DOCSLIB.ORG

Progress in Development of Fluoride Volatility Reprocessing Technology

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
Progress in Development of Fluoride Volatility Reprocessing Technology P1_21 Progress in development of Fluoride volatility reprocessing technology Jan Uhlíř, Martin Mareček and Martin Přeček* Nuclear Research Institute Řež plc, CZ-250 68 Husinec – Řež 130, Czech Republic [email protected] *Present address: Oregon State University,Corvallis, Oregon, USA Abstract – Fluoride Volatility Method is based on direct fluorination of powderized spent fuel with fluorine gas in a flame fluorination reactor, where the volatile fluorides (represented mainly by UF6, partially NpF6) are separated from the non-volatile ones (e.g. PuF4, AmF3, CmF3, fluorides of majority of fission products). The method is regarded to be a promising advanced pyrochemical technology mainly for selected spent oxide fuels of advanced LWRs or GEN IV fast reactors. The technology should be chiefly suitable for the reprocessing of advanced oxide fuels with inert matrixes of very high burn-up and short cooling time, which can be hardly reprocessed by hydrometallurgical technologies. The current research and development work in the area of Fluoride Volatility Method is focused to the experimental program carried out at the semi-technological line called FERDA placed in radiochemical labs of the Nuclear Research Institute Řež plc. INTRODUCTION reprocessing of spent fuel from thermal reactors. They are caused, for example, by higher burn-up, Fluoride Volatility Method (FVM) is regarded as shorter cooling time resulting in a higher amount a promising advanced pyrochemical reprocessing of energy released by the fuel, higher technology, which can be used for reprocessing concentration and amount of plutonium, different mainly of oxide spent fuels coming from future cladding material, presence of metallic sodium, LWRs or GEN IV fast reactors. The technology different fission products composition etc. should be chiefly suitable for the reprocessing of Therefore, countries that were planning the advanced oxide fuel types e.g. fuels with inert introduction of fast reactors also were attempting matrixes and/or fuels of very high burn-up, high to develop suitable methods for reprocessing, content of plutonium and very short cooling because the industrial hydrometallurgical time, which can be hardly reprocessed by PUREX process, employing organic extractants hydrometallurgical technologies. and solvents, was not suited for the fast reactor spent fuel. Hence, the most intensive effort in the Fluoride Volatility Method is based on a development of the FVM was in 1960s, 1970s separation process, which comes out from the and in 1980s together with the development of specific property of uranium, neptunium and fast breeder reactors. [1, 2, 3, 4] plutonium of forming volatile hexafluorides whereas most of fission products (lanthanides) Nowadays, the renewed interest in the dry – and higher transplutonium elements present in pyrochemical reprocessinging methods is irradiated fuel form non-volatile trifluorides. motivated by requirements on the development of advanced fuel cycle technologies devoted to Original intentions of the development of this the planed Gen IV nuclear reactor systems. dry reprocessing method of spent fuel were Among the pyrochemical technologies under motivated in the past by the assumed commercial present development, the Fluoride Volatility utilization of fast breeder reactors. Their Method is almost the only one, which is not application in the power industry can be based on the use of molten salt techniques. economically efficient only in the case of a closed fuel cycle. However, reprocessing of fast The reprocessing technology based on the reactor spent fuel brings about a number of Fluoride Volatility Method consists of the specific difficulties in comparison to the following main operations: ATALANTE 2008 Montpellier (France) May 19-22, 2008 1 P1_21 1. Removal of the cladding material from spent fuel elements 2. Transformation of the fuel into a powder form of a granulometric composition suitable for the fluorination reaction 3. Fluorination of the fuel (the purpose of this operation is the separation of the uranium component from plutonium, minor actinides and most of fission products) Fig. 1. Process flow-sheet of Fluoride Volatility 4. Purification of the products obtained Method CURRENT STATUS OF R&D ON FVM The first two steps represent preparatory stages for FVM itself and they can be realized The Nuclear Research Institute Řež plc has been separately from the FVM. Whereas the early developing FVM in the frame of fuel cycle fluorination techniques of the FVM were devoted to Molten-Salt Transmutation Reactor. fluidized bed processes, now the direct flame Here, the present day R&D represents the fluorination of powderized fuel is considered as follow-up of the former long-lasting R&D the most promising unit operation for future activities in FBR fuel reprocessing. The current industrial application. This method of R&D program is focused mainly to the fluorination in the frame of the FVM, which was technological verification of individual unit firstly used in 1980s, is now under development operations, apparatuses and material research in the Nuclear Research Institute Řež plc. and to the process control.