DOCSLIB.ORG

History, Status, and Potential

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
History, Status, and Potential MOLTEN-SALT REACTORS—HISTORY, STATUS, AND POTENTIAL M.W. ROSENTHAL, P.R. KASTEN, and R.B. BRIGGS Oak Ridge National Laboratory, Oak Ridge, Tennessee 37830 Received August 4, 1969 Revised October 10, 1969 fuel appeared to offer several advantages, so Molten-salt breeder reactors (MSBR's) are experiments to establish the feasibility of molten- being developed by the Oak Ridge National salt fuels were begun in 1947 on “the initiative of Laboratory for generating low-cost power while V.P. Calkins, Kermit Anderson, and E.S. Bettis. At extending the nation's resources of fissionable fuel. the enthusiastic urging of Bettis and on the The fluid fuel in these reactors, consisting of UF4 recommendation of W.R. Grimes, R.C. Briant and ThF4 dissolved in fluorides of beryllium and adopted molten fluoride salts in 1950 as the main lithium, is circulated through a reactor core line effort of the Oak Ridge National Laboratory's moderated by graphite. Technology developments Aircraft Nuclear Propulsion1 program.” The over the past 20 years have culminated in the fluorides appeared particularly appropriate because successful operation of the 8-MWt Molten Salt they have high solubility for uranium, are among Reactor Experiment (MSRE), and have indicated the most stable of chemical compounds, have very that operation with a molten fuel is practical, that low vapor pressure even at red heat, have the salt is stable under reactor conditions, and that reasonably good heat transfer properties, are not corrosion is very low. Processing of the MSRE fuel damaged by radiation, do not react violently with has demonstrated the MSR processing associated air or water, and are inert to some common with high-performance converters. New fuel structural metals. processing methods under development should A small reactor, the Aircraft Reactor permit MSR's to operate as economical breeders. Experiment, was built at Oak Ridge to investigate These features, combined with high thermal the use of molten fluoride fuels for aircraft efficiency (44%) and low primary system pressure, propulsion reactors and particularly to study the give MSR converters and breeders potentially nuclear stability of the circulating fuel system. The favorable economic, fuel utilization, and safety ARE fuel salt was a mixture of NaF, ZrF4, and UF4, characteristics. Further, these reactors can be the moderator was BeO, and all the piping was In- initially fueled with 233U, 235U, or plutonium. The conel. In 1954 the ARE was operated successfully construction cost of an MSBR power plant is for 9 days at steady-state outlet temperatures estimated to be about the same as that of light- ranging up to 1580°F (1133 K) and at powers up to water reactors. This could lend to power costs 2.5 MWt. No mechanical or chemical problems ~0.5 to 1.0 mill/kWh less than those for light-water were encountered, and the reactor was found to be reactors. Achievement of economic molten-salt stable and self-regulating.2 breeder reactors requires the construction and That molten-salt reactors might be attractive for operation of several reactors of increasing size and civilian power applications was recognized from their associated processing plants. the beginning of the ANP program, and in 1956 H.G. MacPherson formed a group to study the technical characteristics, nuclear performance, and THE HISTORY OF MOLTEN-SALT REACTORS economics of molten-salt converters and breeders.3 After considering a number of concepts over a Investigation of molten-salt reactors started in period of several years, MacPherson and his the late 1940's as part of the United States' program associates concluded that graphite-moderated to develop a nuclear powered airplane. A liquid thermal reactors operating on a thorium fuel cycle would be the best molten-salt systems for and that proceeding to the breeder either directly or producing economic power. The thorium fuel via the converter would achieve reactors with good cycle with recycle of 233U was found to give better fuel conservation characteristics. Since many of performance in a molten-salt thermal reactor than a the features of civilian power reactors would differ uranium fuel cycle in which 238U is the fertile from those of the ARE, and the ARE had been material and plutonium is produced and recycled. operated only a short period, another reactor Homogeneous reactors in which the entire core is experiment was needed to investigate some of the liquid salt were rejected because the limited technology for power reactors. moderation by the salt constituents did not appear The design of the Molten-Salt Reactor to make as good a thermal reactor as one moderated Experiment was begun in 1960. A single-fluid by graphite, and intermediate spectrum reactors did reactor was selected that in its engineering features not appear to have high enough breeding ratios to resembled a converter, but the fuel salt did not