-85 and Cs-137 but not for Sr-90. The enrichment level which they proposed for Kr-85 (90%) is no longer valid because they assumed that the neutron capture cross section of Kr-85 was 15 barns. It is now known that the value is about 1.7 barns (see Table 9.C.1) so that much greater enrichment is required than they proposed. Their calculated transmuta- tion rates for Kr-85 are also an order of magnitude too large because of their assumed cross section. The transmutation in thermal fission power reactors of troublesome fission product isotopes with small values of thermal capture (transmutation) cross sections is not possible because of the low values of thermal neutron flux in such reactors. Calculated values of transmutation rates of several such iso- topes are shown on Table 9.C.1 for various neutron flux levels. These results show that, with the exception of 1-129, the transmutation rates differ insig- 1 4 mficantly from the radioactive decay rates for neutron flux levels of 10 neutrons/ (cm ) (sec ) . The neutron flux levels in thermal fission power reac- 13 2 tors are the order of 3 x 10 neutrons/(cm )(sec). Thus, the transmutation of low cross-section isotopes in thermal power reactors does not satisfy the criterion of specific transmutation rate. Steinberg et al. ' specifically proposed the use of special high-flux thermal reactors, such as a flux trap reactor, for transmutation of fission products. Such reactors achieve large values of thermal neutron flux, of the 16 ? order of 10 neutrons/(cm )(sec), by thermalizing leakage neutrons in a medium which essentially does not absorb neutrons. Such a thermal neutron flux level gives specific transmutation rates which would be marginally feasible for some isotopes as shown on Table 9.C.I. However, this neutron flux level cannot be maintained in the presence of absorbing material. At best only a few percent of the fission neutrons could be utilized to produce an interesting specific transmutation rate. In addition, an equal or greater number of fission prod- uct isotopes of high-level waste would be formed in the fission process per transmutation event. Thus, this reactor concept does not satisfy the criteria of overall waste balance and of total transmutation rate. 1) Steinberg et al did not consider the use of a fast breeder reactor for transmutation of fission product isotopes although LMFBRs were considered in ( 2 1 studies at 0RNL. v ' The consideration of the use of LMFBRs for fission prod- uct transmutation arises primarily from the excess neutrons per fission which are destined for the breeding of fissile material. In practical LMFBR designs the excess fraction of neutrons per fission is in the range 0.15 to 0.3. These values are approximately the same as the fraction per fission of fission prod- uct constituent of high-level waste which the LMFBR would produce during opera- tion. Thus, the LMFBR could transmute fission product at only about a break even rate on overall waste balance and would do so at the expense of being no longer a viable breeder of fissile material. In addition, the projected fast 9.C.3 BNWL-1 900 TABLE 9. C . 1 . Properties of Several Important Fission Product Nuclides and Time Required for 99.9% Reduction of Their Inventory by Decay and Neutron Transmutation (Taken from Reference 2) Nuclide . Sr-90 Cs-137 K.r-85 H-3 1-129 Half-1 ife, years f \ Burnout cross section, barns Curies/metric ton in spent fuel (c) Relative hazard in spent fuer 3 m air at RCG/metric ton 3 m water at RCG/metric ton Time required for 99.9% ij\ decay and burnout, years Decay only 14 2 (el * = 10 n/cm -sec* 1 ' $ = 10 15 n/cm 2 -sec^ e ' 4 = 10 16 n/cm 2 -sec^ 5 _ ]0 17 ri/ „2. sec (e) a. Effective thermal cross section in typical spectrum of a PWR having average thermal flux of 2.91 x 10 13 n/cm**sec. b. Per metric ton of uranium charged to a PWR having average specific power of 30 MW/metric ton and burnup of 33,000 MWd/metric ton. c. Volume of air and water potentially contaminated to RCG (10 CFR 20) by the content of a metric ton of spent fuel . , „ d. Indicated times are doubled and tripled for reduction of inventory by factors of 10 and 1C , respectively. e. Average thermal flux assuming spectrum typical of that in a PWR. 28.9 30.2 10.74 12.33 1.6 x 10 7 1.2 0.17 1.8 nil 35 77.C0C I 108,000 11,400 708 0.0367 2.6 x 10 15 2.1 x 10 14 3.8 x 10 10 3.5 x 10 9 1.8 x 10 9 2.6 y 10" 5.4 x 10 9 — 2.3 6.1 x 10 5 288 302 107 123 1.6 x 10' 249 295 106 123 63 112 245 98 123 6.3 17 91 57 123 0.63 1 .8 12 n 123 0.06 1 5 2 neutron flux levels of about 10 n/cm /sec do not allow the attainment of a sufficiently high specific transmutation rate. Thus, the use of an LMFBR to transmute the fission product component of high-level waste is not a feasible concept. (2) Claiborne's report ' gives results of work which have been conducted at 0RNL on fission product transmutation. Tables 9.C.1 and 9.C.2 are taken from this report. The data in Table 9.C.1 illustrates the points made in the above arguments concerning the requirements of high flux and high cross section for 17 2 effective transmutation. Flux levels of the order of 10 neutron/(cm )(sec) are needed to accomplish reasonable transmutation rates. The data in Table 9.C.2 are referred to in Claiborne's report as results of work conducted at 0RNL by Nichols and Blomeke in which they studied the effect of neutron transmutation schemes on radioisotopic inventory and electrical power costs. The rate of decay is essentially unchanged in recycling Sr-90 in either PWRs or LMFBRs. Thus the inventory is unchanged. The high flux reactor was 9.C.4 BNWL-1 900 TABLE 9.C.2 . The Inventory of 90 Sr and Estimated Costs Associated With the Steady-State Production of Electric Power by Various Schemes of Transmutation (Taken from Reference 2) 90 Sr Inventory (megacuries) per 1000 MW(e) of Capacity 1. PWR - conventional operation (b) 2. PWR with complete 90 Sr recycle' ' 3. LMFBR - conventional operation^ 4. LMFBR with complete 90 Sr recycle' ' wi th Plant Factor of 0.8 Estimated Incremental In Outside Cost (a) Reactors Reactors Total [nr ills/kWhr(e)] 3.17 88.0 91.2 64.6 21.5 86.1 0.1 1.39 38.6 39.9 28.9 9.6 38.5 0.1 5. High Flux Isotope Reactor - conventional* 6. High Flux Isotope Reactor conventional vd) 0.11 132. 132. 24 90 Sr recycled 38.3 12.7 51.0 25 7. 98% of power from LMFBRs plus' ' 2% of power from fusion burner 1.36 0.93 0.91 ) 0.31 ( 3.51 8. 90% of power from LMFBRs plus' e ' 10% of power from spallation 1.54 0.16 1.031 0.51 1 3.23 a. In excess of the typical power generation cost of approximately 7 mil ls/kWhr(e). b. Assumes thermal efficiency of 32.5%. c. Assumes thermal efficiency of 41.7% d. Assumes thermal efficiency of 30%. e. One 1000-MW(e) spallation burner reactor associated with nine 1000-MW(e) LMFBRs. The electricity generated by the burner reactor is used internally to generate a 500-Mw beam of 10-BeV protons. 1 5 assumed to produce a minimum thermal flux value of about 2 x 10 neutrons/ 2 (2) (cm )(sec). ; At this flux level, the inventory is reduced by roughly 60%. Claiborne concluded that "this type of reactor would not be an economical source of electric power, however, because of its small size, high refueling cost, and high neutron leakage." Lines 7 and 8 to Table 9.C.2 are strat- egies combining LMFBRs with CTRs, and LMFBRs with spallation accelerators. In summary, it is improbable that transmutation of fission products in fis- sion reactors could meet any of the technical feasibility requirements for the production of stable daughters. An independent evaluation of the feasibility (3) of fission product transmutation was also made by Kubo. ' He concluded that "fission products are not conducive to nuclear transmutation as a general solution to long-term waste management." B. A c t i n i d e s Actinide transmutation with LWRs appears to be quite attractive based on (2 ) the preliminary analyses made to date. Claiborne's study v has produced the (3) (4) most extensive findings on this subject to date. Kubo v ' and Kubo and Rose 9.C.5 BNWL-1900 at MIT have studied Claiborne's results and similarly have concluded that actinide recycle in fission reactors is not only technically feasible but an attractive waste management concept. Their extension to Claiborne's work has been primarily devoted to analyses of the further reduction in potential toxi- city which would accrue from improved efficiency for removal of actinides higher than plutonium from the high-level waste. The principal findings of Claiborne's study are reviewed here. The extension of Claiborne's analysis made by Kubo is briefly summarized. The results of calculations made at PNL confirming Claiborne's conclusions and extending the concept to include other potential strategies are also presented. Review of Claiborne's Work - The basis of Claiborne's study is depicted by the system flowsheet shown on Figure 9.C.I. In this system it is assumed that reprocessing of the fuel takes place 150 days after reactor discharge. The actinide elements present at that time are then recycled into fresh fuel. The recycle actinide material was simply added uniformly to every rod of a 3.3% enriched UO^ fuel loading for a PWR. The U and Pu are kept out of the recycle stream. ORNL-DWG 72-1436 (Taken from Ref. 2) r 3.3% ENRICHED U FEED FUEL ASSEMBLY FABRICATION FEED + RECYCLED -o HEAVY ELEMENTS 99.5 TO 99.9% HE AVY ELEMENTS OTHER THAN U AND Pu FISSION PRODUCTS 0.1 TO 0.5% HEAVY FLEMENTS 100% DAUGHTERS WASTE STORAGE R.EACTOR DISCHARGE 1 CHtMlCAL PROCESSING 99.5 TO 99.9% U + Pu 1 U + Pu STORAGE FIGURE 9 . C . 1 . Flowsheet for Actinide Recycling 9.C.6 BNWL-1 900 The effect of recycle on the inventory of actinides is seen by examination of Table 9.C.3, which lists the amounts of the various actinide isotopes found in a ton of burned fuel. (One recycle covers three years of reactor operation, with one-third of a PWR core discharged each year.) The inventory of most of the heavy elements (e.g., Np, Am, Cm) is seen to increase to saturation in about five recycles. Since at saturation these isotopes are transmuted at 99.5% of the rate at which they are produced, this represents a substantial re- duction of the inventory of toxic species of actinide waste. To put these results in perspective Claiborne used a hazard^ ' index defined as the amount (volume) of air or water required to dilute for release to the environment the residual waste, following reprocessing and removal of materials to be recycled. This index was calculated for dilution of each nuclide to its Radiation Concen- tration Guide Value (RCG). He then defined a hazard measure (or hazard reduc- tion factor) which is the ratio of two indices. The index in the numerator is for waste from fresh fuel, burned and reprocessed, with none of the actinides removed. The index in the denominator is for waste from fuel into which actin- ides have been recycled, burned and reprocessed, with 99.5% of the actinides re- moved for further recycling. Table 9.C.4 presents this hazard measure as a function of the number of re- cycles and the time of storage prior to dilution and release. For a given stor- age time the numerator index is a constant and the denominator index represents the hazard associated with 0.5% of the actinide content of the fuel listed in Table 9.C.3. Thus, the hazard measure decreases as the actinide content of the fuel increases with multiple recycles. The changing isotopic content of the fuel is reflected in the changing time behavior of the hazard reduction factor with recycling. Thus, rapid saturation under recycle of the fuel content of long-lived isotopes seen in Table 9.C.3 is reflected in Table 9.C.4 in the hazard reduction factor for release after 10 years. The increase with recycle of short-lived isotope concentrations is reflected in the hazard reduction fac- tor for release in the short term. The cumulative hazard of the actinide waste from operation of a PWR for 60 years, for cases with and without recycle, is plotted in Figure 9.C.2 for the short term and in Figure 9.C.3 for the long term. Claiborne studied the effect of extraction efficiency on the actinide recycle Hazard reduction factors for the case assuming a 99.9% extraction efficiency are shown in Table 9.C.5 for comparison with the results for the case with 99.5% efficiency presented in Table 9.C.4. The hazard reduction factors are increased We prefer to use toxicity rather than hazard since its connotation is different. However, in this section, the word hazard will be used since it was used by Claiborne in his work and because we are taking data directly from his report. 9.C.7 BNWL-1900 CO T3 -c co +-> C7> S- ^-^ c: ■— . O ro •!- c ■ — Q O) 0) S_ i— Ol QJ O T3 4- >)•■- QJ o c Q£ Ol •!- DC -t-> E O O 4- =t S- O 4- .c -!-> O C (J f0 CD Ol LU -^ CO CQ <— cr» co i i i i »X> LT) LT) LD LO U"> LT) I I *3-r-*or-»cocnoo^u3u-)OOOo <3- <3- co ro ai »— »— i— n m ro -t ^j- ^rooc\jcsjc\j»— t— t— t i i i i i i i i ooooooooo xxxxxxxxx cT>ocococornr^csj"X)Oc\ir^co <3-oc\jCT>i£>u3CT>rnatco^j"*3-«a- i-Ulini-^CO^NO^f-r-i — »— ^r ro oj c\j cm i— • — i— i — ■ — . — i — i i i i i i i i i i i i i ooooooooooooo xxxxxxxxxxxxx i£>oo«*cMomoaoocococ\j m < *r in lo ^r> p — C\JCNJCSIC\J<\JC\JCSJCsJCNJC\JC\JC\J ooooooooooooo xxxxxxxxxxxxx in ^ CO i — cocOr^u^c^jcoco^OKD 01r-lOOCMnC0«- CO ^T LT> LT> IT) co»— .— c\jc\jc\jc\jcocococococo MCSJCVJNCSJfNjCSJCSJfMCVJCSJMCSJ OOOOOOOOOOOOO xxxxxxxxxxxxx r«^ P-. i— i— . — r— . — i— »— *— i — .— r— CO CO CO CO CO CO CO CO CO CO CO CO CO OOOOOOOOOOOOO xxxxxxxxxxxxx P-* »— CTiC\iCOCO*t*T*3-COCOtOCO CXiC\JC\JCOCOCOCOCOCOCOCOCOrO co cr\ cr> cr> o~» cr> cti ct* oi cx> cr> o^ crt C\J(\JC\JC\JC\JC\JCTtc\jCDroLnLnLr)Lr)Lr)LnLn coloi — ^tLr>LOLr)LnLnLnur)Lf)LO ^\£>r^r^r^r^r^r^r^r-«.r»«.r«..r , '-«. Lr)Lnir>tr>LOLr)Lr>Lr)LOLDLnLni-n OOOOOOOOOOOOO xxxxxxxxxxxxx \£> K£> k£) K£) \£) \£> K£> \£) KO <*0 ^G \0 KO LnLnuOLnLnLnunLOLnLnunLnLn CJ^CTi(7iCTiCTiC^C^C^CTi(TiCnCT>W ^•cocococororocococororoco l i i i i i i i i i i i i OOOOOOOOOOOOO xxxxxxxxxxxxx Lni*ocr\r^*3-t£>i — c\j fO O «* N O^ r- CO C\J U"> CO CT> O O . — csjc\jcsjc\jroro^ ~ 1.22 —i ~ 1.21 _i ID 1.20 1. 19 1. 18 1. 17 o PNL CALCULATION CLAIBORNE^ CALCULATION I 12 3 4 5 6 7 RECYCLE STEP 9 10 FI GURE 9 . C . 5 . Comparison of Neutron Multiplication Values in every rod. The calculated isotopics for Np, Am and Cm for this strategy are presented in Table 9.C.7 along with the values calculated for the other strategy. Comparison of the data in Parts A and B of this table shows that the actinide inventory does not get reduced as much by the recycle of actinides concentrated in a few rods as in the recycle in every U0 ? rod. After the ninth recycle, the Np- 237 content is about 28% greater for the concentrated case. The Am-243 and the Cm-244 contents are likewise higher by ^13 and ^27%, respec- tively. Thus, concentration of actinides in 10% of the rods leads to a larger actinide inventory and reduces the transmutation rate. Nevertheless, the tech- nical feasibility criteria are met by this strategy. The increase in uranium enrichment required to achieve the same energy out- put as for 3.3 wt% U0 ? fuel without actinides is shown in Figure 9.C.6 for 9.C.14 BNWL-1900 TABLE 9 . C . 7 . Comparison of Actinide Inventories for Two Recycle Strategies Part A - Actinides Recycled in All Rods Actinide Inventory (Gms/MT of Heavy Metal) Recycle No. NP237 AM241 AM242 AM 243 CM242 CM243 " CM244 CM245 1 521.28 783.92 921. 49~ 993.84 1031.99 " 10 5 2.15 1062.81 1068.45 1 07 1.44 ~~ 1073.02 64,61 76.41 .58 .77 78.6 3 100.41 " 104.67" 105.12 " 104.94"" 104.72 104.58 104.49 7.89 9.95 .12 .28 22.90 76.66 ~116.20"~ 139.81 "152.96"" 160.08 163.88 165.90 166.97 167.53 1.07 5.75 2 3 4 5 6 7 77.51 77.81 77.95"" 78.03 78.06 78.08 ~ 78.10'""' 78.10 .79 .80 ".80" .80 .80 .80 .80 .80 10.15 10.15 — 10.15 " 10.15 10.14 " 10.14 .33 .34 " .34 " .34 .34 " .34 .34~ .34 " 9.63 12.02 ""13.37 14.10 " 14.49 14.70 8 9 104.44 104.42 " 10.14 10.14 ~ 14.81 14.87 Part B - Actinides Recycled in One Rod in Ten Actinide Inventory (Gms/MT of Heavy Metal) Recycle No. NP237 AM241 521.28 64.61 1 811.78 73.37 2 994.46 74.98 3 1118.27 75.85 4 1192.93 ~ 76.31 5 1253.06 76.68 6 1299.96 76.96 7 1334.71 77.15 8 1360.02 ' ~~77.27 9 1379.70 77.35 AM242 AM243 CM242 CM243 " CM244 CM245 .58 78.63 7.89 " .12 22.90 1.07 .71 103.43 ' 111.98" 9.68 .27 76.55 120.63 5.96 .75 9.90 .32 10.44 .78 115.43 9.98 .34 152.73 13.75 .70 116.81 10.01 .35 175.55 ' 16.05 .80 117.66 10.04 .35 193.78 17.88 .81 118.01 10.06 .36 203.55 18.95 .81 118.16 10.07 .36 208.71 19.57 .02" "118.22" 10.08 ~"~".36~ 211. 28~ ~~ 19.91 .82 118.26 10.09 .36 212.44 20.10 both strategies. Present PWR fuel management practices are to discharge one- third of the core each year. Thus, at the end of a cycle each 1/3 fraction of fuel has exposures of 33,000 MWD/MT, 22,000 MWD/MT, and 11,000 MWD/MT. The core reactivity roughly equivalent to the value when the core has an average exposure of about 22,000 MWD/MT. Thus, the neutron multiplication value ob- tained for U0 2 fuel (no recycle) at 22,000 MWD/MT ( k^ = 1.026) was used as a target for setting the enrichment of the recycle fuel. To achieve the same cycle length, the enrichment for recycle in every rod must be increased from 3.3 to about 3.43 wt% U-235 for equilibrium. (The rods containing the actin- ides were enriched increasingly in each recycle to achieve the desired cycle length.) This incremental enrichment was used for calculating the fuel cycle cost penalty which is around $14 million dollars per year (see Section 9.5.2). Recycling actinides in ten percent of the rods also requires an increase in 9.C.15 BNWL-1 900 3.45 — I— 3 o o LD CO » r^ CO co CTl r^ r— CM <* «3- *r *3" ^3- in CM 1 — +-> s- a> Gi- ro S- r_3 CO «=T co CO CO cn CM CM co , — l^~ r-. in r-^ r~~ 1^ r-^ r— r^ CM 10 CT> «* CO CO r— CM co r~. co CO cn cn cn CM in in LD r-~ i~^ o o Ol 00 "3" CM r^ CM cn in CO t— r— CO in CO r^ O 1 — r— r— ^1- «3- 00 r~- lO ^J- cn CM O co CM 3 1 — co in cm in 00 10 co «d- cm o cn co r^- cn co r— o cn cn cn CO CO CO CO CM CM CM ^ in lO CM 00 cr> vo in «d- r*^ ■* * cn cn cr> CO CM CM CM CM >3- ^r -o c ro cm 1 — in in r--- co cn 1— «a- «a- r-. vo l»- CO CM cn r— **■ 00 r^ r-» 10 r^ t-. in 1— 1 — 1^ cm cm cm cn vo t-^ co 00 r-. r>. r». 