[5,6,7] The experimental semi-technological line called FERDA, shown in Fig. 2, was manufactured and FVM PROCESS DESCRIPTION placed in the alpha-radiochemical laboratory of the NRI Řež plc. The technology consists of Flame fluorination reaction of a spent oxide fuel flame fluorinator, series of condensers and is a basic unit operation of the whole process. sorption columns and from distillation column. The reaction between the fuel powder and pure The short-run capacity of flame fluorination fluorine gas is spontaneous and highly reactor is 1 – 3 kg of fuel per hour, but the exothermic. Usual temperature of ignition is over process, as a whole, has a batch character. 250 °C. Subsequently the temperature in the flame can reach in the reactor the range of 1500 – 1700 °C. Based on the fluorination reaction, the main partitioning of spent fuel is realized directly in the fluorination reactor. Whereas the volatile products of the fluorination reaction (typically UF6, PuF6, NpF6 and minority of fission products) leave the apparatus, the non- volatile fluorides (majority of fission products and americium and curium fluorides) remain catched in the fluorinator bottom in the form of ash. Further individual separation of uranium, plutonium and neptunium from those fission products forming volatile fluorides is generally possible by sorption, condensation and Fig. 2. Experimental line FERDA for R&D evaporation of volatile fluorides, thermal on FVM decomposition of plutonium hexafluoride and finally by distillation processes. A total process The main present experimental effort is oriented flow-sheet of the Fluoride Volatility Method is to the mastering of the flame fluorination process shown in Fig. 1. and to the elimination of some bottlenecks of the technology. After the tests done with the ATALANTE 2008 Montpellier (France) May 19-22, 2008 2 P1_21 uranium fuel, the present day program is focused REFERENCES to the verification of the main unit operations with simulated spent oxide fuel constituted from 1. J. J. SCHMETS, Atomic Energy Review, a mixture of uranium oxides and non-radioactive 8(1970), 3. oxides representing selected fission products (mostly lanthanides, Cs, Sr etc.). The next series 2. M. BOURGEOIS, B. COCHET-MUCHY, of experiments should verify the suitability of Bulletin d’Informations Scientific et the technology also for reprocessing of oxide Technoques, 161(1971), pp. 41-50. fuels with inert matrixes. 3. M. A. DEMJANOWICH et al., RIAR report No. 50(565), 1982, RIAR Dimitrovgrad, CONCLUSIONS USSR (Russia). The Fluoride Volatility Method has a good 4. P. NOVÝ et al, NRI report No. 9062Ch, potential to be used within the fuel cycles of 1989, NRI Řež, Czech Republic (in Czech). several current or advanced reactor types for reprocessing of current or future oxide fuel 5. M. MAREČEK, F. LISÝ and J. UHLÍŘ, types. The main attractiveness of technology can Technological development of fluoride be found in the reprocessing of fast breeder volatility method for front-end of molten salt reactors spent fuel owing to the extreme transmutation reactor fuel cycle, Proc. radiation resistance of the used chemical agents InWor for P&T and ADS 2003, Mol, (fluorine gas, inorganic fluorides), which allows Belgium, October 6 – 8, 2003. to reprocess the spent fuels after the cooling time 6. J. UHLÍŘ and M. MAREČEK, R&D of of about six months only. Also the attractiveness pyrochemical reprocessing technologies to use FVM as the “Front-end” technology of dedicated to MSR fuel cycle, Proc. of Molten-Salt Transmutation Reactor is high due GLOBAL 2005, Tsukuba, Japan, October 9 – to the conversion of oxide form of original spent 13, 2005. fuel into fluorides – the chemical form of Molten-Salt Reactors. Another attractiveness of 7. J. UHLÍŘ, M. MAREČEK and M.PŘEČEK, the process comes out from the possibility to R&D on fluoride volatility method for process those oxide fuel types, which are reprocessing of LWR and FR oxide-type insufficiently soluble in nitric acid used in fuels, Proc. ICAPP 2007, Nice, France, May PUREX process, like oxide fuels with inert 13 – 18, 2007. matrixes (ZrO2) or with high content of plutonium. Further advantages of the technology come out from the achieved compactness of the whole process and from the exclusion of any moderating agent. However to reach the separation efficiencies similar those, achieved by current hydrometallurgical processes will require further intensive research and development. The research and development of Fluoride Volatility Method is realized in the Nuclear Research Institute Řež plc thanks to the financial support of the Ministry of Industry and Trade of the Czech Republic, the Czech Power Company ČEZ and the Radioactive Waste Repository Authority. The R&D has a close connection to the national Partitioning and Transmutation program. ATALANTE 2008 Montpellier (France) May 19-22, 2008 3 .