compensate for their higher inventory of fuel. contain thorium and thus was similar to the fuel salt Studies of fast spectrum molten-salt reactors4,5 for a two-fluid breeder. The MSRE fuel salt is a indicated that good breeding ratios could be mixture of uranium, lithium-7, beryllium, and obtained, but very high power densities would be zirconium fluorides. Unclad graphite serves as the required to avoid excessive fissile inventories. moderator (the salt does not wet graphite and will Adequate power densities appeared difficult to not penetrate into its pores if the pore sizes are achieve without going to novel and untested heat small). All other parts of the system that contact removal methods. salt are made from the nickel-base alloy, INOR-8 Two types of graphite-moderated reactors were (also called Hastelloy-N), which was specially considered by MacPherson’s group—single-fluid developed in the aircraft program for use with reactors in which thorium and uranium are molten fluorides. The maximum power is ~8 MWt, contained in the same salt, and two-fluid reactors in and the heat is rejected to the atmosphere. which a fertile salt containing thorium is kept Construction of the MSRE began in 1962, and separate from the fissile salt which contains the reactor was first critical in 1965. Sustained uranium. The two-fluid reactor had the advantage operation at full power began in December 1966. that it would operate as a breeder; however, the Successful completion of a six-month run in March single-fluid reactor appeared simpler and seemed to of 1968 brought to a close the first phase of offer low power costs, even though the breeding operation during which the initial objectives were ratio would be below 1.0 using the technology of achieved. The molten fluoride fuel was used for that time. The fluoride volatility process6, which many months at temperatures > 1200ºF (920 K) could remove uranium from fluoride salts, had without corrosive attack on the metal and graphite already been demonstrated in the recovery of parts of the system. The reactor equipment uranium from the ARE fuel and thus was available operated reliably and the radioactive liquids and for partial processing of salts from either type of gases were contained safely. The fuel was reactor. completely stable. Xenon was removed rapidly The results of the ORNL studies were from the salt. When necessary, radioactive considered by a US Atomic Energy Commission equipment was repaired or replaced in reasonable task force that made a comparative evaluation of time and without overexposing maintenance fluid-fuel reactors early in 1959. One conclusion of personnel. the task force7 was that the molten-salt reactor, The second phase of MSRE operation began in although limited in potential breeding gain, had August 1968 when a small processing facility “the highest probability of achieving technical attached to the reactor was used to remove the feasibility.” original uranium by treating the fuel salt with By 1960, more complete conceptual designs of fluorine gas. A charge of 233U fuel was added to molten-salt reactors had emerged. Although em- the same carrier salt, and on October 2 the MSRE phasis was placed on the two-fluid concept because was made critical on 233U. Six days later the power of its better nuclear performance,8 the single-fluid was taken to 100 kW by Glenn T. Seaborg, reactor was also studied.9 ORNL concluded that Chairman of the US Atomic Energy Commission, either route would lead to low-power-cost reactors, bringing to power the first reactor to operate on service life, and to plan on replacement of the core 233U. at fairly frequent intervals.14 Moreover, During the years when the MSRE was being complexities in the assembly of the core seemed to built and brought into operation, most of the de- require that the entire core and reactor vessel be velopment work on molten-salt reactors was in replaced whenever a graphite element reached its support of the MSRE. However, basic chemistry radiation limit or developed a leak. Under such studies of molten fluoride salts continued through- circumstances, many years of operation of a out this period. One discovery10 during this time prototype reactor would be required to prove was that the lithium fluoride and beryllium fluoride convincingly that the two-fluid core is practicable. in a fuel salt can be separated from rare earths by At about the time the problems associated with vacuum distillation at temperatures near 1000°C. long graphite exposure became evident, a chemical This was a significant discovery, since it provided processing development occurred15 that greatly an inexpensive, on-site method for recovering these improved the prospect for a single-fluid breeder. valuable materials. As a consequence, the study To obtain good breeding performance in a single-- effort looking at future reactors focused on a two- fluid reactor, 233Pa (27.4 day half-life) must be held fluid breeder in which the fuel salt would be up outside the core until it decays to 233U.