1 — W N Ol VO VO VO in r-» o co en o cn o cn co cu 00 co in 00 1 — in in in o 1 — cm co ^r , — VO •a- «=r CO «d- r> 00 00 cn cn 1— cn cn CM E 3 s- > c cu u C XJ •1- O CU 1— CM O O in VD r> 00 cn >■> cu -0 0£ (J Ol CU o; x: 3 f- CMC CT +-> O 4-> M- UJ cn rD cn 9.C.20 BNWL-1 900 TABLE 9. CIO . Incremental Fuel Cycle Costs For Increased Enrichment Due to Recycle of Actinides When Also Recycling Plutonium Recycle No. Incremental cost mllls/kwhe Volume Fraction of Actinide Region Incremental Enrichment in U Region Core Averaged Incremental cost mills/kwhe 1 .16 .169 .027 2 .71 .080 .057 3 .70 .104 .073 4 .68 .106 .034 .081 5 .64 .108 .075 .089 6 .61 .112 .101 .095 7 .59 .115 .118 .100 8 .57 .118 .135 .103 9 .55 .121 .148 .107 Equili brium .42 .144 .271 .132 TABLE 9.C.11. Economics Parameters Used Reactor Life Reactor Cost Interest on Reactor Cost Fabrication Loss (Recycled) Reprocessing Loss UF fi Conversion Cost Separative Duty Tails Composition Plutonium Value Fuel Cycle Working Capital Interest Fuel Inventory Interest U0p Fabrication Cost Mixed Oxide Fabrication Cost Reprocessing Cost U 3 8 Cost 30 years $400.00/KW 10% 1% 0.5% $1 .50/kgU $32.00/kg 0.3% U-235 $10.00/gm fissile 10% 8% $80.00/kgU $85.00/kg Heavy Metal $40.00/Kg $17.64/kg 9.C.21 BNWL-1 900 TABLE 9.C.12 . Composition Of Plutonium Recovered From Actinide Region Plutonium Composition, Wt.% Recycle Mo. Pu-238 Pu-239 Pu-240 Pu-241 •Pu-242 1 4.95 39.57 26.37 18.88 10.22 2 12.19 24.10 24.92 22.04 16.75 3 15.39 15.25 23.10 21.75 24.51 4 17.50 11.72 20.53 20.53 29.72 5 18.59 10.72 18.52 19.19 32.97 6 19.18 10.27 17.17 18.10 35.27 7 19.48 9.97 16.25 17.28 37.01 8 19.62 9.73 15.61 16.65 38.40 9 19.64 9.52 15.14 16.17 39.53 Equilibrium 16.54 7.38 13.28 13.46 49.34 9.C.22 BNWL-1 900 REFERENCES 1. M. Steinberg, G. Wotzak, and B. Manowitz, Neutron Burning of Long Lived Fission Products for Waste Disposal , BNL-8558, 1964. 2. H. C. Claiborne, Neutron Induced Transmutation of High Level Radioactive Waste , ORNL-TM-3964, December 1972. 3. A. S. Kubo, Technology Assessment of High-Level Nuclear Waste Management, ScD. Thesis, Department of Nuclear Engineering, Massachusetts Institute of Technology, Cambridge, Mass, April 1973. 4. A. S. Kubo and D. J. Rose, "On Disposal of Nuclear Waste," Science , Vol. 182, No. 4118, 1205-1211, December 21, 1973. 5. E. T. Merrill, ALTHAEA-A One-Dimensional Two-Group Diffusion Code with an Effective Four-Group Burnup , BNWL-462, May 1971. 6. E. A. Eschbach, et al., QUICK - A Simplified Fuel Cost Code , HW-71812, January 2, 1962. 7. M. J. Bell, "ORIGEN, The ORNL Isotope Generation and Depletion Code," ORNL-4628, Oak Ridge National Laboratory, May 1973. 8. Code of Federal Regulations , Title 10, Part 50, Appendix F. APPENDIX 9.D TRANSMUTATION BY FUSION (CTR) REACTORS SUMMARY 9.D.1 DETAILED RESULTS 9.D.4 A. The Transmutation Chain 9.D.5 B. Transmutation Calculations 9.D.8 90 1 37 - The Fast Neutron Flux Transmutation of Sr and Cs 9.D.8 - Thermal Neutron Transmutation of Dilute Target Samples 9.D.11 1 07 - Transmutation of Quantitative Amounts of Cs in a 9.D.14 Moderating Blanket REFERENCES 9.D.19 9. D.l APPENDIX 9.D TRANSMUTATION BY FUSION (CTR) REACTORS SUMMARY The unique features of a fusion reactor (otherwise called a Controlled Thermonuclear Reactor CTR) as a waste transmutation device are the high energy of the neutrons available, and the high flux anticipated. Due to the high energy, (n,2n) reactions can contribute significantly to the process, while the high flux and high source strength makes waste trans- mutation at reasonable rates appear possible a priori. Analyses indicate: (1) For the specific isotopes analyzed, calculations indicate that transmutation decreases the radiological toxicity. That is, the ultimate daughter products created are non-radioactive stable isotopes. This is necessary for the feasibility of any trans- mutation scheme. Waste transmutation is a function of the neutron flux spectrum and this conclusion does not necessarily apply to all neutron transmutation schemes. A particularly attractive transmutation rate for these fission products (FP's) has not been demonstrated by the analysis to this point. However, is appears that detailed parametric studies performed on specific target loading techniques may indicate ways to achieve transmuta- tion rates several times larger than the natural decay rates of these isotopes. (2) It appears that there will be adequate physical space in a CTR blanket for fission product loadings, and that the thermal burden of these fission products on the blanket will not be too great. 9. D.2 (3) It has been shown that a few CTR power plants can, in principle, transmute all of the Sr-90 and Cs-1 37 created by an essentially all fission-reactor electrical power economy. (4) The daughter products produced appear to be approaching well defined equilibria after a year or so of operation. They should, therefore, cause few perturbations on the characteristics of operating CTR power plants. (5) The question of the economics of CTR power plants as fission product burners was found to be most interesting. First of all, it was shown that a transmutation scheme which is a net absorber of neutrons (n,y) would be relatively expensive. Secondly a CTR transmutation scheme based on the n,2n reaction could be a net producer of neutrons for long irradiation periods. This observation has a number of implications. First of all, it indicates that the process could be a net producer of income. Secondly, it shows that the process may not require early separa- tion of the daughter products. Because certain fission products may prove to be of value to CTR power plants, this finding has some implications as to what strategy should be used in applying retriev- able storage concepts. The real economic factors for the trans- mutation of radioactive waste in a CTR would appear to be not in the neutronics but in the manufacture of FP irradiation targets. Among other considerations are possible increased capital costs associated with using the CTR for this purpose and increased operational costs for handling and safety. 9.D.3 (6) The transmutation of heavy elements in a CTR is a separate con- sideration. As a rule the longer the natural half-life of a particular isotope and the larger the cross section of that isotope, the more attractive that isotope is for transmutation. Consequently, while Sr-90 and Cs-137 have been found to be possible transmutation targets, heavy element wastes are more attractive for transmutation in a CTR. This is because their neutron cross sec- tions for transmutation are significantly higher. The CTR will be better for neutron-induced transmutation than other known approaches if it attains the higher neutron flux and excess neutron capacity currently predicted. An additional reason why transmutation of heavy isotopes appears particularly attractive in a CTR has to do with the specific neutronic characteristics of a CTR. Analysis has demonstrated the importance of high-energy neutron threshold reactions in CTR transmutation. It is concluded that the (n,2n), (n,3n), and fission reactions could be of major importance in the transmutation of heavy isotopes in a CTR. These reactions would result in either the transmutation of isotopes back up the chain (i.e., Pu-240->-Pu-239) or fission. In either case, the net effect could be a rapid decrease of the potential radiological burden on the environment. In addition, the average thermal energy deposited in the CTR blanket per fusion neutron will be enhanced through fission. Thus, the confinement conditions on the fusion plasma would be several times lower than the conditions required for energy breakeven without the presence of fission. 9. D.4 DETAILED RESULTS The CTR is of interest for the possible transmutation of radioactive waste because of both the large neutron source and high neutron fluxes which have been projected for fusion reactors. The most troublesome fission product nuclei are those which have relatively small neutron reaction cross sections. Consequently, high values of neutron flux are required for reasonable transmutation rates. The requirements on neutron flux-level are obtained by comparing neutron-induced transmutation rates with radioactive decay rates. If the assumption is made that, as a minimum, the transmutation rate needs to be about equal to the natural radioactive decay rate then X - a (Dl) where A is the radioactive decay constant and o and

ZD Q O Q£ Q_ on o < O o I 1- oo o Q_ =3 O CO 10 -4 ._' 10 10 -6 H- 10 - ~~™lx " — / 91 Zr — - / Sr 89 Y — 90 Y — s 91 Y jf y^x — 91 Sr 1 1 1 - TIME (YEARS! FIGURE 9. D.I . Buildup of Sr-90 Daughter Products with Time 9.D.10 FIGURE 9.D.2, 1 2 3 TIME (YEARS) Buildup of Cs-137 Daughter Products with Time 9. D.ll Zr-89 which decays shortly (78.5 hour T, , ? ) to Y-89 the net result is only a stable nucleus and the Zr-90 fraction shown on the figure contains some Y-89. Similarly, the transmutation chains for the Cs-137 target were terminated at Ba-135 and La-139 so that the fractions shown on the figure for these isotopes also contain product nuclei. The neutron production rates per neutron absorbed in the mixed nuclide samples of the initial Sr-90 and Cs-137 targets were also calculated as a function of irradiation time. The values of neutron production per neutron absorbed were about 1.75 for the Sr target and about 1.50 for the Cs target. Neither value changed more than a few percent over the three year irradiation which was calculated. Thus it was concluded that in dilute targets the FP nuclei contributed neutrons to the economy of the CTR blanket. Thermal Neutron Transmutation of Dilute Target Samples - Preliminary studies sponsored by the USAEC Division of Controlled Thermonuclear Research have been made at PNL, in which the intent was to establish an approximate upper bound to the highest transmutation rates which could be achieved utilizing the neutrons produced by a controlled thermonuclear fusion reaction. The results of these studies help to establish the types of radioactive waste targets which are worthy of more serious and practical investigation. The detailed results of the calculations have been published in a BNWL document^ 6 ' and summarized in an American Nuclear Society presentation. ' 9.D.12 Consistent with the intent of the study as given above, calcula- tions have been made using a moderating blanket, the initial portion of which is beryllium, to obtain multiplication of the fusion plasma source neutrons through the beryllium n,2n reaction. The geometry chosen for the calculations was essentially that of the CTR calculational (4) standard benchmark, ' a TOKOMAK with 200-centimeter inner vacuum wall radius and a blanket 100 centimeters thick. Calculations have been made for both D-D and D-T fusion plasmas, each with a fusion reaction rate 2 producing a neutron power loading at the inner wall of 10 MW/m . The neutron distribution in the blanket was calculated using the ANISN code. ' A maximum thermal flux value of almost 3x10 neutrons/(cm ) (sec) was found to occur at the inner surface of the beryllium moderator in the absence of any absorbing vacuum wall material. The transmutation calculations which were made utilized the neutron energy spectra and flux values representative of that of the first centimeter thickness of the Be blanket. Radioactive waste inventories representative of an expected LWR economy were established using results obtained from the ORIGEN code^ ' for chemical isotopic mixtures of fission products Sr and Cs and of actinide isotopes. As before, the transmutation code ALCHEMY was used to calculate the isotopic concentrations for dilute targets as a function of irradiation time. A com- puter code was written to transform the calculated isotopic activities to dispersal toxicity values. The results of these calculations are shown on Table 9.D.1 for both D-D 2 and D-T neutron sources and for neutron wall loadings of 1 and 10 MW/m . The 9.D.13 TABLE 9. D.I. Toxicity Half-Life^ as a Function of Neutron Wall Loading (P ) Toxicity Half-Life, Years Target Plasma p w = o.o P =1.0 MW/m 2 w P =10.0 MW/m 2 Sr-90 D-T 32.0 7.6 1.3 Sr< b > D-T 32.0 8.0 7.4 Sr-90 D-D 32.0 4.7 0.48 Cs-137 D-T 30.0 19.0 5.6 Cs< b > D-T 14.0 12.0 4.0 Cs-137 D-D 30.0 16.0 4.4 Kr-85 D-T 11.0 8.6 6.1 1-129 D-T 1.6xl0 7 0.35 0.27 Actinides D-T 460 (c) 0.60 0.056 Actinides D-D 460 (c) 0.23 0.037 a. Defined as the time required for the toxicity index from a particular isotope and its daughters to decay to one-half of its original value. The toxicity index itself is the volume of air or water required to dilute the nuclide to its Radiation Concentration Guide value. b. This is isotopic mixture of Sr or Cs from LWR fuel at 33,000 MWD/MT and 10 years cooling. c. Am-241 only. 9.D.14 expected large enhancement in reduction of actinide toxicity due to the neutron irradiation is apparent. The actinide toxicity values do not include the effects of fission product created in the fission process. A similar drastic reduction in toxicity values is obtained for a target of 1-129. The reduction in toxicity values for D-D source neutrons relative to D-T source neutrons for the same neutron wall loading arises primarily from an increase of about a factor of two in thermal neutron flux. The relative merits of isotopic separation of Sr and Cs cannot be clearly deduced from just the hazard values listed on Table 9.D.I. More detailed study of economic and other factors is required to produce any value judgments in this area. Transmutation of Quantitative Amounts of Cs-137 in a Moderating Blanket The results of the preceeding studies of irradiations of Cs-137 have shown that there are some potential hazard reductions when dilute samples are irradiated in either fast or thermal neutron flux spectra. The in- dicated reductions were marginal in benefit, and further studies were considered to be desirable. The studies of dilute targets had not quantitatively assessed the quantity of Cs which could be subjected to transmutation. As a result, it was decided to study the transmutation rates as the loading of Cs in a Be moderating blanket was progressively increased. The results of this study have been summarized in an American Nuclear Society presentation^ ' and are published in more detail elsewhere. ' The calculational model of this study was intended to represent a more realistic blanket configuration which was, as before, built upon the standard calculational benchmark model with a D-T fusion plasma with 9. D.15 2 a neutron wall loading of 10 MW/m . In these calculations an inner vacuum wall of 0.5-centimeter thickness of niobium was placed at the 200-centimeter radius as shown in Figure 9.D.3. This was followed by a 3-centimeter coolant zone (void in the calculations) and a second 0.5-centimeter Nb wall. The target zone was the next 20-centimeters annular thickness of the blanket. The target zone was followed by an annular zone of beryllium of 70-centimeter thickness followed by a 6-centimeter thickness of lithium. The calculated transmutation rates in the target zone were found to not be \/ery dependent on whether the 70-centimeter zone of beryllium was replaced by graphite or whether its thickness was significantly reduced. In the initial calculation the 20-centimeter thick target zone was filled to 80% volume with beryllium and a 20% volume void space to allow for gas cooling channels. The infinite dilution reaction rates of Cs-137 were then calculated using the ANISN code. The calculated maximum and volume averaged reaction rates for the neutron capture and n,2n processes which were obtained are shown as the first entry of Table 9.D.2. The Cs-137 concentration was then progressively increased, a given volume fraction of Be being replaced by Cs-137 at the atom density of elemental Cs as indicated on the following entries of the table. The total specific transmutation rates for Cs-137 are seen to be not yery dependent on Cs concentration, the main effect being an increase of the (n,2n) reaction rate to almost compensate for the decrease in the neutron capture rate as the Cs concen- tration is increased. In view of the above results, the beryllium moderator was removed from the target zone and the effect of filling this void with different volume 9.D.16 O O CO - o +-> ^r ^ CNJ o C (XI re _ i — LU CD CD o o T3 C\J ^ i. < -a h— c C/> -t-> ~~~ oo CD Q LPv a: i— | 1 i— < O — ■ -o Q CD < ■i— 4- a: •i— ■o O • ro Q CT> LU DC U3 QJ co to <_> s- o 4- +-> o o 03 to QJ QJ Cd 4-> 03 c q: o c to o =3 +-> fO ^ E E O to -^ E O 03 I— S- 1— o O)--^ > 03 * N ■!■" s »— -^ to -M CM S- o \ rO QJ i— QJ 4- l >> 4- h- *■ *• LU 9.D.17 CO ^i- r-*. >— to en o r*» r^ i--» CO CO CO cr> c C o CM •1 — n 4-> QJ c <_) 4-> A rt3 03 -e- - E o r - •^ n X +-> CD c ra o +-> A ?■ 03 03 -e- cu OH O o; V c o • 1 — i- +-> 03 u QJ 03 >1 s_ *■»— • 4- c o ^ — ■•» • f— S- +-> 03 u a> o3 >> S- *— -" 4- to r~- co i — c\j o cr> to to r-- co en en co CM CM CM C\J CM cm cm o o o o o o o O O CM CTi CM CM CM lil «t O * 00 ID ifi "vf ^" *tf" CO CM CM CM O O O O O O O u to QJ 1 to o f— CM VO CO X E CM CM o CO CO CM 00 CO LO to O to i — CTt i — r»« to r— i— i— O o o 03 +-> O c o • r— ** — * +-> -Q 03 *-— ^. +-> CU s_ 13 +-> >! E 03 tO CtL c E 03 •■^ S- CJl 1- -X-. O LO LO ■— .