Load more

Recommended publications
  • Separation of Fluoride Residue Arising from Fluoride Volatility Recovery of Uranium from Spent Nuclear Fuel
    University of Tennessee, Knoxville TRACE: Tennessee Research and Creative Exchange Masters Theses Graduate School 5-2004 Separation of Fluoride Residue Arising from Fluoride Volatility Recovery of Uranium from Spent Nuclear Fuel Jennifer L. Ladd-Lively University of Tennessee - Knoxville Follow this and additional works at: https://trace.tennessee.edu/utk_gradthes Part of the Chemical Engineering Commons Recommended Citation Ladd-Lively, Jennifer L., "Separation of Fluoride Residue Arising from Fluoride Volatility Recovery of Uranium from Spent Nuclear Fuel. " Master's Thesis, University of Tennessee, 2004. https://trace.tennessee.edu/utk_gradthes/2557 This Thesis is brought to you for free and open access by the Graduate School at TRACE: Tennessee Research and Creative Exchange. It has been accepted for inclusion in Masters Theses by an authorized administrator of TRACE: Tennessee Research and Creative Exchange. For more information, please contact [email protected]. To the Graduate Council: I am submitting herewith a thesis written by Jennifer L. Ladd-Lively entitled "Separation of Fluoride Residue Arising from Fluoride Volatility Recovery of Uranium from Spent Nuclear Fuel." I have examined the final electronic copy of this thesis for form and content and recommend that it be accepted in partial fulfillment of the equirr ements for the degree of Master of Science, with a major in Chemical Engineering. Robert M. Counce, Major Professor We have read this thesis and recommend its acceptance: Barry B. Spencer, Paul Bienkowski, Fred Weber Accepted for the Council: Carolyn R. Hodges Vice Provost and Dean of the Graduate School (Original signatures are on file with official studentecor r ds.) To the Graduate Council: I am submitting herewith a thesis written by Jennifer L.
    [Show full text]
  • Properties of Selected Radioisotopes
    CASE FILE COPY NASA SP-7031 Properties of Selected Radioisotopes A Bibliography PART I: UNCLASSIFIED LITERATURE NATIONAL AERONAUTICS AND SPACE ADMINISTRATION NASA SP-7031 PROPERTIES OF SELECTED RADIOISOTOPES A Bibliography Part I: Unclassified Literature A selection of annotated references to technical papers, journal articles, and books This bibliography was compiled and edited by DALE HARRIS and JOSEPH EPSTEIN Goddard Space Flight Center Greenbelt, Maryland Scientific and Technical Information Division / OFFICE OF TECHNOLOGY UTILIZATION 1968 USP. NATIONAL AERONAUTICS AND SPACE ADMINISTRATION Washington, D.C. PREFACE The increasing interest in the application of substantial quantities of radioisotopes for propulsion, energy conversion, and various other thermal concepts emphasizes a need for the most recent and most accurate information available describing the nuclear, chemical, and physical properties of these isotopes. A substantial amount of progress has been achieved in recent years in refining old and developing new techniques of measurement of the properties quoted, and isotope processing. This has resulted in a broad technological base from which both the material and information about the material is available. Un- fortunately, it has also resulted in a multiplicity of sources so that information and data are either untimely or present properties without adequately identifying the measurement techniques or describing the quality of material used. The purpose of this document is to make available, in a single reference, an annotated bibliography and sets of properties for nine of the more attractive isotopes available for use in power production. Part I contains all the unclassified information that was available in the literature surveyed. Part II is the classified counterpart to Part I.