Load more

Recommended publications
  • Arxiv:2003.07462V2 [Cond-Mat.Mtrl-Sci] 8 Jun 2020
    The Structure of Molten FLiNaK Benjamin A. Frandsena,∗, Stella D. Nickersonb, Austin D. Clarkb, Andrew Solanob, Raju Barala, Jonathan Williamsb, J¨orgNeuefeindc, Matthew Memmottb aDepartment of Physics and Astronomy, Brigham Young University, Provo, Utah 84602, USA. bDepartment of Chemical Engineering, Brigham Young University, Provo, Utah 84602, USA. cNeutron Scattering Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831, USA. Abstract The structure of the molten salt (LiF)0:465(NaF)0:115(KF)0:42 (FLiNaK), a potential coolant for molten salt nuclear reactors, has been studied by ab initio molecular dynamics simulations and neutron total scattering experiments. We find that the salt retains well-defined short-range structural correlations out to approximately 9 A˚ at typical reactor operating temperatures. The experimentally determined pair distribution function can be described with quantitative accuracy by the molecular dynamics simulations. These results indicate that the essential ionic interactions are properly captured by the simulations, providing a launching point for future studies of FLiNaK and other molten salts for nuclear reactor applications.1 Keywords: molten salt reactor, FLiNaK, total scattering, pair distribution function, molecular dynamics Molten salt reactors (MSRs) are a promising nuclear reactor concept in which fuel and/or fertile material are dissolved directly into a halide salt coolant. This has significant benefits over traditional light water reactors (LWRs) that are in operation today, including the capability of producing medical radioisotopes and electricity simultane- ously in large amounts [1] and the possibility of reactor designs that prevent proliferation of weaponizable material, eliminate the risk of meltdown events, and avoid producing long-lived transuranic nuclear waste [2, 3].
    [Show full text]
  • Molten Salts As Blanket Fluids in Controlled Fusion Reactors [Disc 6]
    r1 0 R N L-TM-4047 MOLTEN SALTS AS BLANKET FLUIDS IN CONTROLLED FUSION REACTORS W. R. Grimes Stanley Cantor .:, .:, .- t. This 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. om-TM- 4047 Contract No. W-7405-eng-26 REACTOR CHENISTRY DIVISION MOLTEN SALTS AS BLANKET FLUIDS IN CONTROLLED FUSION REACTORS W. R. Grimes and Stanley Cantor DECEMBER 1972 OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee 37830 operated by UNION CARBIDE CORPORATION for the 1J.S. ATOMIC ENERGY COMMISSION This 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, com- pleteness or usefulness of any information, apparatus, product or process disclosed, or represents that its use would not infringe privately owned rights. i iii CONTENTS Page Abstract ............................. 1 Introduction ........................... 2 Behavior of Li2BeFq in a Eypothetical CTR ............3 Effects of Strong Magnetic Fields .............5 Effects on Chemical Stability .............5 Effects on Fluid Dynamics ...............7 Production of Tritium ..................
    [Show full text]
  • Recent Research of Thorium Molten-Salt Reactor from a Sustainability Viewpoint
    Sustainability 2012, 4, 2399-2418; doi:10.3390/su4102399 OPEN ACCESS sustainability ISSN 2071-1050 www.mdpi.com/journal/sustainability Article Recent Research of Thorium Molten-Salt Reactor from a Sustainability Viewpoint Takashi Kamei Research Institute for Applied Sciences, 49, Tanaka-Oi-cho, Sakyo-ku, Kyoto 606-8202, Japan; E-Mail: [email protected]; Tel.: +81-75-701-3164; Fax: +81-75-492-0679. Received: 3 July 2012; in revised form: 20 August 2012 / Accepted: 24 August 2012 / Published: 27 September 2012 Abstract: The most important target of the concept “sustainability” is to achieve fairness between generations. Its expanding interpolation leads to achieve fairness within a generation. Thus, it is necessary to discuss the role of nuclear power from the viewpoint of this definition. The history of nuclear power has been the control of the nuclear fission reaction. Once this is obtained, then the economy of the system is required. On the other hand, it is also necessary to consider the internalization of the external diseconomy to avoid damage to human society caused by the economic activity itself, due to its limited capacity. An extreme example is waste. Thus, reducing radioactive waste resulting from nuclear power is essential. Nuclear non-proliferation must be guaranteed. Moreover, the FUKUSHIMA accident revealed that it is still not enough that human beings control nuclear reaction. Further, the most essential issue for sustaining use of one technology is human resources in manufacturing, operation, policy-making and education. Nuclear power will be able to satisfy the requirements of sustainability only when these subjects are addressed. The author will review recent activities of a thorium molten-salt reactor (MSR) as a cornerstone for a sustainable society and describe its objectives and forecasts.