— CM CM CTl 00 CTi CTl LO i— i— CM CO 03 CD-— - > 03 ■" — . •i— ** — ** to +J CM s- O \ 03 CLI r— QJ 4- 1 >i 4- I— UJ o o CTt CTl CO 1^ CO CTl CTl CTl CTi CTl CO O c c 0J o CM CD •1— CU #* 03 -t-J +J c S- o 03 A cu ro or -e- > QJ D <=c o V >- •1 — QJ «% +-> +-> c o 03 A 03 CU -e- QJ t> CC V to i o r— X c •i- S- +-> 03 O QJ 03 >, LT) CTl r-% oo 1 — , — CTi i — i — CO to >vf CM O i — r— ■ — ^~ CM CM CM O o o o o O O o o o o o O O c o * — s •1 — S- +-> 03 (J QJ 03 >i 4- 00 CTl CTi , — LO O 00 CM r— r^ CM r-«» LO CO «* ^d- CO CO CM CM CM O O o O O O O o o CTi tO LO i— O «d" CO r— CO r^ *d- r^ co oo r->. to to .— .— .— o c O QJ QJ •i- CO E 4-> &S 3 U ^~. ^~ 03 tO o i- O > LL. &s LO o o o LO CM CO o o •^Hs. oo r^ ^*^, \ — ^ •^ o ^•N^ ** — , LO o o o o o LO CM LO LO 00 1 — S- fO • OJ -C >> +-> CD ^ c QJ QJ a. ' — F QJ o O +J • r- 03 > QJ S- T QJ q-q: 1— CTl (_) CM CM 4- o O o S- QJ 4- 4-> O QJ E >l +-> s- •i — QJ > Q. •1 +J ^-^ u 4-> 03 " — * 3 >>s: 03 O O QJ O -a CM OJ to > QJ •1— O +-> Z3 u -a 03 o o s- •1— Q. ■a 03 CU i. (J •r- LTl > QJ QJ -o T3 3 i — to O • i— c -C 03 -Q 9 e D.18 fractions of only Cs-137. From the calculated results of this study as shown on Table 9.D.2 it is again apparent that the total specific activities are not yery sensitive to the amount of Cs loaded or to the presence of beryllium. The maximum practical amount of Cs which can be transmuted is the 291 kg/yr per meter length of blanket as obtained for an 80% volume of Cs-137 in the target zone. This quantity of transmuted Cs-137 has been compared with the amount of inventory Cs-137 by considering the Cs-137 produced in a LWR. For this particular Cs-137 loading, the Cs-137 production rate in a LWR would equal the transmutation rate in a CTR blanket of the same thermal power if only one percent of the CTR blanket were loaded for the Cs-137 transmutation. The conclusions of this study are that interesting quantities of Cs-137 could be transmuted under the projected CTR blanket loading conditions. The reductions in Cs-137 toxicity are, however, still projected to be at most a factor of about three. In addition, the buildup of product nuclei has not yet been studied in order to establish the requirements of periodic chemical processing and associated costs. 9.D.19 REFERENCES 1. W. C. Wolkenhauer, "The Controlled Thermonuclear Reactor as a Fission Product Burner," Trans. Am. Nucl. Soc, v ol. 15, p. 92, 1972. 2. W. C. Wolkenhauer, "The Controlled Thermonuclear Reactor as a Fission Product Burner," BNWL-SA-4232, June 1972. 3. W. W. Engle, Jr., "A User's Manual for ANISN," K-1693, Union Carbide Corporation, Nuclear Division, Oak Ridge Gaseous Diffusion Plant, Oak Ridge, TN. 4. International Working Sessions on Fusion Reactor Technology, ORNL, CONF-710624, June 28-July 3, 1971. 5. B. H. Duane, "Time-Variant Isotopic Transmutation GE-HL Program ALCHEMY," Physics Research Quarterly Report, October, November, December 1963 , USAEC Report HW-80020, pp. 4-7, 1964. 6. W. C. Wolkenhauer, B. R. Leonard, Jr., and B. F. Gore, "Transmutation of High-Level Radioactive Waste with a Controlled Thermonuclear Reactor," BNWL-1772, September 1973. 7. W. C. Wolkenhauer, B. R. Leonard, Jr. and B. F. Gore, "Transmutation of High-Level Radioactive Wastes with a Controlled Thermonuclear Reactor," Trans. Am. Nucl . Soc. , vol. 17, p. 52, November 1973. 8. M. J. Bell, "ORIGEN - The ORNL Isotope Generation and Decay Code," ORNL-4628, May 1973. 9. B. F. Gore and B. R. Leonard, Jr., "Transmutation in Quantity of 137 Cs in a Controlled Thermonuclear Reactor," Trans. Am. Nucl . Soc. , vol. 17, pp. 52-53, November 1973. 10. B. F. Gore and B. R. Leonard, Jr., "Transmutation of Massive Loadings of 137 Cs in the Blanket of a Controlled Thermonuclear Reactor," Nucl. Sci. Eng. , vol. 53, p. 319, 1974. APPENDIX 9.E COMMENTS OF PEER REVIEWERS SUMMARY COMMENTS Dr. J. L. Crandall Dr. D. G. Foster, Jr. Dr. R. L. Hellens Prof. A. S. Kubo Prof. H. W. Lefevre Prof. C. J. Poncelet Prof. B. H. Spinrad Page 9, ,E, .1 9, ,E. ,3 9. ,E, .6 9, ,E. ,16 9, ,E, .27 9. ,E, .33 9, ,E, ,35 9. ,E. ,41 9.E.1 APPENDIX 9.E COMMENTS OF PEER REVIEWERS SUMMARY Seven people technically expert in the various areas reviewed the draft copy of the study, High-Level Waste Disposal Alternatives: TRANSMUTATION PROCESSING BNWL-B-301 Section 9, and provided their comments. The peer group met with PNL and AEC personnel in Seattle on January 14th to discuss the report, give their views, and resolve differences of opinion. Those in attendance were: Peer Group Dr. J. L. Crandall Dr. R. E. Hellens Prof. A. S. Kubo Prof. H. N. Lefevre Prof. C. J. Poncelet Prof. B. H. Spinrad Dr. D. G. Foster, Jr. Savannah River Laboratory Combustion Engineering, Inc. U.S. Military Academy University of Oregon Carnegie-Mellon Institute Oregon State University Los Alamos Scientific Laboratory AEC R. W. Ramsey PNL C. M. Heeb B. R. Leonard, Jr. R. C. Liikala T. I. McSweeney K. J. Schneider In summary, there was general agreement with respect to conclusions on technical feasibility. Those areas where concensus or near-concensus opinion exists and where the report was modified to reflect these opinions are as follows: 1. Actinide recycle in LWRs seems technically feasible but not necessarily attractive. 2. The LMFBR is probably more attractive for actinide recycle. An immediate Research and Development goal should be an evaluation of the concept. 3. If actinide recycle in LWRs is to be done, the need for immediate irradia- tion experiments is urgent. 9.E.2 4. The actinide recycle concepts in LWRs calculated by Claiborne and by PNL almost certainly represent a worst case in terms of reactor operation and economic penalty. 5. The concept of actinide recycle in LWRs as a burnable poison needs to be evaluated. 6. The use of a captive fission reactor, perhaps special-purpose, perhaps part of the National Repository, deserves serious consideration as an alternative to actinide recycle throughout the fission power reactor industry, LWR or LMFBR. 7. Actinide recycle in LWRs needs to be evaluated as a hazard relative to plutonium recycle as a base case. 8. The concept of the transmutation of actinides in a near-critical fission assembly driven by an accelerator neutron source is not attractive since the same job can be accomplished in fission reactors at much less cost. 9. The utilization of thermonuclear explosives to transmute radioactive waste is probably not publicly acceptable, even if feasible. 10. Considering transmutation of waste in fusion reactors, it should be emphasized that CTRs do not presently exist. 11. Transmutation of waste in fusion reactors should be based on a captive, isolated CTR--not a CTR power economy. Additional analysis and discussion has resolved the main difference of opinion regarding the feasibility of using accelerators and nuclear explosives as transmutation devices. This report reflects the comments provided by the peers. OSR 24- A I I ( (REV 1-72) SLLPOK 9.E.3 E. I. du Pont de Nemours & Company INCORPORATED ATOMIC ENERGY DIVISION Savannah River Laboratory Aiken, South Carolina 29801 810-771. 2670 TEL 803-824 6331 VVU AUGUSTA. Dr. B. R. Leonard, Jr. Staff Scientist Battel! e Northwest Laboratories Battel le Boulevard Richland, Washington 99352 January 10, 1974 Dear Dr. Leonard: At your request I have reviewed BNWL-B-301 , Section 9, the Transmutation Processing section of the Battel le investigations of high level waste disposal alternatives for their Advanced Waste Management Studies. The report accomplishes its main purposes in admirable fashion, identifying those transmutation processes worth a detailed study and providing a comprehensive and systematic review of essentially all the possibilities. Actually I must admit it was not until I had worked through the report references that I realized just how necessary such a unified study was to correct the many misconceptions in the literature. I am in general agreement with the report's conclusions that few processes except transmutation in inexpensive neutrons from the still-to-be-developed fusion reactors or actinide transmutation in fission reactors hold current promise. However, as a small quibble, there might also be limited promise to byproduct transmutation in nuclear explosions performed primarily for gas stimulation or similar Plowshare purposes, even though the report quite properly demonstrates that the nuclear explosions would not be attractive for waste transmutation alone. As usual, in a basic study of this sort the key questions are more philosophical than technical, and I would like to raise two of these questions. The first concerns the urgency for the actinide recycle. As clearly developed in the report and in most other waste management studies, the hazards from the byproduct actinides (excluding plutonium) are several orders below the fission product hazards for the first few hundred years of waste storage, but do become limiting (along with some of the long-lived fission products, such as 12 ^I) at very long times. The current consensus seems to be tr.at the fission products can be safely put into retrievable engineered storage for 50-100 years pending development of ultimate disposal methods, and the question becomes why should this not also be done with the byproduct actinides, which, due to their lower heat generation, are probably much easier to store in the interim period. There seem to be three possible reasons why the actinides might merit special treatment. All are connected with the fact that the actinides need to be safeguarded for more than a million years whereas most of the fission products will B. R. Leonard 9.E.4 have decayed to safe levels in less than a thousand years. First, the ultimate disposal methods chosen for the fission productsmay not be suitable for the longer time demands of the actinides. Second, the combined costs of interim plus ultimate disposal of the actinides may exceed that of immediate transmutation. Third, the actinide hazards may be greater than should be accepted for interim storage. I believe that the first two possibilities can be eliminated. In the first regard, many of the disposal methods being proposed for the fission products, such as disposal in space, do apply to the actinides and, even if special methods are required, it is almost certain that in 50-100 years less expensive alternatives can be developed than transmutation in light water reactors. Similarly, in the second regard, both the possibility of relatively inexpensive future disposal methods and time-value-of-money considerations say that immediate actinide transmutation will be less expensive than interim storage and later disposal only if it is in fact less expensive than the interim storage alone, which is demonstrably untrue. Thus the real question comes down to whether interim storage is safe for the byproduct actinides, which might contaminate the earth for a million years if they were released. A direct answer can be given only by detailed biological and transport studies. However, the problem can also be put into perspective by recognizing that, on a gram-for-gram basis, the plutonium produced in the reactors offers approximately equal hazards to the byproduct actinides and that any extraordinary circumstances which could lead to actinide release from engineered storage - war, technological breakdown, etc., - are even more likely to lead to plutonium releases. Thus in viewing the hazards reduction from byproduct actinide burnup, the percentage base should not be from the byproduct actinides alone but from all actinides including plutonium. Since the byproduct actinides are only produced at about a tenth the rate of the plutonium, this change in bases greatly reduces the importance unless a fuels recycle program is proceeding to burn up the plutonium without touching the byproduct actinides. My own prejudices on actinide treatment in view of this analysis are that partitioning of the actinides may be justified at the present time since it greatly broadens the fission product storage options and might be much more difficult to perform at a later date. Also, the separated actinides may eventually be valuable products in their own right as radiation, heat, or neutron sources. Likewise, it would be desirable to have more data on fast reactor transmutation of the actinides. However, little additional work seems justified on light water reactor transmutation at this time, particularly since some pertinent data are still coming in from the various transplutonium programs (Savannah River, for example, expects to publish an evaluated multigroup cross section set by June). The other philosophical question I would like to raise is on the use of dilution to RCG as a hazard index. Certainly it makes a simple and definite index, and it is used sensibly and effectively throughout the report. However, as a general purpose index it has obvious disadvantages since it takes no account of half-life or dis- persibility. Worse, if the ultimate disposal method effectively eliminates the waste, say by shooting it into the sun or even by guaranteed geologic dispersal, such a hazard index has no meaning at all, the required index being merely a yes-no index of those nuclides that have to be removed and those that don't. B. R. Leonard The following are a few pertinent notes I made while reading the report. Page 9.3 As noted in text, hazards index considered is probably too simple for general use. 9.5 Need economic criteria as well as physical ones. 9.20 Seems to consider that undesirable products are formed on a ratio of 1 per fission. If plutonium is not considered, more like 0.1-0.2 per fission. 9.23 Repeated in my copy. 1 29 9.25 I could probably be transmuted quite effectively in a spallation reactor. 9.47 Some of the required cross sections are available in new SRL studies (in press) for 252 Cf production. 9.55 I personally put the partitioning studies at the highest priority in the research programs, since the text shows clearly that good partitioning is basic to most other handling methods. 9.60 Potential reduction in hazards by two orders of magnitude is true only if plutonium produced in reactor is excluded from consideration. 9.61 Since these programs always seem to cost more than anticipated, it is doubtful if economics will be attractive under even the "better" methods. 9.66 I believe costs increase by M'Vyear, not 0.1^/year. 9.76 Two rather than to. 9. A. 4 Apparently the energy calculation is performed on the oasis of one undesirable product per fission but since accelerator electrical efficiency is also neglected overall balance, it is still approximately correct. 9.A.28 U-235 for U-233? Sincerely yours, J. L. Crandall, Director Advanced Operational Planning JLC/fy LA-UR 74-74 TITLE: 9.E.6 REVIEW OF PNL STUDY ON TRANSMUTATION PROCESSING OF HIGH LEVEL WASTE AUTHOR(S): D . Graham Foster, Jr., Editor SUBMITTED TO: I 8 ©a Presented at meeting held at the Battelle Seattle Research Center for preliminary study of alternative methods of disposal of high level radioactive waste for the USAEC Division of Waste Management and Transportation, January 14, 1974. By acceptance of this article for publication, the publisher recognizes the Government's (license) rights in any copyright and the Government and its authorized, representatives have unrestricted right to reproduce in whole or in part said article under any copyright secured by the publisher. The Los Alamos Scientific Laboratory requests that the publisher identify this article as work performed under the auspices of the U. S. Atomic Energy Commission. of the University of California LOS ALAMOS, NEW MEXICO 87544 Form No. S3G Si No. 2629 1/73 UNITEO S7ATES ATOMIC CNESCY COMMISSION CONTRACT W-7JOS-CNG. 33 9.E.7 REVIEW OF PNL STUDY ON TRANSMUTATION PROCESSING OF HIGH LEVEL WASTE Editor: D. G. Foster, Jr. Contributors: T. R. England D. G. Foster, Jr. D. R. Harris J. W. Healy H. S. Jordan E. A. Knapp K. D. Lathrop R. E. MacFarlane D. W. Muir R. G. Shreffler P. D. Soran R. F. Taschek SUMMARY We have reviewed briefly a draft of BNWL-B-301, Section 9: "Trans- mutation Processing, Advanced Waste Management Studies, High Level Waste Disposal Alternatives," at. the request of Pacific Northwest Laboratories. This document describes a preliminary study of transmutation concepts (accelerators, thermonuclear explosives, fission reactors, fusion reac- tors) for the management of fission-product and actinide waste generated in the nuclear-energy economy. We strongly support the principal con- clusion of the study; i.e., that disposal of higher actinide nuclides by recycling in fission power reactors appears to be attractive with near- term, high-likelihood technology, and that major R&D emphasis is appro- priate in that area (page xi) . Effective, large-scale applications of this concept will depend on R & D which affects the economics of recycle utilization of the actinides including plutonium. With respect to dis- posal of fission-product nuclides, we feel that the work done to date may have discarded fission reactor transmutation of selected nuclei (af- ter chemical separation) prematurely, whereas a more detailed study may demonstrate feasibility for special cases. We support the PNL conclusion that transmutation both of actinides and of fission products in fusion reactors appears to be sufficiently at- tractive that a modest R&D effort is projected in this area. We feel, however, that the case for transmutation by particle accelerators is stronger than suggested and deserves an R & D effort to determine the utility of GeV proton bombardment with or without fission boosting. In this connection we describe a fast actinide burner, either stand-alone or driven by a proton accelerator, which appears to have economic potential. Contrary to the PNL conclusion, we suggest that transmutation of high-level waste by neutrons from an underground nuclear explosion may be very promising, rather than infeasible. 