    [Show full text]
  • System Studies of Fission-Fusion Hybrid Molten Salt Reactors
    University of Tennessee, Knoxville TRACE: Tennessee Research and Creative Exchange Doctoral Dissertations Graduate School 12-2013 SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS Robert D. Woolley University of Tennessee - Knoxville, [email protected] Follow this and additional works at: https://trace.tennessee.edu/utk_graddiss Part of the Nuclear Engineering Commons Recommended Citation Woolley, Robert D., "SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS. " PhD diss., University of Tennessee, 2013. https://trace.tennessee.edu/utk_graddiss/2628 This Dissertation is brought to you for free and open access by the Graduate School at TRACE: Tennessee Research and Creative Exchange. It has been accepted for inclusion in Doctoral Dissertations by an authorized administrator of TRACE: Tennessee Research and Creative Exchange. For more information, please contact [email protected]. To the Graduate Council: I am submitting herewith a dissertation written by Robert D. Woolley entitled "SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS." I have examined the final electronic copy of this dissertation for form and content and recommend that it be accepted in partial fulfillment of the equirr ements for the degree of Doctor of Philosophy, with a major in Nuclear Engineering. Laurence F. Miller, Major Professor We have read this dissertation and recommend its acceptance: Ronald E. Pevey, Arthur E. Ruggles, Robert M. Counce Accepted for the Council: Carolyn R. Hodges Vice Provost and Dean of the Graduate School (Original signatures are on file with official studentecor r ds.) SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS A Dissertation Presented for the Doctor of Philosophy Degree The University of Tennessee, Knoxville Robert D.
    [Show full text]
  • RFI) DE-FOA-0002533 on Cleaning up Radioisotope Enventories (CURIE)
    U.S. Department of Energy Advanced Research Projects Agency – Energy (ARPA-E) Request for Information (RFI) DE-FOA-0002533 on Cleaning Up RadioIsotope Enventories (CURIE) Introduction The purpose of this RFI is to solicit input for a potential future ARPA-E research program focused on the development of technologies that would enable the effective management of the Nation’s commercial used nuclear fuel (UNF). The goals of this RFI are to (1) solicit information about reactor fuel needs for both the current commercial light-water reactor (LWR) fleet and future advanced reactors, and (2) seek insights into technology gaps and/or cost drivers that may be hindering economical recycling of existing LWR UNF.1 This information is needed to help ARPA-E identify ways in which the Nation’s roughly 86,000 MTU2 inventory of UNF, which has been increasing by approximately 2,000 MTU per year, can best be recycled to support current and advanced reactor fuel needs. Such activities are consistent with ARPA-E’s statutory goals, which include supporting the development of transformative solutions for addressing UNF.3 ARPA-E is interested in information about technologies with the potential to make recycling UNF at least as economical, safe, and secure as the current once-through fuel cycle.4 Such technologies would enable a UNF treatment facility to be economically constructed, managed, and operated; yield an actinide product that is cost-competitive with natural uranium (U) obtained from traditional mining and milling; and generate significantly lower
    [Show full text]
  • 12.2% 130,000 155M Top 1% 154 5,300
    We are IntechOpen, the world’s leading publisher of Open Access books Built by scientists, for scientists 5,300 130,000 155M Open access books available International authors and editors Downloads Our authors are among the 154 TOP 1% 12.2% Countries delivered to most cited scientists Contributors from top 500 universities Selection of our books indexed in the Book Citation Index in Web of Science™ Core Collection (BKCI) Interested in publishing with us? Contact [email protected] Numbers displayed above are based on latest data collected. For more information visit www.intechopen.com Chapter Fast-Spectrum Fluoride Molten Salt Reactor (FFMSR) with Ultimately Reduced Radiotoxicity of Nuclear Wastes Yasuo Hirose Abstract A mixture of NaF-KF-UF4 eutectic and NaF-KF-TRUF3 eutectic containing heavy elements as much as 2.8 g/cc makes a fast-spectrum molten salt reactor based upon the U-Pu cycle available without a blanket. It does not object breeding but a stable operation without fissile makeup under practical contingencies. It is highly integrated with online dry chemical processes based on “selective oxide precipita- tion” to create a U-Pu cycle to provide as low as 0.01% leakage of TRU and nominated as the FFMSR. This certifies that the radiotoxicity of HLW for 1500 effective full power days (EFPD) operation can be equivalent to 405 tons of depleted uranium after 500 years cooling without Partition and Transmutation (P&T). A certain amount of U-TRU mixture recovered from LWR spent fuel is loaded after the initial criticality until U-Pu equilibrium but the fixed amount of 238U only thereafter.