    [Show full text]
  • Molten-Salt Technology and Fission Product Handling
    Molten-Salt Technology and Fission Product Handling Kirk Sorensen Flibe Energy, Inc. ORNL MSR Workshop October 4, 2018 2018-10-16 Hello, my name is Kirk Sorensen and I’d like to talk with you today about fission products and their handling in molten-salt reactors. One of the things that initially attracted me to molten-salt reactor technology was the array of options that it gave for the intelligent handling of fission products. It represented such a contrast to solid-fueled systems, which mixed fission products in with unburned nuclear fuel in a form that was difficult to separate, one from another. While my focus will be on our work on molten-salt reactor fission product handling, many of the principles are general to molten-salt reactors as a whole. Fundamental Nuclear Reactor Concept In its simplest form, a nuclear reactor generates thermal energy that is carried away by a coolant. That coolant heats the working fluid of a power conversion system, which generates electricity from part of the thermal energy and rejects the remainder to the environment. coolant working fluid fresh fuel electricity Power Nuclear Heat Conversion Reactor Exchanger System spent fuel heated water or air coolant working fluid The primary coolant chosen for a nuclear reactor determines, in large part, its size and manufacturability. The temperature of the coolant determines the efficiency of electrical generation. Fundamental Nuclear Reactor Concept In its simplest form, a nuclear reactor generates thermal energy that is carried away by a coolant. That coolant heats the working fluid of a power conversion system, which generates electricity from part of the thermal energy and rejects the remainder to the environment.
    [Show full text]
  • Molten-Salt Fast Reactors
    MOLTEN-SALT FAST REACTORS L. G. Alexander Oak Ridge National Laboratory Oak Ridge, Tennessee Thorium, plutonium, and uranium chlorides and fluorides are soluble in mixtures of the halides of Li, Be, Na, K, Mg, and other metals. Their fluoride solutions, at least, are compatible with INOR (an alloy consisting primarily of nickel). They are also compatible with graphite, a structural material as well as a moderator having many desirable properties for high-temperature application. Thus, many embodiments are possible for molten-salt reactors, ranging from simple one-fluid, one-region systems externally cooled to complex internally cooled, two-region, two- fluid systems. The capabilities of only a few of the more obvious systems have been studied so far. The nuclear and economic potentials of several thermal reactors have been evaluated heretofore, (1-3) and recently a limited program for the preliminary evaluation of fast molten-salt reactor concepts was instituted. Appropriate background studies were performed: (1) a survey was made of data available about the thermal and physical properties of molten fluoride and chloride salt mixtures, (2) the compatibilities of selected reactor materials with various reactor coolants and with molten-salt fuels were studied, (3) potential processing methods for irradiated molten fluoride and chloride reactor fuels were reviewed, and (4) data available for the nuclear properties of chlorine, nickel, and other nuclides of interest were reviewed. The compositions and physical properties of typical molten-salt
    [Show full text]
  • Status of Fast Spectrum Molten Salt Reactor Waste Management Practice December 2020
    PNNL-30739 Status of Fast Spectrum Molten Salt Reactor Waste Management Practice December 2020 Stuart T Arm David E Holcomb* Robert L Howard Brian Riley *Oak Ridge National Laboratory Prepared for the U.S. Department of Energy under Contract DE-AC05-76RL01830 Choose an item. DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor Battelle Memorial Institute, nor any of 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. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof, or Battelle Memorial Institute. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. PACIFIC NORTHWEST NATIONAL LABORATORY operated by BATTELLE for the UNITED STATES DEPARTMENT OF ENERGY under Contract DE-AC05-76RL01830 Printed in the United States of America Available to DOE and DOE contractors from the Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831-0062; ph: (865) 576-8401 fax: (865) 576-5728 email: [email protected] Available to the public from the National Technical Information Service 5301 Shawnee Rd., Alexandria, VA 22312 ph: (800) 553-NTIS (6847) email: [email protected] <https://www.ntis.gov/about> Online ordering: http://www.ntis.gov Choose an item.