9.E.8 I. INTRODUCTION This report was prepared in response to a request from the Pacific North- west Laboratories of the Battelle Memorial Institute for a review of a draft of BNWL-B-301, Section 9: "Transmutatation Processing, Advanced Waste Manage- ment Studies, High Level Waste Disposal Alternatives." This section discusses transmutation processing for disposal or utilization of radioactive wastes from nuclear-energy sources. It was not possible in the time available to pre- pare a review which is appropriate to the importance of the problem. According- ly, what follows is a composite of views expressed by the contributors, grouped into sections devoted to related concepts. Comments expressing the views of one or more contributors are arranged as a lettered sub-section. Some of these views are somewhat disparate in detail, reflecting both the short time available for editing and technical differences in judgement. There is a general consen- sus, however, on the on the conclusions stated in the summary. II. CONCEPTS FOR EVALUATING TRANSMUTATION AS A METHOD OF WASTE MANAGEMENT A. Most of the contributors feel that the dilution "hazard index" is an appro- priate indicator for beginning exploration of disposal methods. However, it ignores some vital details and may be misleading in making a final de- cision. The effective volume of air or water which dilutes an accidentally released radionuclide is difficult to predict. In order to determine the hazard to the general population, the transport properties of the material released need to be considered in detail, and the pathways by which it can affect the general population explored individually to determine which one presents the greatest hazard. A simple change in chemical or physical form can drastically alter the effective hazard index. B. There is a semantic problem in the words "hazardous" and "waste." At present, certain hazardous fissile and fertile nuclides are considered useful, whereas other actinide nuclides are considered waste or a "nui- sance." The very long-lived products of reactors have such low specific activities that it may be misleading to lable them as problems. For ex- ample, 9^Tc -£ S treated by PNL as a waste to be disposed of, although yym Tc is routinely injected into humans for gamma radiography of the brain, and apparently causes no biological problems after it decays to Tc. In this respect it appears no more desirable to transmute it to some shorter-lived species than to transmute the 2 ^ 2 Th used in the mantles of gasoline lan- terns. C. Table 9.3 gives the hazard index for some important fission products and for tritium. We feel that the discussion of transmutation would benefit if the hazard index for each nuclide was included when it is discussed. It would place these in better perspective to tabulate also the hazard indices of 232 Th, natural and slightly enriched U, and fuel-grade ' U and Pu. The latter are hazards, but certainly not "waste." 3 9.E.9 D. In considering transmutation, it is instructive to factor the hazard index, h = I(l/x), where I contains the biological information and 1/t is the mean decay rate per nucleus. Then the hazard index integrated over all time is H = t(I/t) = I, which is independent of i. Thus, unless transmutation changes the biolog- ical considerations the cumulative hazard is not changed by shortening the half life. The primary advantage of such a transmutation is then in de- creasing the time span over which the waste must be protected. Note, how- ever, that in fissioning actinides the predominant radiation is changed from short-range a emission to more penetrating 3 and y» which complicates further handling. If a nuclide with mean life t is taken from storage where it has a mean probability per unit time of catastrophic release P Q and placed in a trans- mutation device which shortens its hazard mean life toT t with a catastropic re- lease probability P t , the time- integrated hazard to the population is changed from P T to Pt-Tf. Clearly we would like to require °° 1 P 1 PT^~ t t 00 t t P„ t LOO If we hypothesize that in a reactor, as opposed to a salt bed, P t /P Q = 100, then we must require that the mean life be shortened by at least a factor of 100 (rather then "several times") before transmutation can be justified. This makes the transmutation of such fission products as ^'Cs an ^ " u Sr, which have the highest hazard indices, even more unattractive than esti- mated by PNL, but makes transmutation of the very long-lived wastes more attractive even with low conversion rates. We must also require that the transmutation products be either stable (or so long-lived as not to be considered radiological hazards) or else have half lives so short that they decay essentially completely in acceptably few years or decades. E. These considerations underline the fundamental fact that any processing of wastes involves temporarily increasing both the hazard to workers in the operations of refueling and processing, and the probability of catastrophic release. In marginal cases, this would favor long-term storage over trans- mutation. F. Our contributors feel strongly that the feasibility of any transmutation cannot be proved or refuted without a detailed examination of each nuclide in question and of its possible pathways to the population and environment in the event of an accidental release from any point of the processing/ refueling operations. Since each nuclide may require its own strategy of management, we concur with the PNL emphasis on the importance and value of chemical, and perhaps even isotopic, partitioning. 9.E.10 G. It is too early to assessaccurately the prospects for laser-induced photo- chemical separation of elements and of isotopes. However, we feel that this possibility might become very important within, perhaps, a decade, and is certainly more promising than the gamma-ray laser. It might be expected to lower the cost of partitioning, improve the stripping of actinides from the fission-product stream, and perhaps profoundly alter the economical strate- gies of reactor design and fuel recycling. Even without such a breakthrough, Kubo and Rose have pointed out the substantial advantage in pushing the con- centration of Pu in the fission-product stream below 10 _l ^ a/o by convention- al means. The draft report does not explain why the separation of fission products from Pu is so much more complete than the separation of Pu from the fission products. III. ACTINIDE RECYCLE IN FISSION REACTORS A. Our contributors are virtually unanimous in endorsing the promise of acti- nide recycle in fission reactors. The PNL report emphasized actinide re- cycle in fission power reactors, primarily LWR's, but special-purpose re- actors also may have promise. Certain actinides produced during fission power reactor operation, e.g., plutonium isotopes, are already projected for recycle as power reactor fuels. The effective, large-scale recycle of higher actinides in fission reactors will depend on the successful resolution of a number of technical and eco- nomic factors. Many of these factors are common both to problems of plutoni- um recycle and to higher-actinide recycle and should be examined together. The design of actinide-recycle systems presents a large number of options, and it is not clear that the economic costs' and benefits can be reliably pro- jected without careful optimization. In general, it appears that among the penalties of actinide recycle are additional shipping hazards and costs, ad- ditional hazards in the event of a loss-of-coolant accident, and additional complexity and cost of the design process. Among the benefits of higher- actinide recycle are reduced costs for long-term actinide storage and re- duced fuel-separation costs. It is not completely clear that there is an economic penalty associated with inclusion of higher actinides in plutonium fuel. It is possible that, with optimized reactor design, increased uranium enrichment and increased control-rod swing can be avoided. The cost of al- ternative actinide disposal must be borne by the power producer if the bene- fits of actinide recycle are to be apparent. B. It is worth underscoring the PNL caution that any transmutation blanket or loading added to a reactor, whether fission or thermonuclear, acts as a parasite in competition with breeding. The loading of waste which would still permit breeding is not necessarily large enough for the reactor econ- omy to consume its own higher actinides at equilibrium. Clearly calcula- tions directed at this question should have a high priority in Phase 2 of the program. 9.E.11 C. Kubo and Rose have claimed that the "relative waste toxicity" of the fission products in the early decades of decay exceeds that of all actinides to- gether by a factor of more than 500. Thus, the immediate problem of high- level wastes is overwhelmingly that of the fission products. This fact will probably have to be borne in mind in establishing funding priorities for de- veloping all-actinide recycling technology. D. We emphasize that every industrial process generates waste. Processing re- cycled actinides currently generates wastes which are sometimes inordinately expensive to recover chemically, and they may be in forms more dangerous than before processing. To prevent accumulation of such nominally low-level waste the definition of "scrap" must be set low enough to discourage discarding in- convenient forms as "waste" — a conclusion which runs counter to the present trend. In fact, the economic penalty of actinide wastes may have to be in- creased artificially (or a bonus established for recovering scrap) at the taxpayers' expense in order to insure that commercial power plants will ac- cept the modest penalties incurred by recycling, as projected by the PNL study. 235 E. In present-day U-fueled reactors the buildup of higher actinides is due mostly to the presence of ^-* 8 U as a fertile material. If isotope separa- tion becomes dramatically cheaper in the near future, it might be useful to compare the economics of highly-enriched uranium burners with the possible cost reduction in managing the higher-actinide waste. The short-terra effect would be to delay the buildup of higher actinides until the technology for managing them becomes economically competive. 232 238 F. It is often forgotten that Th and U are probably the only long-lived actinides which do not have finite critical masses. In other words, it is quite possible that an assembly fueled by the higher actinides alone could be made to go critical and consume the undesirable nuclides with a bonus of additional power production. The spectrum would probably have to be much harder than contemplated for current designs of fast breeder, and the de- layed-neutron fraction does not appear easy to predict, so control might be more difficult than envisioned in present fast-reactor designs. However, as a "waste burner" which is not required to make economical power, such a reactor could be sited in a remote area and optimized for safety rather than electrical efficiency. The total thermal power of enough burners to consume the actinide waste of an equilibrium recycle economy would be in the range of a few to 10 percent of the thermal power in the nuclear-power industry as a whole, in direct proportion to the fraction of higher actinides in the overall actinide inventory. The extremely hard spectrum at reasonably high flux might also be attractive for testing components and as a neutnn source for other measurements. G. As long as the delayed-neutron fraction is sufficient to provide a clear margin between delayed and prompt critical such a burner should be control- lable. To increase the margin of control, it might be worth returning to the slow-fast coupled-reactor concept. A thermally-critical region with its characteristically long generation time would be used to dictate the response time of the entire assembly, which would be subcritical without the thermal region, but still have a very hard spectrum in the burner region. Similar 9.E.12 designs have also been studied for the thorium thermal breeder, in which power level (and neutron economy) was to be controlled by adjusting the coupling (neutron leakage) between the regions, rather than by parasitic absorption. The multiplying blanket might be either inside or outside the thermal seed, or merely adjacent to it. H. A still safer design would involve a subcritical assembly fueled with the unwanted actinides and driven by a high-power proton accelerator. This variation resembles one cited by Claiborne. A very rough study- at LASL suggests that k = 0.97, so that (1-k) -1 = 30, is an appropriate value for the actinide assembly. An efficient accelerator of the LAMPF design would produce neutrons by spallation directly in an actinide target at a cost of 50-100 MeV/n. Realistic efficiency estimates (see Section VII) suggest that electricity generated by steam from the multiplying region at this rather modest multiplication would just supply the required 100 MeV/n in the accelerator beam.. Control would be achieved by adjusting the ion. source of the accelerator. Conservative cost estimates, assuming that production-line accelerators would cost half as much as the prototype, yield a capital cost of about 20% of the capital cost of the power reac- tors serviced by the actinide burner. The possibility of offsetting un- expectedly high energy costs by higher multiplication in the burner region is obvious. Similar designs have recently been studied for plutoniura breeding and as neutron sources for experiments. IV. FISSION PRODUCT RECYCLE IN FISSION REACTORS A. We concur with the pessimistic PNL view on transmutation of fission products in fission reactors. However, we are not satisfied that the study to date has conclusively disproved the feasibility for all prominent fission products, Studies done originally for the prediction of reactor after-heat have dis- closed many complex and unexpected transmutation chains. We suspect that chemical and isotopic partitioning of the fission-product waste will be re- quired for useful lowering of hazard index. B. In particular, thermal radiative capture in the rare earths from Ce through Dy is clearly unprofitable, since most of the products are no less hazard- ous than the targets. Placing Sm in an epithermal flux might deserve further study, but the yield of this nuclide is small. From the arguments given in Section II-D, even thermal transmutation might be useful on the fis- sion products with half lives greater than 10^ y. It is not clear whether these were studied by PNL. C. PNL appears to have limited consideration of sources of troublesome radionuclides to the fuel itself. Neutron interactions with cladding and structural materials can also produce long-lived products such as °^Nb and °3z r . Zr-clad fuel elements at typical burnup, for example, Q 6 will have had about 0.5% of the cladding converted to ^ J Zr. 9.E.13 V. TRANSMUTATION IN FUSION REACTORS A. CTRs are expected to be relatively "clean," in the sense that their burden of radionuclides will be much smaller than in fission reactors. They ap- pear attractive for siting near population centers for this reason. Use of such reactors to transmute large loads of radioactive waste from fission reactors would greatly increase the hazard to the population, relative to the hazards associated \^ith underground storage. As with fission reactors, the parasitic blanket of unwanted nuclides competes with breeding — in this case tritium — which may prevent usefully fast destruction of the waste nuclides. B. Useful destruction of nuclear waste by plasma devices need not wait for commercial power-producting CTRs. Dense-plasraa-f ocus devices at present produce fluxes too small to be attractive, but seem likely to be capable of being scaled up before self-sustaining CTRs become available. Laser- pellet concepts may also prove useful in producing a non-self-sustaining plasma device for transmutation applications. VI. TRANSMUTATION USING NUCLEAR EXPLOSIONS AS A NEUTRON SOURCE A. We feel that this alternative has been discarded prematurely. The basic PNL assumptions were so far from optimum that the conclusions were unreasonably pessimistic. Most questionable are the assumptions that 5 kt is an appro- priate size of device and that transmutation would be limited to one cycle. Studies at LASL suggest 100-kt devices at a depth of 1.5 km as an attractive base case. The ground motion would be quite acceptable at this depth. The cost of drilling holes much larger than 60 cm is not prohibitive. The cost has been estimated at roughly 500 k$ for the device and an equal amount for site preparation (if the cost is prorated over^ several explosions). The de- vice can be relatively clean, x^ith a small fission yield. The cavity can probably be reused repeatedly, with the products from each shot being re- covered, partitioned, and recycled. The above parameters lead to a cost of about $5j30(ymole of fission products transmuted. If we set out to transmute the worst 10% of the fission products from reactors, the cost of transmutation comes to 0.4% of the value of the electricity produced by the reactors. Thus, ion I^S 99 this technique looks quite promising for I, Cs, and Tc. B. An old Plowshare concept is again under study, aimed at producing economical electric power from steam using underground explosions as the heat source. If this concept can be combined with burning the most offensive fission-reactor products, the process may prove to be very attractive VII. TRANSMUTATION BY CHARGED- PARTI CLE ACCELERATORS A. We have described in Section III-H an accelerator-driven barely-subcritical actinide burner which looks more attractive than any purely accelerator- induced process. We have also referred briefly in Section V-B to rum— self- sustaining plasma devices which might loosely be classed as accelerator 9.E.14 In addition, we feel that the case against direct accelerator