    [Show full text]
  • Fuel Cycle Processes Directed Self-Study Course! This Is the Third of Nine Modules Available in This Directed Self-Study Course
    MODULE 3.0: URANIUM CONVERSION Introduction Welcome to Module 3.0 of the Fuel Cycle Processes Directed Self-Study Course! This is the third of nine modules available in this directed self-study course. The purpose of this module is to be able to discuss the NRC regulations of and describe conversion facilities; identify the basic steps of the dry fluoride volatility conversion process and contrast with the wet acid digestion conversion process; identify sampling and measurement activities for the dry conversion process and the radiological and non-radiological hazards associated with the dry conversion process. This self-study module is designed to assist you in accomplishing the learning objectives listed at the beginning of the module. There are five learning objectives in this module. The module has self-check questions and activities to help you assess your understanding of the concepts presented in the module. Before you Begin It is recommended that you have access to the following materials: ◙ Trainee Guide ◙ Sequoyah Fuels Accident Slides (on CD accompanying course manual) ◙ “Release of UF6 from a Ruptured Model 48Y Cylinder at Sequoyah Fuels Corporation Facility: Lessons-Learned Report," U.S. Nuclear Regulatory Commission, NUREG-1198, June 1986. ◙ “Assessment of the Public Health Impact from the Accidental Release of UF6 at the Sequoyah Fuels Corporation Facility at Gore, Oklahoma," U.S. Nuclear Regulatory Commission, NUREG-1189, Vol. 1, March 1986. Complete the following prerequisites: ◙ Module 1.0: Overview of the Nuclear Fuel Cycle How to Complete this Module 1. Review the learning objectives. 2. Read each section within the module in sequential order.
    [Show full text]
  • Estimation of Alkali Metal Mole Percent and Weight of Calcined Solids for ICPP Calcine
    INEL-95/0184 Estimation of Alkali Metal Mole Percent and Weight of Calcined Solids for ICPP Calcine B. H. O'Brien Published March 1995 Idaho National Engineering Laboratory High-Levei Waste Engineering and Project Management Department Lockheed Idaho Technologies Company Idaho Falls, Idaho 83415 Prepared for the U.S. Department of Energy Assistant Secretary for Environmental Management Under DOE Idaho Operations Office Contract DE-AC07-94ID13223 DISTRIBUTION OF THIS DCCl-^T IS UNLIMITED (j$ I DISCLAIMER Portions of this document may be illegible eiectronic image products, images are produced from the best avaiiabie original document. ABSTRACT An updated method is given for estimation of the weight of calcined solids and volume reduction factor for calcine, and mole percent sodium plus potassium in calcine produced from radioactive waste in a fluidized-bed calciner at the Idaho Chemical Processing Plant (ICPP). It incorporates new information on a calcine chemistry from a study by K. N. Brewer and G. F. Kessinger in which they determined the compounds formed during calcination by both high temperature thermodynamic equilibrium calculations and by analyses of pilot-plant calcines. An explanation of the assumptions made in the calculations, along with several example calculations and comparisons with the previous calculation methods are included. This method allows calculation of the heat generation rate and sodium content of the calcine, which are used to determine the suitability of the calcine for storage in the ICPP bin sets. Although this method accurately predicts the weight of calcine and mole percent Na+K for its intended purpose, the compounds predicted should only be used as a first approximation for other purposes since the calculation does not incorporate all of the compounds, such as mixed-metal oxides, which may form during calcination.