    [Show full text]
  • Remote Operations and Maintenance Framework for Molten Salt Reactors
    ORNL/TM-2018/1107 Remote Operations and Maintenance Framework for Molten Salt Reactors David E. Holcomb Charles L. Britton, Jr. Venugopal K. Varma Louise G. Worrall December 2018 Approved for public release. Distribution is unlimited. DOCUMENT AVAILABILITY Reports produced after January 1, 1996, are generally available free via US Department of Energy (DOE) SciTech Connect. Website www.osti.gov Reports produced before January 1, 1996, may be purchased by members of the public from the following source: National Technical Information Service 5285 Port Royal Road Springfield, VA 22161 Telephone 703-605-6000 (1-800-553-6847) TDD 703-487-4639 Fax 703-605-6900 E-mail [email protected] Website http://classic.ntis.gov/ Reports are available to DOE employees, DOE contractors, Energy Technology Data Exchange representatives, and International Nuclear Information System representatives from the following source: Office of Scientific and Technical Information PO Box 62 Oak Ridge, TN 37831 Telephone 865-576-8401 Fax 865-576-5728 E-mail [email protected] Website http://www.osti.gov/contact.html This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of 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. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof.
    [Show full text]
  • 231Pa and 232U Production in a Fusion Breeder to Aid Nonproliferation with Thorium Fission Fuel Cycles
    Vallecitos Molten Salt Research Report No. 5 Rev.1 September 9, 2014 231Pa and 232U production in a fusion breeder to aid nonproliferation with thorium fission fuel cycles Ralph W. Moir Abstract 231Pa is made especially copiously in a fusion reactor blanket by n,2n reactions on 232Th owing to fusion’s uniquely high neutron energy being well above the reaction threshold unlike fission neutrons. The 231Pa can be extracted for use in making thorium cycles more proliferation resistant or left in the fusion reactor’s blanket to produce 232U for self- protection of the produced 233U or some of both. Typical fusion production rates of 231Pa 233 and U are of order 0.1 and 2 kg per full power year per MWfusion, respectively. Neutrons captured in 231Pa produce 232U that contributes to making 233U a 231 nonproliferant. Pa revenues per Wnucleary range from 0.08 $ to 0.5 $ depending on blanket design and market value of isotopes. By comparison, the electricity revenues is typically 0.1 $ at Q=2 and falling for Q<2 (Q=fusion power/input power). 1. Introduction This note describes a fusion breeder designed to produce 231Pa, 232U and 233U for use in molten salt reactors that are described in a companion note.1 Special emphasis is given to nonproliferation of weapons useable materials. As opposed to a fission breeder reactor, a fusion breeder reactor can produce far more fissile material for the same amount of nuclear power by an order of magnitude so that we can think of a fusion breeder being located in a site and supplying the isotopes (231Pa, 232U and 233U) to dozens of fission reactors at various sites.
    [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]
  • A New Look at Molten Salt Reactors By: Louis Qualls, Ph.D
    A New Look at Molten Salt Reactors By: Louis Qualls, Ph.D. What if you could deliver a source of clean, abundant energy that produces little waste and could be a driver for major societal change? What if that energy source is adaptable—built on a small scale for isolated communities, on a large scale for grid-scale electrical production, or perhaps for heat production to support industry? America demonstrated that technology over 50 years ago and today molten salt reactors are getting a fresh look as a modern source of clean energy. or most Americans, it is difficult to F imagine the need for energy, clean air, or clean water. However, in many parts of the world, people have little energy available to them, and in many others, the air and water are polluted. Energy is a driving force behind modern living and the need for energy is growing. Figure 1 shows the consumption of energy in the United States over the course of its history by resource [1]. The United States benefits from a diverse set of energy sources but remains predomi- nantly reliant on fossil fuels for electricity and transportation, while emission-free energy sources such as nuclear and renewables Figure 1: Energy consumed in the United States. provide smaller but significant amounts. The Source: U.S. Energy Information Administration (July 2018). transportation sector is powered mostly by released energy. The possibility of establishing chain fis- petroleum, which is the largest single source of energy sion reactions was soon postulated, and the first successful consumed. The U.S., and other developed nations, are demonstration of self-sustaining nuclear fission was per- looking beyond fossil fuels.