transmutation is enough weaker than suggested by PNL to justify further exploration before abondoning this approach. B. The initial conclusion of the PNL study is, of course, clearly correct. Charged particles of a few 10' s of MeV show no promise at all. C. We offer two comments on specific details of the discussion of spallation- neutron sources. The tritium production in the Be moderator proposed for ING is a bonus, not a penalty. Tritium breeding is the heart of first-gen- eration CTR concepts. Also, the penalty attributed to radionuclides produced in the spallation target is overstated, by an amount that remains to be ex- plored. While it is true that roughly 7 radionuclides would be produced per 19 fission-product nuclei consumed, under the basic assumptions of transmu- tation waste-management only those new nuclides with half-lives greater than those of the waste to be transmuted can be counted as additional nuisances. The entire strategy is based on reducing the half lives of the waste nuclides. D. Spallation itself is a rather interesting form of transmutation. Studies with an intranuclear-cascade-plus-evaporation code show that the product nuclei may range from the target nucleus down to about half the original mass for proton or neutron energies of about 1 GeV. Unlike fission products, spalla- tion products are predominantly on the neutron-poor side of the valley of sta- bility, because the Coulomb barrier inhibits the evaporation of charged par- ticles. Although most codes omit the fission channel, it is certain that di- rect bombardment of actinide targets above 1 GeV leads mostly to fission, which is the desired effect. If the proton energy is high enough, the yield is probably close to one fission per incident proton, plus still-uncalculated further transmutations from secondary particles. We shall outline below a brief calculation which suggests that this is too low a yield to be attrac- tive. E. In computing the energy cost of spallation-induced transmutation, the PNL study reaches a lower limit by assuming that 100% of the input electrical power can be converted to beam power. Surprisingly, this is much less un- realistic than it seems. LAMPF produces 0.8 MW of beam power at about 15% overall efficiency, which could probably be raised to about 50% without stretching current technology unreasonably. LAMPF cost about $4/W for the accelerator alone. This figure could be cut by a factor of 2 to 10 in mass production. F. The PNL study gives almost no consideration to direct spallation of radio- active waste by protons in the GeV range. The Japanese proposal to use 10-GeV protons on 1-^'Cs is mentioned briefly in an appendix. From Monte Carlo cal- culations at 0.8 GeV, we extrapolate a cost of about 50 MeV/nucleon emitted, which implies about 200 new nucleons per incident proton at 10 GeV. In an infinite Cs target, 100% of the neutrons and a substantial fraction of the protons will transmute -*-^'Cs nuclei Into products which have either a shorter half life or a lower specific activity than the target. Radiative capture of either a neutron or a proton, and most reactions which emit one nucleon, lead directly or quickly to stable nuclei. Unfortunately, even an optimistic calculatic: .hows that the 50 MeV/ transmutation required is 1.5% 9.E.15 of the thermal power generated in the reactor which produced the Cs, or about 10% of the original power when realistic efficiencies are included. The terawatt nuclear power economy envisioned for the pinnacle of the fission-power age will generate about 1 kg/s of 13 'Cs. For the 65-^A ion sources proposed for the Canadian Intense Neutron Generator the yield at 10 GeV would only be of the order of 1.5 kg/d, so that 5 x 10^ acceler- ators would be needed. VIII. EXOTIC FORMS OF TRANSMUTATION 8S We suggest that, for transmutation of Kr by raising it to the shorter-lived excited state, inelastic scattering of 14-MeV neutrons from a CTR or other d-t device would have a much higher yield than Coulomb excitation. IX. COMMENTS ON THE PROPOSED FUTURE PROGRAM A. We have remarked above that certain particular topics deserve a more thorough study. We feel that the present study has dismissed a few possi- bilities prematurely. Of these, accelerator-driven subcritical burning of a higher-actinide core and transmutation of partitioned fission products by underground explosions look particularly promising. B. It is our feeling that the time allotted to development of transport methods (Task 3.0) is grossly overestimated. Adequate codes already exist, and are estimated to require only a few man-years to collect and adapt. The time allocated for evaluation of nuclear data is more realistic. We question however, the omission of a modest program of new microscopic measurements, primarily on selected fission products, in tandem with some additional de- velopment of nuclear models for fission yields, decay systematics, and cross sections. X. COMMENTS ON THE PRESENTATION A. Many of our contributors found the present draft rather repetitious. Most of them thought that at least parts of it give insufficient detail to make clear how much exploratory work has actually been done. If the detail of exposition matches the detail of investigation, parts of the study are too shallow to meet the weeding-out goal of Phase 1. B. The draft is replete with minor grammatical errors. C. We have found that the symbol MT is more likely to be interpreted as megaton than as metric ton. In the Systeme Internationale 1000 kg = 1 t. The cor- responding English unit is properly' designated 1 short ton (sh tn) . Simi- larly, there are other sloppy uses of units, such as KW for kW. 9.E.16 Review Comments on BNWL-B-301, Sec. 9: Advan ced Waste Management - Transmutation Processing Dr. Robert L. Hellens January 17, 1974 I enjoyed very much participating in the review of your draft report and in the discussions which took place subsequently in Seattle. The group, no doubt, came to some specific conclusions before the end of the meeting, and I gather that my comments during the discussion were recorded. In summary, I would like to emphasize several points. 1. Section 9 of your report does not go far enough in distinguishing which actinides and which parts of the chains are the main contributors to the long-term hazard. As Bernie Spinrad pointed out, it is suggestive that Claiborne's hazard curves showed a marked change with the amount of U and Pu separated from the waste stream. Unfortunately, this can be interpreted in a variety of ways and I think this needs to be spelled out much more clearly in your review of the ORNL work. 2. I believe that actinide recycle is feasible in power reactors but the anticipated gains need to be made more explicit in your report by the removal of some of the ambiguities which are mentioned in my notes and which were also empha- sized by other members of the panel. I don't think that it has been shown that adequate gains could be made in hazard reduction in a practical application to warrant the appreciable effort which would be involved. If it should turn out to be an efficient approach, I would suggest that you devote more effort, in the pro- posed program, to the evaluation of methods of inserting actinides into a power reactor and to the evaluation of the changes in reactor operation, safety con- siderations and in reactor maintenance operations which would result. 3. The concern expressed by Claude Poncelet regarding the distribution of actinides among the number of power reactors could be substantially allayed if it can be shown that at the time of irradiation the hazards arising from the actinides are substantially smaller, even with recycle, than those normally arising from the 9.E.17 2. plutonium and fission products in the fuel. I suspect that the latter might well be the case and this would strengthen your proposed solution to the problem in an important area. If I can be of any further assistance to you in connection with this study, particularly in those aspects which are concerned with power reactor utilization, please let me know and I will be more than happy to assist insofar as I can. January 14, 1974 Time limitations have allowed me to read with care only those sections of the report that deal with the recycle of actinides and fission products through light-water reactors. Since the report views actinide recycle as the most promising form of waste hazard reduction by transmutation, and, in fact, speaks of this approach as both feasible and economic, concentration on this area seems warranted. The concept is not new, having been studied some years ago at BNL and more recently in ORNL, but the idea of using a power reactor as an irradiation facility for radio- active wastes is unconventional, to say the least, from the point of view of the reactor designer. I suspect that the Brookhaven concept grew up around the HFBR and en- visioned a special purpose reactor devoted to waste hazard reduction; in such a reactor, the energy and waste balance (p. 9. 3) would have to be very favorable and my intuitive feeling tends to favor looking for such an optimized "garbage incinerator" to avoid the dispersion of wastes among many utility-operated power reactors. 1. General Comments on Report Form On the whole, I found the report well organized and relatively easy to read. There are some ambiguities in language which can be attributed to either haste or too much familiarity with the subject on the part of the writers, (for example, p. 9. 30 on the distinctions between the 99. 5% and 99. 9% recovery curves of Claiborne is very unclear - and quite unnecessarily so). Beyond these normal difficulties, the report stops short of distinguishing between the various chemical 9.E.18 3. elements whose isotopes produce the various components of the net waste hazard. It may well be that further chemical separations, so that smaller but more toxic components could be subjected to higher flux levels, would not aid in reducing waste levels, but the question is, I believe, worthy of comment for the audience familiar with reactors but not with the hazards of actinide components; this latter audience comprises most reactor design experts. The discussion of placement of the actinides in the core is limited to mix- ture with either uranium or mixed-oxide fuel. This approach would be considered with great reluctance by most reactor designers because of (i) the importance of achieving rated power and any change which could be prejudicial to the fuel per- formance could not be tolerated; (ii) inclusion of radioactive materials in the fuel, such as Pu, is expected for some time to increase the rod fabrication cost ^•100% until pellet manufacturing and fuel rod loading is automated; (iii) the reprocessing of fuel containing unusually high actinide content may well increase the costs for the large masses of fuel in the core to prohibitive levels. With these points in mind, one is led to consider the use of target rods composed, for example, of uranium tailings or inert diluents and high actinide loadings. Some appraisal of the core rod fraction necessary and the resulting actinide mass evolution (certainly larger than in the uniform dispersion case) would be helpful in providing a rough limiting case which would be viewed as more accept- able in not interfering with the fuel performance. It might also prove to produce so small a reduction in waste hazard that the approach through irradiation in power reactors could be eliminated from consideration. Development Program At first sight, the calendar length of the development program seems excessively long, 15 years, in view of the Figure 9. 11, which shows that if the program started in 1975 we could expect to have 500 Te of actinide inventory on hand to deal with by the completion of the program. At that point the rate onstruction of fission reactors should be nearing its maximum value and the 9.E.19 4. problem would seem to be well out of hand. I can accept the possibility that a reactor can reduce its own output of actinide waste without serious redesign of the core, added safety and maintenance problems, etc. , but dealing with an in- itial inventory of this size in addition to recycle tends to prejudice the evaluation of the approach. It would help in evaluating the program plan to have this question of timing gone into in more careful detail. As mentioned in the text, improved separation of the heavy metals from the reprocessing waste stream can result in substantial gains in the waste hazard reduction. Is enough time devoted to this question in the program? Perhaps the technology is already available, but no clue as to its state seems to be given in this volume of the report. In the safety evaluation of the actinide recycle, the impact of the increased loading of radioactive heavy metals in the core on the inspection, maintenance, repair and possible carry-over into the steam side of the plant should be addressed under conditions of assumed failures of the target rods. It may be possible simply to show that compared to the hazards and/or containination allowed from failed uranium or plutonium recycle rods, the actinide target rods can be expected to release so much less additional toxic material that the problem is second order either because of the number of rods, their low probability of failure, or because of their, content. If the content is more hazardous than a Pu rod, one would have to consider a common tendency for failure of target rods and this would entail an extensive target rod test program which is expensive and lengthy. These comments are all predicated on a mode of operation in which heavy load- ings of actinides are incorporated in a small number of displaced fuel or poison rod locations so that fuel and actinide rods are entirely separated. In reviewing the Research and Developtnent Program Schedule, Figure 9. 13, it appears to me that the irradiation tests performed in the fifth year are too late and too brief to serve the required purpose of demonstrating that irradiation in typical flux levels is an efficacious means of reducing the actinide hazards by significant margins. The program should demonstrate that inventory 9.E.20 5. rise rates are indeed as estimated by calculations, as in Table 9.4. Thus, if it is at all feasible, the irradiations should start, say, in the second year of the program and continue for at least six years with typical fresh loadings of actinides being introduced along with the already irradiated material to support the theoretical contention of the approach to an equilibrium content in a fixed number of target rods in a given reactor. During the irradiation program, the test rods should be measured for re- activity wo-rth in a facility like the ARMF relative to boron glass standards, to ensure that no unforeseen problems of uncertainties in the actinide chain struc- ture, decay constants, cross sections, competition between various (n, x) re- actions, etc. , are present. Measurements should be made at the end of each irradiation on a time schedule guided by calculations so that the observed re- activity worth can be translated back to the value applicable during the irradiation if short half-life components are present with appreciable content and cross sec- tions. After the recycled actinides have been mixed with the fresh component, selected to represent the output from a processed uranium or Pu recycle batch, a second set of reactivity worth measurements needs to be made to provide a check for calculations of the isotopic and reactivity worth evolutions during the next cycle. In my opinion, the reactivity worth measurements and accompanying calculations are vital to the success of the intent of the program if the use of power reactors is seriously considered as a source of neutrons for the irradia- tion and subsequent reduction of the actinide inventory. It would be quite dif- ficult to persuade reactor vendors or utility operators to incorporate such re- cycle rods in a core without some assurance that the perturbation of the fuel enrichments, the cycle length, and refueling interval were well-known. In addi- tion, the influence of such rods on core power pattern would need to be pre- dicted with reasonable assurance to satisfy both operators and AEC regulatory staff that plant down-rating would not occur. It could be argued that the basic 9.E.21 6. measurements mentioned above would be made unnecessary by the demonstration irradiations scheduled to start in Year 12 and extending into Year 15. However, this phase is really one of final proof- testing , and by this point all questions of feasibility, efficiency, degree of hazard reduction, value of process, should have been disposed of at a much earlier stage. There are two other types of measurements which I feel should be incor- porated at an early stage in conjunction with the irradiation sequence mentioned above for the years of roughly 2-8. The first are measurements of the activi- ties of the actinide pellets which contribute to the radioactive hazard at the begin- ning and end of each of irradiation cycles. Admittedly, these measurements would yield activities far different from those of concern in the long term, judging from Figure 9. C. 14, but they could possibly provide data for confirmation that the calculated reductions shown in Figures 9. C. 14 and 9. C. 15 could be expected and that the process is efficient in hazard reduction. The possible value of such measurements should at least be examined. The second type of confirmatory measurement that would have fairly obvious advantages in showing that the trans- formations induced by neutrons and decay among the actinides are properly under- stood is mass spectrometer determinations of isotopic populations. The area of isotopic measurements with a large conglomeration of radioactive nuclei of closely spaced mass values may be difficult to deal with, but judging from the Yankee-Rowe data the measurements are possible and would provide additional substance to the actinide recycle data being generated in this program. Because of the recycle the population of remote isotopes, which do not figure prominently in uranium or even Pu recycle cores, should become large and the verification of the predictions concerning the content and reactivity worth of these constituents should be a primary goal of the development program outlined in this report. Having mentioned these additions to the proposed development program, I recognize that they could be very costly extras and would require more ex- tensive justification for inclusion than given above. It seems to me that the 9.E.22 7. program would be proceeding on a rather weak foundation without support of an experimental nature to show beyond reasonable doubt that the theoretical ex- pectations would be achieved in practice. Before discussing some reservations I have on the reactor design impli- cations of actinide recycle, I would like to reinforce the comments made in the report (p. 9. 