    [Show full text]
  • Actinide Separation Chemistry in Nuclear Waste Streams and Materials, December 1997
    NEA/NSC/DOC(97)19 NEA NUCLEAR SCIENCE COMMITTEE ACTINIDE SEPARATION CHEMISTRY IN NUCLEAR WASTE STREAMS AND MATERIALS December 1997 NUCLEAR ENERGY AGENCY ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT FOREWORD The separation of actinide elements from various waste materials is a significant problem facing developed countries. The issue arises primarily because of the potential long-term hazard of many of the actinides, but is also due to the regulatory requirements associated with actinide waste disposal, which are different from those associated with other radioactive wastes. The different regulations are in turn related to the different health hazards and generally longer half lives of the actinides. This issue is of continuing interest to the OECD/NEA Member countries primarily in relation to waste produced within power station fuel recycle. Similar problems exist for waste produced in the past as a result of nuclear weapons production programmes and wastes likely to be produced in the future emerging from operations required for waste disposal. The Nuclear Science Committee, representing the interests of the basic R&D in the OECD/NEA Member countries, has established a task force of experts in actinide chemistry to review the current and developing separation techniques and chemical processes of interest in separating actinides. This task force was commissioned with the preparation of a report on the subject. The task force members are listed in Annex 1. This report is intended to provide a timely and representative guide to technical journals and other sources of separation chemistry information on the actinide elements. An important objective is to provide information that may be used to conserve energy and protect humans and the environment.
    [Show full text]
  • The Study of Alternate, Solid-Phase Fluorinating Agents for Use in Reactive Gas Recycle of Used Nuclear Fuel
    University of South Carolina Scholar Commons Theses and Dissertations 2013 The tudS y of Alternate, Solid-Phase Fluorinating Agents for Use in Reactive Gas Recycle of Used Nuclear Fuel Dillon Inabinett University of South Carolina Follow this and additional works at: https://scholarcommons.sc.edu/etd Part of the Nuclear Engineering Commons Recommended Citation Inabinett, D.(2013). The Study of Alternate, Solid-Phase Fluorinating Agents for Use in Reactive Gas Recycle of Used Nuclear Fuel. (Master's thesis). Retrieved from https://scholarcommons.sc.edu/etd/3587 This Open Access Thesis is brought to you by Scholar Commons. It has been accepted for inclusion in Theses and Dissertations by an authorized administrator of Scholar Commons. For more information, please contact [email protected]. THE STUDY OF ALTERNATE, SOLID-PHASE FLUORINATING AGENTS FOR USE IN REACTIVE GAS RECYCLE OF USED NUCLEAR FUEL by Dillon Inabinett Bachelor of Science University of South Carolina, 2012 Submitted in Partial Fulfillment of the Requirements For the Degree of Master of Science in Nuclear Engineering College of Engineering and Computing University of South Carolina 2013 Accepted by: Travis W. Knight, Major Professor Joshua Gray, Second Reader Lacy Ford, Vice Provost and Dean of Graduate Studies © Copyright by Dillon Inabinett, 2013 All Rights Reserved. ii DEDICATION I would like to dedicate this work to my mother, for her endless support throughout all of my endeavors. Though life may not have always been perfect, the one thing that remains true is her love and affection. iii ACKNOWLEDGEMENTS I would like to thank my two advisors, Dr. Travis Knight and Dr.