    [Show full text]
  • Open Problems in Reprocessing of a Molten Salt Reactor Fuel
    SK01ST117 " AIR SvixjKMwn.x VV1-K KcKIM Hink. ami S«lcly. S«|«eii*« l«!2. SUM. Minom. Ku..u Open problems in reprocessing of a molten salt reactor fuel. Vladimir Lelek, Radim Vocka Nuclear Research Institute Rez, Czech Republic [email protected], [email protected] Abstract. The study of fuel cycle in a molten salt reactor (MSR) needs deeper understanding of chemical methods used for reprocessing of spent nuclear fuel and preparation of MSR fuel, as well as of the methods employed for reprocessing of MSR fuel itself. Assuming that all the reprocessing is done on the basis of electrorefining, we formulate some open questions that should be answered before a flow sheet diagram of the reactor is designed. Most of the questions concern phenomena taking place in the vicinity of an electrode, which influence the efficiency of the reprocessing and sensibility of element separation. Answer to these questions would be an important step forward in reactor setout. 607 Page 1 of 8 I. Introduction Spent nuclear fuel contains highly radioactive isotopes and isotopes with a very long half-life (>IOOO years) which, if stored, could be potentially hazardous for environment. The main aim of the Molten Salt Demonstrational Transmuter (MSDT) is to proof the possibility of closing of the fuel cycle, that means of conversion of these radioactive isotopes (mainly plutonium and minor actinides) either to stable ones or to fission products with relatively short half-life (~30 years). The first question that must be asked is what will be the neutron balance of such a facility. Let us suppose that an asymptotic state has been reached, where fuel of stable isotopic composition is added to the reactor and only fission products with short half-life and stable isotopes are extracted.
    [Show full text]
  • Efficient and Timely Production of Valuable Radioisotopes (In a Molten-Salt Reactor)
    Efficient and Timely Production Of Valuable Radioisotopes (in a Molten-Salt Reactor) A. DeVolpi, PhD Nuclear Applications Company 1371 Corte Paguera, Oceanside, CA 92057 All rights reserved ABSTRACT Abundant and valuable Mo-99, tritium, He-3, and Pu-238 radioisotopes could be provided by a small multipurpose liquid-fueled molten-salt reactor (MSR). This enterprising means would satisfy high-priority national production and nonproliferation goals consistent with global threat reduction. It would conform to Congressional legislation requiring domestic, affordable, and proliferation-resistant radioisotope supplies for medical use, as well as meet increasing requirements having national-security applications. A single 10-100 MW(th) reactor, based on proven American technology, would fulfil growing deficiencies — at full cost-recovery, more likely a profit. Optimized for enhanced radioisotope production, the reactor would likely have to be government prioritized and located on a government reservation, in part to conform with global threat-reduction goals. It offers potential commercial product value of $billions/year, while reducing worldwide incentive and conditions for nuclear proliferation. Contemporary Requirements Ample and timely supplies of special radioisotopes — readily fulfilling objectives of the American Medical Isotopes Production Act — could be profitably produced in a single, small nuclear reactor. Nuclear Applications Company, a California/Delaware small business, is promoting consideration of a small-scale dedicated molten-salt reactor (MSR) to create abundant quantities of crucial radioisotopes at competitive cost. The proposal is based on decades of U.S. reactor development — some personally realized or witnessed at Argonne National Laboratory — for production of valuable radioisotopes, such as tritium (T-3), helium- 3 (He-3), molybdenum-99 (Mo-99), and non-fissile plutonium-238 (Pu-238).
    [Show full text]