54) concerning the development of strategies for insertion of recycled actinides into the reactor without appreciably degrading the power -producing capability of the plant. Of course, the lowest actinide level should be achieved in a uniform mixture of fuel and waste. If this configuration can be achieved in an economical fashion at all, it will probably have to be deferred until the time when all fabrication processes are remote and when contamination of the pro- duction line by highly toxic material can be accepted. This has obvious impact on the conditions for maintenance, repair, inspection, clean-up, etc., opera- tion required on the production line. Since the bulk of fuel fabrication until 1990 will probably be uranium, there is not a high incentive for complete automation of the fuel production line at a much earlier stage, unless it can be shown to be cheaper. Consequently, an approach using target rods in the unused control rod locations of PWR cores is most attractive. The actinide content of the rods will need to be higher, no fuel can be included since little cooling is provided in these channels, and the depletion of actinides will be reduced due to flux self-shielding. The whole fuel management pattern of the core will probably be strongly affected by the presence of the target rods and their change in worth during irradiation can roll the power enough during one fuel cycle to degrade the plant operation unless their charac- teristics are quite accurately predictable. The actinide content from roughly 50,000 - 60,000 fuel rods will have to be inserted into roughly Z000 rods of slightly smaller diameter in a (rod cluster control) RCC core or into about 400 rods of larger diameter in a Combustion Engineering PWR core. Thus, the concentration of actinides should be increased by about a factor of 40 for each 9.E.23 fuel batch contribution when transferred from the fuel to the target rods. The increase sounds large, but when the relatively small reactivity worth of the actinides, other than U and Pu, is recalled, one suspects that the target rods would not be very black, particularly if the actinide oxides are mixed with a low cross section diluent such as AI2O3 or MgO. As later actinide wastes are cycled into the target rods, the blackness will increase and flux depression will occur, with a corresponding loss in the efficiency of recycle to reduce the waste hazard level. How far these various effects will change the results of earlier calculations will need to be evaluated in detailed fuel management calculations in at least two-dimension to see whether stable fuel management patterns can be found with acceptable power distributions. One further point should be mentioned in connection with the large number of 1200 MWe plants expected to be commissioned in the early 1980's. If the pro- posed recycle program is well along, these reactors could start reducing the backlog of wastes in first-core loadings where no previous inventory from the particular reactor itself needs to be included. The difficulty is that data from the proposed development program would need to be available to the core designers well in advance of the start of waste recycle so that the planned loadings can be reflected in uranium enrichment orders and in fuel loading studies. Once again it strikes me that the program, as outlined, is too slow to attack the waste build- up during the time when large quantities of neutron flux are becoming increasing- ly available and the buildup of waste is proceeding at an accelerating rate. The last comment on the development program, as outlined in the report, is that the most serious problem of actinide production has not been specifically noted. This, I believe, occurs in the plutonium recycle mode of fueling for thermal reactors since the contents of Am and Cm are substantially increased by the repeated irradiation of the various Pu isotopes. It would be interesting to see Curves 9. 8 and 9. 9 redrawn for a reactor operating in the SGR (self- generating recycle) mode- -perhaps these have already been generated but have 9.E.24 not been included in the report. If the uranium and SGR modes do indeed have different actinide hazard consequences, the measurements program connected with the experimental sequential irradiations may have to be substantially ex- panded to demonstrate that the actinide recycle scheme does, in fact, provide sufficient reduction to justify its full development and use in power reactors. 3. Comments on Appendix C This appendix provides some additional detail to that given in earlier sections of the report, but not quite enough to be as useful in forming an opinion as it might be. The PNL inventories, using only 10% of the fuel rods rather than all, are useful in showing the effects of higher actinide densities in the rods and deviations from Claiborne's calculations. However, the choice of fuel enrich- ments, both uranium and later plutonium, to provide the same energy output from a target rod as from a normal fuel rod is not a particularly good or valid way of estimating enrichment penalties for recycle in a full core. Nor is the idea, as I have mentioned before, of mixing fuel with recycle waste a very desirable choice, The section (p. 9. C. 23) dealing with the recycle of actinides with plu- tonium fuel is not very clear in the presentation of the recycle schemes, either for the plutonium or for the actinides. However, the weakest point is that the primary emphasis seems to be on the cost increment of waste recycle, rather than on the very'large increase in the actinide hazard associated with Pu recycle. Seeing this increase in Table C8, one wonders first if this is representative of the entire core or only the actinide loaded Pu rods and, second, what impact these very large increases would have on the amplitude and shape of the hazard curves 9. 8 and 9. 9. To my mind, an evaluation of this sort could have an im- • portant bearing on the choice of this program as a vital part of the waste man- agement scheme. In early calculations of plutonium recycle, we have observed the marked increase of the Am and Cm isotopes on the reactivity loss at end of cycle. The 9.E.25 10. calculations were primarily aimed at evaluations of some of the low mass actinide concentrations. The attached table of the reactivity worth of various actinides from cell calculations, though old, is probably indicative of the mag- nitudes of the changes in reactivity worth to be expected as the mode of refueling is changed. It should be kept in mind that these calculations assume complete removal of all actinides other than U and Pu isotopes in recycling, so that the reactivity loss due to Am and Cm are due to buildup only in the cycle in question. I gather from the tables in the report that very substantial further reactivity losses could be anticipated in later stages of plutonium recycle if the actinides were to be recycled. 4. Conclusions The comments I have made have largely been critical in tone with the intent of improving the perspective provided by this proposal. The importance of actinide waste reduction needs to be emphasized and it appears that it may be most valuable in dealing with wastes generated in the late 1970's through 1990 and beyond by plutonium recycle in thermal reactors. The proposed de- velopment program seems minimal, particularly in the area of the measurements which would demonstrate the validity of the approach. If I were making the proposal, I would suggest two phases. A short first phase would be defined to confirm the need, feasibility and efficiency of actinide recycle in power stations. If the first phase were successful and com- pelling, the second phase should be considerably more substantial in some of the areas I have outlined above than that now proposed. If the problem is serious and the solution effective, the job should be thoroughly done. 9.E.26 REACTIVITY WORTH OF SOME HEAVY ISOTOPES at 33,000 MWd/te Isotope 3.3 w/o U0 ? 1st Gen. SGR Equil. SGR Np-237 0.51 0.14 0.15 Pu-238 0.15 0.03 0.03 Am-241 0.12 0.71 0.89 Am-242 0.01 0.10 0.13 Am-243 0.09 0.89 1.95 Cm-244 0.01 0.19 0.41 Cm-245 0.01 0.10 0.22 TOTAL 0.90 2.16 3.78 U-236 1.04 0.20 0.22 Pu-240 6.27 12.51 14.05 Pu-242 0.34 1.64 3.37 Arthur Kubo 9 - E -27 31 December 1973 COMMENTS ON Advanced Waste Management Studies High Level Waste Disposal Alternatives Section 9: Transmutation Processing BNWL-B-301, August 1973 I. General Comments: 1. Accelerators, Explosives, andCTR's: I agree with your dismal prognosis for accelerators and explosives as potential waste processing units. The prospects of a CTR waste burner appear promising but as yet premature depending upon its own technological break through. In this I do agree that at present a continuous though modest interest be maintained in CTR's to provide coordinated development of both waste management and CTR technologies. 2. Feasibility and Order of Merit: It seems clear what four factors comprise the feasibility criteria; however, it is not so clear what elements comprise the order of merit criteria to be used in comparing transmutation schemes with each other and with other strategies described on pages iii and iv. It is assumed that Section 9.4 through 9.6 generally make up the order of merit criteria. If this assumption is valid, the following suggestions are offered: a. Hazard Index: The reduction in long-term waste toxicity, as measured by the hazard index, using nuclear transmutation will be achieved by improved technology and a more complex fuel cycle. To appropriately use this toxicity measurement, it should properly be bounded on either end. On one extreme, as is pointed out, are the untreated high-level wastes. On the lower extreme no such limit presently exists. Some reasonable lower limit should be identified, either technological or economic. Thus bounded, . 1 . 9.E.28 it makes more meaningful your results of a two-order of magnitude improvement in long-term waste toxicity costing about 0.2 mills per KWHe for actinide recycle in a LWR. b. Environmental Impact: Two impacts of actinide recycle not discussed in detail but deserve more attention are dilution of the high-level wastes due to chemical processing, and changes in the actinide waste forms. The former presumably is discussed in Section 4 (Potential for Waste Partitioning), while the latter rightfully seems appropriate in this present section. Current estimates by ORNL ( Sources of Transuranium Solid Waste and their Influence on the Proposed National Radioactive Waste Repository ,CRNL-TM- 3277, February 2, 1971) indicate about 0.5 w/o of the heavy metal will be lost at both fuel preparation and manufacturing facilities (also the value assumed in Table Cll, p 9.C.31). These losses reappear predominantly as low-level alpha contaminated wastes currently termed alpha-wastes by the AEC. Because of the higher levels of alpha and neutron activity of the trans-plutonium isotopes, these low-level wastes will no longer be as benign and easily handled as a purely plutonium contaminated waste. 3. Hazard Index: The use of the hazard index as defined is a most convenient measure of waste toxicity; however, it is disproportionately sensitive to the fraction of Np237 alpha decay chain present in the waste. Only two of the eleven radioactive daughters of Np237 have specified RCG's; the remainder assume the default values assigned by 10CFR20. This circumstance is not as extreme for the other three alpha decay chains. Future research could be directed towards resolving this 'imbalance' but for the nonce some clarification is needed. . 2 . 9.E.29 4. Useful Actinides: As conceived and if implemented, routine actinide recycle might emerge circa 1990. During the interim, it is projected that Np237 (as a progenitor of Pu238), and possibly the curium and californium isotopes, may become commercially valuable materials. If these prognostications are reasonably true, commercial use of the extractants might partially off-set fuel preparation, manufacturing, and reprocessing costs; reduce neutron radiation difficulties; and modify to some extent core physics and in-core fuel management. This aspect of actinide commercialism certainly does not lessen the ultimate requirement for disposal of the 'useful' actinides nor should it be made pivotal in determining physical feasibility of actinide recycle in fission reactors, but it is a possibility worth investigating. 5. Waste Partitioning: The technological-economic limits of actinide-fission product partitioning should be of major concern and one of the first milestones to the Recycle-Transmutation Program. Some reasonable lower limit (for extraction) should be delineated (see l.a. above) from which competing options can be judged. My studies indicate that long-term toxicity reduction is not as optimistic when applied to LMFBR fuels. This circumstance obtains because of the more plentiful quantities of trans-plutonium isotopes generated in the fast reactor coupled with a poorer extraction efficiency for these isotopes (about 0.99) with current technology. However, waste partitioning is the only currently feasible option expanding alternative and should be considered on its own merits. 6. System Model There is a need to look at waste management on a total system basis. It is quite possible that less than optimal solutions can obtain if too narrow a view of the fuel cycle is framed. 9.E.30 II. Specific Comments: Page Comment 9.12 Heretofore, workers in the field have estimated the final dis- (3-6)* posal cost for high-level wastes at but a few percent of the nuclear fuel cycle cost. Thusly, the exact strategy and timing Note: '*• (lines from top of page inclusive) seemed not as important as it now appears if waste management costs approach 15-20 percent (your estimate) of the fuel cycle cost or as much as 100 percent if more exotic means are needed. 9.27 As is rightly observed, there is incomplete/inconsistent data (18-20) at present to definitively predict recycle in fast reactors. This shortcoming, as is pointed out in Section 9.3.2, should receive major emphasis in future actinide-recycle research programs. 9.32 There is some question if any advantage exists in the short-term (Figure 9.8) by recycling actinides as Figure 9.8 implies. If the fission product hazard for the comparable time period is superimposed, it dominates. Additionally the increased handling problems and neutron generating alpha-wastes makes such near-term savings moot. 9.70 Space disposal competes with nuclear transmutation as a 'now* (11-13) method of waste disposal, albeit more costly. 9.77 & 9.78 It is not entirely clear why the actinide release and attendant environmental insult from a fabrication plant should decrease so dramatically over the current U-Pu fabrication facility. The bulk of the fuel would remain the standard U-Pu fuel, while 9.E.31 the added transuranics tend to be more toxic per unit weight, thereby exacerbating the problem. 9.82 The impact of extracting the actinides from the high-level wastes (1-3) (23-25) on transportation appear (in addition to your discussion): 9.83 (1-9) a. Although the actual waste volume/weight will decrease with actinide extraction, the waste-inert material volume/ weight most possibly will increase. This is due to the more difficult chemical processing necessary to make the actinide- fission product cut. b. The reduction of toxicity due to extracting the actinides from the high-level wastes is almost inconsequential vis-a-vis the fission products on the near-term, in particular during transportation of the wastes. Additionally the neutron source strength does not vary appreciably between the no-recycle and recycle high-level wastes due to the isotope composition of the actinides. c. Recycling actinides will result in a reduction of actinides in the solidified high-level wastes but trace amounts of this contaminant will be present in the fuel manufacturers' solid alpha-wastes. Previously considered alpha-wastes were inconsequential neutron sources (e.g. 3440 n/sec-gm Pu238). Conversely, the heavier trans-plutonium isotopes are extremely active alpha and neutron emitters (e.g. 2.3 X 10' n/sec-gm and 3320 (&) Ci/gm Cm242) . High energy alpha emitters are problematic in handling because of the (ctf,n) reactions (note: (o(,n) reactions in oxides are of the order of 0.5 X 10 4 n/sec-Ci; J. P. Vaane, "Hazards Connected with the Handling of Transuranium Elements • 5 9.E.32 and Methods used for the Protection of Personnel", Actinides Review, 1, 1969, pp 337-370). Presently, alpha-wastes can be shipped using DOT approved containers of negligible shielding. With actinide recycle, this may not be possible. 