    [Show full text]
  • (Msrs): Coupling Spent Fuel Processing and Actinide Burning
    Molten Salt Reactors (MSRs): Coupling Spent Fuel Processing and Actinide Burning Dr. Charles W. Forsberg* Oak Ridge National Laboratory P.O. Box 2008 Oak Ridge, Tennessee 37831-6179 Tel: (865) 574-6783 Fax: (865) 574-9512 E-mail: [email protected] Professor Ehud Greenspan University of California, Berkeley 4107 Etcheverry Berkeley, California 94720-1730 Tel: (510) 643-9983 E-mail: [email protected] Manuscript Date: July 25, 2003 File: MSR.HiltonHead.Combined.2003 Advances in Nuclear Fuel Management III American Nuclear Society Hilton Head, South Carolina October 5–7, 2003 The submitted manuscript has been authored by a contractor of the U.S. Government under contract DE-AC05-00OR22725. Accordingly, the U.S. Government retains a nonexclusive, royalty-free license to publish or reproduce the published form of this contribution, or allow others to do so, for U.S. Government purposes. _________________________ *Oak Ridge National Laboratory, managed by UT-Battelle, LLC, for the U.S. Department of Energy under contract DE-AC05-00OR22725. Advances in Nuclear Fuel Management III (ANFM 2003) Hilton Head Island, South Carolina, USA, October 5–8, 2003, on CD-ROM, American Nuclear Society, La Grange Park, IL (2003) MOLTEN SALT REACTORS (MSRs): COUPLING SPENT FUEL PROCESSING AND ACTINIDE BURNING Charles Forsberg Ehud Greenspan Oak Ridge National Laboratory University of California, Berkeley P.O. Box 2008 4107 Etcheverry Oak Ridge, Tennessee 37831 Berkeley, California 94720-1730 [email protected] [email protected] Keywords: Molten Salt Reactor, Transmutation, Actinide Burning ABSTRACT Molten salt reactors (MSRs) are liquid-fueled reactors that can be used for burning actinides, production of electricity, production of hydrogen, and production of fissile fuels (breeding).
    [Show full text]
  • Identification of Potential Waste Processing and Waste Form Options for Molten Salt Reactors
    Identification of Potential Waste Processing and Waste Form Options for Molten Salt Reactors Prepared for U.S. Department of Energy MSR Campaign B.J. Riley,(a) J. McFarlane,(b) G.D. DelCul,(b) J.D. Vienna,(a) C.I. Contescu,(b) L.M. Hay,(a) A.V. Savino,(a) H.E. Adkins,(a) (a)Pacific Northwest National Laboratory (b)Oak Ridge National Laboratory August 15, 2018 NTRD-MSR-2018-000379, PNNL-27723 Approved for public release. Distribution is unlimited. DISCLAIMER This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed 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. References herein to any specific commercial product, process, or service by trade name, trade mark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof. Identification of Potential Waste Processing and Waste Form Options for Molten Salt Reactors iv August 15, 2018 SUMMARY The overall summary of the waste management envelope discussed in this report is represented by the diagram shown in Figure S1. The
    [Show full text]
  • Nuclear Fuel Cycle Technologies for Future Low-Carbon Society 70 Nuclear Fuel Cycle Technologies for Future Low-Carbon Society
    Nuclear Fuel Cycle Technologies for Future Low-carbon Society 70 Nuclear Fuel Cycle Technologies for Future Low-carbon Society Tetsuo Fukasawa, Dr. Eng. OVERVIEW: Establishing a nuclear fuel cycle that can recover usable fuel Shusaku Sawada material from spent nuclear fuel so that it can be reused in a nuclear reactor Kikuo Okada is essential to meet objectives such as securing the long-term availability of an energy source that emits almost no CO and improving Japan’s very low Masashi Shimizu 2 level of energy self-sufficiency. The Monju prototype fast breeder reactor was restarted on May 6, 2010 and this will be accompanied by the planned completion in the future of the Rokkasho Reprocessing Plant, an interim storage facility for spent fuel, and other associated facilities. Discussions regarding the construction of a second reprocessing plant are expected to start during the 2010 financial year. It is anticipated that a demonstration fast-breeder reactor will be completed around 2025 and the introduction of commercial fast-breeder reactors will take place around 2050 provided economic and other requirements have been satisfied. In addition to its involvement in reprocessing, Monju, interim storage, and other such projects, Hitachi is also working on the development of future technologies such as fast breeder reactors and advanced reprocessing with the aim of creating a low-carbon society in the future. INTRODUCTION energy sources such as photovoltaic and wind power RISING international concerns about climate change also emit no CO2 in the generation process, they are have led to a growing number of countries showing characterized by requiring a large area of land and an interest in nuclear power which can generate being dependent on weather conditions.
    [Show full text]