9*82 The concept of a cycle is not entirely clear. It is inferred (4-17) that a cycle extends from one fueling to the next, much as is the practice in nuclear fuel management. Claiborne uses the total irradiation period as a cycle (three refueling periods for a PWR). Since actinide recycle takes so very long to achieve equilibrium, it seems some distinction should be made. 9.85 (7-13) The comments of the second paragraph appear inconsistent with Tables 9.10 and 9.11 (see previous comment). 9. A. 7 The term *F' is not clearly defined in equation All. 9.C.12 & .16 It is not clear what the heavy lines indicate on the tables, possibly quasi-equilibrium. 9.C.22 Figure C5 has no scale for the abscissa. 9.E.33 i n i v i r s i tv or <> K r. GO N Department of Physics CO 1.1 l.CI ())•' II It KRAI. ARTS EUGENE, OREGON 974°} telephone (code 503) 686-4751 January 8, 197U Dr. Bowen R. Leonard, Jr. Battelle Northwest Richland, Washington 99352 Dear Bo: I approached section 9 of BNWL-B-301 with some hesitancy since it is a massive document. After going through it several times, I find myself comfortable with it. It strikes me as a well done study which should lead to easy management decisions. It seems to me though, that those decisions must be influenced strongly by sections 1 through 8 which I have not seen. Let me first comment on transmutation schemes other than reactor processing of the Actinides . I can find no fault with Appendix A except that I think most energy balance and transmutation rates are stated optimistically. I think that your conclusions could be strengthened if there were any need to strengthen them. On page 9 .20 however, the factor of 25 obtained by comparing 200 MeV per fission with 5000 MeV per proton produced transmutation is overstated if one limits attention to a particular species such as Cs . The appropriate first number should then be 900 MeV/ -*'Cs (page 9 -A. 32) giving a ratio closer to 5- Since you discuss coulomb exitation using °->Kr as an example, it might be worth mentioning that neutron inelastic exitation of the same state by a fission neutron spectrum might be more feasible. One can manage with jl = 3 neutrons , and in either a fission spectrum or a CTR spectrum there should be plenty of those. Also, have you considered the possibility that neutron proximity might significantly alter beta decay rates by angular momentum coupling? (Since you consider Nobel Prize level schemes like GRASERs I thought that I should slip that in.) I wouldn't be surprised if someone eventually defends the proposition that transmutation of hazardous materials in large volume can only be done 9.E.34 Dr. Bowen R. Leonard, Jr. January 8, 197 1 * Page 2 safely with neutral radiation. The surface to volume ratio of material to be processed, and protection of cladding integrity both point to that conclusion. I will pass by Appendix B without comment. Appendix D is incomplete because much more nuclear data, and a better defined CTR is needed to make that section complete. I would regard Appendix D as an interesting possibility which should not be used to distract either the people working on that program or the general public. Everyone "knows" that CTRs will be "clean". Don't spoil that illusion. I think that I would worry some about a CTR loaded with 50 kg of Cs. Now on to the Actinides. I think that the scale of the problem is well stated by Table C 6, p. 9-C.18. The actinide of most significance is clearly ^39p u and it is also one which will persist for a time which is long compared to human records. We clearly have accepted the long term hazard associated with storage of 239pu in our weapons inventory. If we accept the one hazard, does it make all that much sense to devote a rather large effort to eliminate a small part of the total long term hazard? (I suspect it does.) Metallic storage of actinides like weapons are stored should be possible. The heat burden from actinides is probably also small enough to allow salt bed disposal of at least them. The energy balance arguments aren't very convincing for two reasons: l) U can't be worse than say 'Np as a breeding material, 2) from a reactor operations point of view, more time would be lost from a jam on loading an actinide rod than would be lost on a jam with a clean rod. In summary, if I had to make the management decision, before I would recommend proceeding with this program I would spend a large amount of time with sections 1 through 8 of the report. Sincerely yours, Harlan W. Lefevre Professor of Physics HWL/js 9.E.35 Camegie -IVIei lOfl University Nuclear Science and Engineering Division Schenley Park Pittsburgh, Pennsylvania 15213 [412] 621-2600 January 7, 1974 Mr. B. R. Leonard, Jr. Staff Scientist Battelle Northwest Laboratories Richland, Washington 99352 Dear Mr. Leonard: I greatly appreciate this opportunity to review the Battelle study on high level radioactive waste transmutation as documented in the draft report, "Advanced Waste Management Studies - High Level Waste Disposal Alternatives; Section 9: Transmutation Processing". The management of long-lived radioactive wastes -Prom -Fission nuclear power plants is a crucial problem and transmutation offers the only possibility of elimina- tion as opposed to disposal. I therefore highly commend you and your staff for your efforts in this area. I concur with the study's conclusions regarding the lack of technical feasibility for transmutation processes involving the use of accelerators or nuclear explosives. The technical feasibility of burning actinides and long-lived fission products in a fusion system appears to be well established, However, the concept of burning the wastes in central station CTR's seems to me very objectionable, since it would introduce in a relatively clean system high level and long-lived radioactive material. This would negate one of the major advantages of fusion systems and could lead to substantial public health hazards. I would suggest that fusion systems be designed for the sole purpose of burning high level radioactive wastes, thus leading to the possibility of siting at the reprocessing plant and to the possibility of achieving engineering feasibility independent of CTR power plant engineering feasibility. I have strong reservations regarding the desirability and real feasi- bility of actinide recycle in central station fission power reactors. The benefits are limited arid these do not appear to outweigh the added risks accruing to plant operation, as well as the increased difficulties in fuel fabrication and fuel reprocessing, which have real influences of plant re- liability and public health hazards. 9.E.36 Mr. B. R. Leonard Battelle Northwest Laboratories Pape 2 Specific and detailed comments on the draft report follow. 1. The study only considers existing or anticipated devices, whose major purposes are other than waste transmutation. This probably limits both feasibility and neutron source characteristics. For example, a fusion facility could be developed for the specific rmrpose of waste transmutation. Research and development programs should be recommended in this area. Note that engineering feasibility of such a facility could be reache"d sooner or independent of CTR power plant feasibility. 2. The conclusions drawn in the study ^avor those processes, such as actinide recycle in fission reactors, that have relatively immediate, although limited, applications. More emphasis should be nlaced at this time on more comprehensive and effective long-range methods. 3. The concept of burning highly radioactive wastes in central station power plants is questionable. It may be better, as a matter of principle, to decouple radioactive waste transmutation from electrical power generation. That is, the neutrons required for transmutation should not necessarily have to come from processes whose function is to generate electrical power. The public and the power industry might much more readily accept a concept where radioactive wastes would be burned in a facility located at the reprocessing plant, apart from electrical power generation. 4. The recycle of actinides in -Fission power reactors may prove un- acceptable to the electric utility industry and perhans the fuel manufacturers This is essentially based on a cost-benefit argument. - The benefits of actinide recycle in -fission reactors are limited. The arguments advanced in the study to explain the non-feasibility of burning long-lived fission products in fission reactors, that is if they could be burned effectively this would imply they would burn in situ at the time o f uro- duction, are partly applicable to actinides. The study shows that the actinides reach an equilibrium in a recycle scheme after about 15 years. Essentially, given that the neutron fluence nvt is the parameter of importance in transmutation processes, the concept here is to stretch the time of exposure t. However, as depicted in Figures 9.8, 9.9, and 9.11 of the report, the reduction in the hazard index is about a factor of 10, whereas many orders of magnitude are probably necessary to cause a significant hazard reduction. That is, with a continuing increase in central station fission power plants, the total hazard coming from actinides would still be very significant even with actinide recycle. 9.E.37 Mr. B. R. Leonard Battelle Northwest Laboratories Page 3 - With actinide recycle, the actinide inventory at any given time in a power reactor is increased. The necessity for remote fabrication of the recycled fuel rods implies a probable decrease in quality control. These considerations imply additional burdens on both the safety and reliability of the plants, and the prospect of severe licensing problems. Given the limited benefits, these additional problems may make actinide recycle very unappealing to the electric utility industry. Note that the concepts proposed in the study would burden the entire fission reactor economy with actinide recycle. - Actinide recycle would impose relatively severe manufacturing, logistical, and quality control problems on the -Fuel manu- facturing industry. It may be difficult to justify the cost of special actinide recycle fabrication plants. 5. Considering long-range goals, it may be necessary to rely on more than one process for effective transmutation. For example, the wastes could be exposed sequentially to a number of different neutron spectra. 6. The recycle of actinides in uranium and/or plutonium bearing fuel rods has serious implications in terms of fuel manufacturing, quality control and. reliability, safety and operations. Fuel fabrication research and de- velopment requirements may be much more extensive than envisioned in the study. An alternate concept would be to recycle the actinides as specially fabricated poison rods that could be utilized in the same way as the burnable poison rods currently in use in LWR's. This would minimize fabrication cost, increase quality control, and alleviate safety, reliability and licensing concerns. Another concept would be to locate the actinide targets in the reflector of LWR's, which experiences a very high thermal neutron flux, and essentially utilize leakage neutrons. The feasibility of locating targets in the reflector space is, however, open to question. 7. The energy balance argument used in the discussion of accelerator de- vices on pages 9.20, 9.23 and 9.24 of the report, appear incorrect. The energy released in the power reactor where the specific waste products to be transmuted were created is more than 200 Mev/waste. That is, the specific fission product yields and the neutron balance leading to actinide production must be considered. 8. I fail to fully understand the explanation for the differences be- tween Clairborne's results and the PNL results as far as the actinide concen- trations in the recycle scheme are concerned. The actinides are present in relatively small quantities in the fuel and I would not expect spatial self- shielding to be very significant. Certainly it cannot explain a difference of 0.05 in k of the fuel. Are the actinides even worth this much? The dif- 00 ference in Pu concentrations (~5%)are also significant. 9.E.38 Mr. B. R. Leonard Battel le Northwest Laboratories Page 4 9. The enrichment penalty in actinide recycle first mentioned on page 9.34 should he worded as a 4% increase in enrichment, and not a 4% additional U-235 content. Note that this implies a roughly proportional increase in kw/ft. Also, a 4% increase in U-235 enrichment implies a correspondingly larger increase in total additional uranium requirement, as opposed to the comment made on page 9.83 of the report. 10. In recycling actinides in a few rods as opposed to every rod, one must consider the effect on local peaking factor in the fuel assemblies. 11. The results for actinide and Mission product transmutation in fusion systems appear very promising and significant. The reduction by a factor of 3 for Cs-137 (page 9.38) is probably significant, given the relatively short half-life of Cs-137. 12. The strategy of recycling actinides in fission power plants until fusion power plants come on line, and then to feed actinides and selected long-lived fission products to the fusion plants, appears questionable. Since the burning of actinides in fission plants is only partial and since long- lived fission products must be stored in any event, why not also store the actinides and await the development of fusion systems, thus avoiding the problems attendant to recycling in fission power plants. 13. The total cost for the proposed RP 7 D program to develop actinide recycle in fission reactors may be much higher than the $20 million arrived at in the study. Major cost may accrue to the fuel fabrication and re- processing technology R§D. 14. Will government subsidies to the electric utilities have to be considered to make up for the economic cost penalty that actinide recycle in fission power plants would imply, particularly during the demonstration stages? 15. With reference to the research and development requirements, I question the need for methods development. Existing methods are probably more than adequate. On the other hand, uncertainty and/or lack o^ cross section data is probably a much more important consideration. A substantial effort would therefore be required in the area of nuclear data. Verification will not be an easy task, given that the existing data on actinide isotopics from irradiated fuel measurements are not that good and plentiful. 16. The neutronics and fuel management analyses will have to consider all the various engineering design and safety factors, including: effect on kw/ft and hot channel factors; effect on shutdown margin; effect on reactivity coefficients; effect on control rod worth; effect on accident analyses. It is likely that engineering design and safety considerations would dictate the recycle strategy, and not fuel cycle or management considerations. 17. The experiments briefly discussed on page 9.53 of the report in relation to the research and development requirements are not defined or specified in the study. Neutronics experiments are probably not required or 9.E.39 Mr. B. R. Leonard Rattelle Northwest Laboratories Page 5 meaningful. Experiments to directly measure basic cross section data, and irradiation tests in reactor during the demonstration program, are probably the more important requirements. It should be noted that irradiation tests in operating fission nower plants would require extensive and costly design and safety analyses and licensing procedures. 18. Safety may be much more of an issue in actinide recycle than the study makes it to be. In particular, as opposed to the statement made on page 9.85 of the report, it is felt that the risk associated with actinide recycle in fuel elements of central station Mission plants would be greater than that associated with normal reactor operation. 19. With reference to the comment made at the bottom of page 9.60 of the report, there is currently no Pu recycle in LWR's, safety and licensing issues having yet to be resolved. 20. With reference to the fuel cycle strategies discussed in section 9.5 on Capital and Operating Costs, the differentiation between fuel rod fabrica- tion and ^uel assembly fabrication does not come through. That is assemblies, not -Puel rods, are loaded into the core, and individual rods cannot be removed from spent fuel assemblies excent at the reprocessing plant. Thus, if acti- nide bearing fuel rods are fabricated at the reprocessing plant site, the entire fuel assembly may have to be fabricated there. Also, it would not be feasible to expose actinide bearing fuel rods longer than other fuel rods in the assembly in which they are inserted. Individual rods could not be shipped from the reprocessing plant to the nower plant, since they must be fabricated into fuel assemblies. 21. The economics analyses appear favorable. However, there may be hidden cost in added safety and licensing problems and in a potential decrease in plant reliability. 22. It is suggested in the study that the fuel fabrication and reproces- sing nlants be located at the same site. Although there is a great deal o^ merit in this concept, it is questionable whether the limited benefits ac- cruing from actinide recycle would justify modifications of current industry plans for both fuel fabrication and reprocessing plants, given the high capital investments involved. 23. As indicated in the study, the effects of actinide recycle on ac- cidents in fuel processing plants are probably minimal. 24. Transmutation of long-lived fission products in fission reactors indeed appears to be non-feasible. 25. The rationale used on top of page 9.C.23 is not clear. The volume- averaged core burnup at the end of a cycle is about 22,000 MWD/MT. The replacement of the 33,000 MWD/MT fuel with fresh fuel increases the core k _ f to about 1.1, which is higher than the value when the core has an average exposure of 22,000 MWD/MT (k = 1.026 in the study). 9.E.40 Mr. B. R. Leonard Battel le Northwest Laboratories Page 6 26. Recycle of actinides in plutonium bearing rods would probably lead to engineering design and safety problems more severe than for recycle in normal U0_ fuel. 27. The use of novel fuel designs, such as the MgO fuel rods proposed in the study, is questionable because of the many potential problems that can be expected with the use of novel fuel types. 28. The necessity to rely on one or more core regions which drive regions of the core with low k imply very poor power distributions, parti- cularly in PWR's. 29. The concept of burning fission reactor radioactive wastes in a central station CTR plant is not appealing because it introduces actinides and long-lived fission products in a system which otherwise is free of these materials. It would make more sense to develop special facilities based on the fusion process to burn fission reactor wastes. 30. The very encouraging results obtained with the fusion processes indicate that major R§D efforts should be concentrated on these processes, as opposed to fission processes. I look forward to discussing these matters more fully with you and your colleagues and the other reviewers. Sincerely yours, Claude G. Ponce let, Chairman Nuclear Science and Engineering CGP:GB 9.E. 41 HIGH-LEVEL WASTE DISPOSAL ALTERNATIVES A Review by Bernard I. Spinrad GENERAL COMMENTS 1) As is recognized in the study, the case for transmutation processing rests primarily on the ability to partition nuclear wastes. This seems to be self-evident, but the study provides adequate objective confirmation. 2) The Kubo and Rose studies cited, and the draft Section 4 of the BNWL study, make a sufficient case for partitioning as a valid technique and a desirable one. The major advantage on the one hand is the possibility of separating out those fission products of significant activity whose half- lives are less than about 50 years. Particularly in the Kubo and Rose study, it is shown that a variety of techniques which have been proposed and criticized for ultimate disposal of mixed wastes can be successfully justified for 1000-year disposal, during which time the radioactivities of this short-half-life set of nuclides will decay to harmless levels. 3) Given the assumption that the results of partitioning and disposal of "decades-long" half-life fission products can be achieved, transmutation of the actinides becomes extremely important as an indicated technique for disposal of the next most dangerous set of by-products of fission power. 4) Although not part of the review of Section 9, my opinion is that Section 4 is unduly timid with regard to both the potential of various chemical processes for realizing high degrees of separation, and the costs of so 9.E.42 doing. The conservatism seems to arise from the fact that the potential and economics of such processing seems to be based on past and current chemical-engineering practice, rather than on that basic chemistry which permits future engineering to be developed. For processes to be developed, this line of thought does not seem appropriate. 5) I would guess that some further qualitative differences exist between the effect on the biosphere of medium-lived nuclides such as most of the actinides, and of long-lived nuclides among the fission products such as Tc and I 129 . At first glance, I am somewhat surprised at the concern over these latter, and would think that their effects would be yery hard to discern against background radiation effects. An in-depth study as to whether concern over these long-lived nuclides is justified would seem to me appropriate to any follow-on to the current project. At a minimum, studies defining that value of half-life which is a "threshold for concern" ought to be initiated. Clearly, U-235 @ 7xl0 8 yr. is a "safe" radioactivity, U-236 @ 2.4xl0 7 yr. is "safe" in small quantities and U-233 @ 1.6xl0 5 yr. is "unsafe" in large amounts, even setting aside critical ity hazards. 6) Transmutation of actinides within the fission cycle is exactly equivalent to recycle of actinides. I do not think that the term "transmutation" improves the public image of the process, which factor is a significant one if "recycle" is an equally valid describing term. 7) In line with my comment #4 above, I am uneasy about accepting current practice of permitting ca. 0.5% of the Pu to be lost to waste during the reprocessing cycle. I think recovery can be improved without major cost penalty, and in this case there is the additional incentive of diminishing 9.E.43 the possibility of diversion of significant quantities of Pu from the MUF in the waste stream. Thus, I think it's worth a lot more R&D. Even though the current standard 200-fold reduction in Pu content seems to be sufficient to make Am and Cm the major "villains" among the remaining actinide content, I think that any large total amount of Pu sent to ultimate waste would be unacceptable for a large nuclear industry. 8) As a reactor physicist, I am concerned with the elaboration of reactor physics projects in the suggested development program for actinide recycle. There is a lot of room for improvement of cross-section data, but beyond that existing methods and computer programs are adequate for answering such general questions as may be posed (such as, for example, the order of magnitude of cost penalties to the nuclear fuel cycle which actinide recycle would bring about), while detailed questions about particular recycle schemes are best answerable within the context of specific scheme proposals. 9) There are policy questions as to what administrative or legal steps would be effective and appropriate in order to achieve actinide recycle within our nuclear industrial framework. While not a part of the present BNWL study, they should be addressed by AEC in any future follow-on studies. SPECIFIC COMMENTS ON SECTION 9 9.1 - In 9.1.2, it is assumed in the text and illustrated in Figs. 9.2 and 9.3 that partitioning is to be performed after "waste" has been rejected from the chemical reprocessing plant. I have registered skepticism as to the ultimate validity of this assumption. In any case, the assumption is not necessary to the conclusions of this subsection, and ought therefore be edited out. 9.E.44 The part of 9.1.2 labeled "Transmutation Cycles" is unnumbered. I believe it is the most salient part of the subsection, and as such ought to be put into the front of 9.1 as a numbered subsection (perhaps 9.1.1). 9.2 - On page 9.30, the text does not indicate whether Claiborne's strategy recovers residual U and Pu from the waste stream to the same recovery factor from that stream as the other actinides. This uncertainty beclouds subsequent discussion. In fact, I interpret Fig. 9.9 to suggest that its bottom curve is dominated by the hazard of the 0.5% plutonium not removed. On page 9.37, I believe that the last two sentences of 9.2.3 could be much more optimistically and positively worded. There is no question in my mind but that fast reactors are much more efficient actinide burners than are thermal reactors. Indeed, fast breeders would produce far less actinide by-product (at least, trans-plutonic) than thermal reactors, a matter which could be ultimately of great economic advantage to them. Section 9.2.4 seems to me to be correct as regards fission products, but potentially very misleading as regards actinides. The reason is that the s/ery high neutron fluxes cited for CTR devices are not much more effective for actinide destruction than are the much lower fluxes in fission reactors. The important matter is to reduce burnup half-lives to the order of a year or less. The use of "hazard half-life" as a relative figure of merit gives inordinate weight, for the actinides, to processes which burn them up in unnecessarily short periods. In section 9.2.5, I believe care should be taken not to assume that CTR is an established technology. I would add appropriate qualifying comments to 9.E.45 the references to CTR if the subsection remains organized as in the draft. Better would be to segregate the fission and fusion recycle schemes into separate subsections, indicating the independent strategies. The sentence beginning on the bottom of page 9.39 contains an editorial mistake. In Fig. 9.10, why are long-lived FP to be returned to the fuel cycle? Fig. 9.12 is a good illustration for the "mixed" cycle, but the fact that it is such a cycle, or the words "CTR" or "fusion" should be incorporated into title or legend. Another figure should be added to Section 9.2 to illustrate the workings of a pure CTR economy, cutting into the by-products of a preceding fission economy. Summary Comments on 9.2 (a) It is not clear what is the fate of that portion of the Pu not recovered in reprocessing, during that part of the cycle in which "actinides" are recovered. (b) A discussion of the improvement in actinide burnup due to the use of fast, rather than thermal, reactors as burners should be included. The gains are demonstrably substantial, and, even without detailed calculations, the generalities of fast-neutron cross-sections permit a definite and positive statement. (c) Discussion of the potential usefulness of CTR techniques should be introduced by qualifications throughout. Such machines do not exist, and our estimate of how we would cycle wastes through them is consequently 9.E.46 vastly vaguer than is the case for fission machines. Relative reduction of hazard half-life is a poor figure of merit to employ. 9.3 - In Section 9.3.2, I believe that the data evaluation task should be clearly divided into separate subtasks, one for actinides and one for FP. I believe it would be useful and appropriate to recognize that many aspects of the task are already under way in other AEC-supported programs, that coordination with these programs is needed, and that with such coordination the work to be done is likely to be not so enormous. The appropriate AEC programs are, of course, those of the Reactor and Research Divisions. In Section 9.3.3, I recommend the title "System Modeling" over "Methods Development". It is more correct, and does not over-generalize the task which is contemplated. I do not think that the task of "Neutronics Analysis" described in 9.3.4 should take as long as scheduled, nor should it be very costly. Given cross-sections, all the codes which have been validated for reactor design should be routinely useful. I do not think that the review of actinide production data should be included in the task described under 9.3.5. That review seems to me to be an important subtask under "Data Evaluation", being the major validation of the cross-sections adopted. Section 9.3.6 describes a number of activities in "catch-all" fashion. Many of these are not independent jobs, but add-ons to programs which will, for other reasons, be on-going in both industry and government. This thought, and the cost savings which may be achieved by recognizing it, ought to be expressed. 9.E.47 Section 9.3.8 has questions associated with it which are covered in the summary comments belov;. Summary Comments on 9.3 (a) I would be happy if the write-up of the neutronics programs, and particularly those involving large experiments, would be in the spirit of "bowing to the inevitable". As written, it is largely "programming the unnecessary". That is, it ought to be clearly stated that reactor physics is now a well enough based discipline that its predictions, given valid cross-section data, are qualitatively unexceptionable. I recognize that pseudo experts will question this and that some group of reactor physicists who are concerned about their employment will support the question; but I would rather see the issue joined than have it conventionally included in recommendations. (b) As written, the development program is much more elaborate than I think it ought to be. Insufficient attention is paid to work in colla- boration with other AEC programs. As a practical matter, it is conventional to propose "fat" programs, but I here register my objection to this convention. 9.4 - I do not see why the task labeled "demonstration" in Fig. 9.13 could not be in fact the beginning of commercial operation. However, since LMFBR's are likely to be the optimum vehicles for actinide transmutation, and since they are likely not to be in commercial use much before 1990, the total schedule would not be affected by official demonstrations. 9.E.48 9.5 - The high costs associated with remote fuel fabrication are used in 9.5.1. These costs are expected to be subjected to very large reductions, particularly as the LMFBR program progresses into recycling high-burnup Pu in quantity. The uncertainty which this comment illustrates is, however, much reduced by the strategy proposed by BNWL, and the conclusions of the section are therefore valid. For reasons cited above, I suspect that the fabrication cost penalty cited in Table 9.6 is too high. For reasons given in my comments on Appendix C, I also do not think there should be as large an enrichment penalty as is cited. On page 9.66, electricity is priced at 23 mills/kwh. Current base-load costs of generation are ca. 10 mills/kwh for new nuclear plants. On page 9.69, the interesting matter of HTGR use for actinide recycle is brought up. Prospects may be worth further elaboration, or incorporation in the program described in 9.3. Also on page 9.39, mention is made of the possibility of fuel reuse without reprocessing (LWR-LMFBR). I have been mentioning this possibility in public lectures for about 15 years, so it can hardly be described as a recent idea. 9.6 - This is in general a well -reasoned section. Risks from transportation may be minimized if nuclear parks are used for all operations of the nuclear industry, as suggested by Weinberg, or even if operations other than power generation are isolated in parks. As opposed to accident hazards, the milling operations of the nuclear industry routinely increase the potential exposure of the biosphere to 9.E.49 significant quantities of radium. Should recovery and transmutation of Ra be considered? Should mill-tailing practice be compared with actinide disposal practice? This section may be the proper place to discuss the reciprocal impacts of actinide recycle and safeguards on each other. The net results will probably be favorable to both. However, a separate partitioning facility would incur incremental costs for safeguards - both for technological security as represented by analytical facilities, and for plant security. This latter would have to be particularly stringent in consideration of the contained hazards of the facility. Appendix A - There is, of course, a real question of cost feasibility for spallation devices. Even ING, based on promotional cost estimates, has too high a cost per neutron for any uses except research. Appendix C - Comments have already been expressed on the need for removing Pu along with the other actinides. It seems that Claiborne did not contemplate Pu extraction - to which I attribute the (to me) disappointingly low hazard reduction factors of Table C-4. The 1000-year decay time is the touchstone to me, and a factor of 3.5 activity reduction does not seem worthwhile on which to base a whole new technology. Even the factor of ca. 5 in Fig. C-3 is not exciting. The numbers for 99.9% actinide recycle (Table C-5) are more significant, and justify that number as a target. The reasons cited for differences between PNL calculations and those of Claiborne seem reasonable, and give confidence to many of the calculations generated. 9.E.50 I am instinctively skeptical about the results of C-5 with regard to the enrichment penalty to be incurred by incorporation of actinides in fuel. My reason is that the actinides are quite valuable burnable poisons. Their incorporation ought to be credited against reduction of boron and rare-earth requirements in the reference fuel, and I wouldn't be surprised if optimum use resulted in an actual enrichment reduction. It is also likely that insufficient credit was taken for fast fission of the actinides: all the higher ones are much more easily fissionable (lower effective threshold energies and higher cross-sections) than U-238 with fast neutrons. CONCLUDING REMARKS In spite of the detailed criticisms noted above, the report has merit. I would urge that more emphasis should be given to certain recommendations: (a) Follow-up studies on detailed schemes - similar in coverage to Claiborne's work - for actinide recycle in LMFBR's. (b) More complete fuel cycle studies on actinide recycle in LWR's, looking to optimization of cost with actinide recycle. The optimization to be emphasized is that of the whole cycle, rather than any single part. There is a strong tendency for reoptimization to reduce sharply the effect of a cost penalty on a single part of the cycle; and on this basis I expect the cost penalties for actinide recycle which have been derived in this report to be upper limits rather than order-of-magnitude estimates. May I repeat the observation that the recommended R&D program for actinide recycle seems to me to be overly elaborate, requiring perhaps more money and time than is necessary? Consideration should be given to the possible 9.E.51 effect that such a recommendation might in consequence adversely affect the chances of the program be adopted at all. Distr-1 DISTRIBUTION BNWL-1900 Special Distribution No. of Copies OFFSITE 1 AEC Chicago Patent Group 9800 South Cass Avenue Arqonne, IL 60439 4 U.S. Atomic Energy Commission Headquarters 1717 H Street, N.W. Washington, DC 20545 Public Reading Room 2 U.S. Atomic Energy Commission Advisory Committee on Reactor Safeguards Washington, DC 20545 H . G . 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