Littérature scientifique sur le sujet « Partitioning, Transmutation, Nuclear Technology, Nuclear Reactors »
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Articles de revues sur le sujet "Partitioning, Transmutation, Nuclear Technology, Nuclear Reactors"
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Ruskov, Ivan, Andrei Goverdovski, Walter Furman, Yury Kopatch, Oleg Shcherbakov, Franz-Josef Hambsch, Stephan Oberstedt et Andreas Oberstedt. « Neutron induced fission of 237Np – status, challenges and opportunities ». EPJ Web of Conferences 169 (2018) : 00021. http://dx.doi.org/10.1051/epjconf/201816900021.
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Nowadays, there is an increased interest in a complete study of the neutron-induced fission of 237Np. This is due to the need of accurate and reliable nuclear data for nuclear science and technology. 237Np is generated (and accumulated) in the nuclear reactor core during reactor operation. As one of the most abundant long-lived isotopes in spent fuel (“waste”), the incineration of 237Np becomes an important issue. One scenario for burning of 237Np and other radio-toxic minor actinides suggests they are to be mixed into the fuel of future fast-neutron reactors, employing the so-called transmutation and partitioning technology. For testing present fission models, which are at the basis of new generation nuclear reactor developments, highly accurate and detailed neutron-induced nuclear reaction data is needed. However, the EXFOR nuclear database for 237Np on neutron-induced capture cross-section, σγ, and fission cross-section, σf, as well as on the characteristics of capture and fission resonance parameters (Γγ, Γf, σoΓf, fragments mass-energy yield distributions, multiplicities of neutrons vn and γ-rays vγ), has not been updated for decades.
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KITAMOTO, Asashi, et MULYANTO. « Grouping in Partitioning of HLW for Burning and/or Transmutation, with Nuclear Reactors ». Journal of Nuclear Science and Technology 32, no 6 (juin 1995) : 565–76. http://dx.doi.org/10.1080/18811248.1995.9731744.
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Shlenskii, Mikhail, et Boris Kuteev. « System Studies on the Fusion-Fission Hybrid Systems and Its Fuel Cycle ». Applied Sciences 10, no 23 (26 novembre 2020) : 8417. http://dx.doi.org/10.3390/app10238417.
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This paper is devoted to applications of fusion-fission hybrid systems (FFHS) as a powerful neutron source implementing transmutation of minor actinides (MA: Np, Am, Cm) extracted from the spent nuclear fuel (SNF) of nuclear reactors. Calculations which simulated nuclide kinetics for the metallic fuel containing MA and neutron transport were performed for particular facilities. Three FFHS with fusion power equal to 40 MW are considered in this study: demo, pilot-industrial and industrial reactors. In addition, needs for a fleet of such reactors are assessed as well as future FFHSs’ impact on Russian Nuclear Power System. A system analysis of nuclear energy development in Russia was also performed with the participation of the FFHSs, with the help of the model created at AO “Proryv”. The quantity of MA that would be produced and transmuted in this scenario is estimated. This research shows that by the means of only one hybrid facility it is possible to reduce by 2130 the mass of MA in the Russian power system by about 28%. In the case of the absence of partitioning and transmutation of MA from SNF, 287 t of MA will accumulate in the Russian power system by 2130.
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Inoue, Tadashi, Masahiro Sakata, Hajime Miyashiro, Tetsuo Matsumura, Akihiro Sasahara et Nobuya Yoshiki. « Development of Partitioning and Transmutation Technology for Long-Lived Nuclides ». Nuclear Technology 93, no 2 (février 1991) : 206–20. http://dx.doi.org/10.13182/nt91-a34506.
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Bourg, Stéphane, Andreas Geist, Jean-Marc Adnet, Chris Rhodes et Bruce C. Hanson. « Partitioning and transmutation strategy R&D for nuclear spent fuel : the SACSESS and GENIORS projects ». EPJ Nuclear Sciences & ; Technologies 6 (2020) : 35. http://dx.doi.org/10.1051/epjn/2019009.
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Processes such as PUREX allow the recovery and reuse of the uranium and the plutonium of GEN II/GEN III reactors and are being adapted for the recycling of the uranium and the plutonium of GEN IV MOX fuels. However, it does not fix the sensitive issue of the long-term management of the high active nuclear waste (HAW). Indeed, only the recovery and the transmutation of the minor actinides can reduce this burden down to a few hundreds of years. In this context, and in the continuity of the FP7 EURATOM SACSESS project, GENIORS focuses on the reprocessing of MOX fuel containing minor actinides, taking into account safety issues under normal and mal-operation. By implementing a three-step approach (reinforcement of the scientific knowledge => process development and testing => system studies, safety and integration), GENIORS will provide more science-based strategies for nuclear fuel management in the EU.
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Varlachev, Valery A., Evgeny G. Emets et Yana A. Butko. « Technology for Silicon NTD Using Pool-Type Research Reactors ». Advanced Materials Research 1084 (janvier 2015) : 333–37. http://dx.doi.org/10.4028/www.scientific.net/amr.1084.333.
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Neutron transmutation doped silicon is an important material for electronics that is based on the conversion of30Si into31P through a30Si (n,γ) →31Si reaction taking place during the neutron irradiation and followed by the beta decay of31Si into31P. The production of such silicon requires high homogeneity. The paper describes a new facility for NTD of silicon ingots of up to 5 inches in diameter and presents the experimental results that were obtained at IRT-T research nuclear reactor.
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Bergelson, B. R., A. S. Gerasimov et G. V. Tikhomirov. « Transmutation of actinides in power reactors ». Radiation Protection Dosimetry 116, no 1-4 (20 décembre 2005) : 675–78. http://dx.doi.org/10.1093/rpd/nci249.
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Maltseva, T., А. Shyshuta et S. Lukashyn. « Modern Methods of Radiochemical Reprocessing of Spent Nuclear Fuel ». Nuclear and Radiation Safety, no 1(81) (12 mars 2019) : 52–57. http://dx.doi.org/10.32918/nrs.2019.1(81).09.
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The paper is devoted to the history of development and the current state of technological and scientific advances in radiochemical reprocessing of spent nuclear fuel from water-cooled power reactors. Regarding spent nuclear fuel (SNF) of NPP power reactors, long-term energy security involves adopting a version of its radiochemical treatment, conditioning and recirculation. Recycling SNF is required for the implementation of a closed fuel cycle and the re-use of regeneration products as energy reactor fuels. The basis of modern technological schemes for the reprocessing of the spent nuclear fuel is the “Purex” process, developed since the 60s in the USA. The classic approach to the use of U and Pu nuclides contained in spent nuclear fuel is to separate them from fission products, re-enrich regenerated uranium and use plutonium for the production of mixed-oxide (MOX) fuel with depleted uranium. The modern reprocessing plants are able to deal with fuel with further increase of its main characteristics without significant changes in the initial project. In order to close the fuel cycle, it is needed to add the following technological steps: (1) removal of high-level and long-lived components and minor actinides; (2) return of actinides to the technological cycle; (3) safe disposal of unused components. Each of these areas is under investigation now. Several new promising multi-cycle hydrometallurgical processes based on the joint extraction of trivalent lanthanides and minor actinides with their subsequent separation have been developed. A number of promising materials is suggested to be potential matrices for the immobilization of high-level components of radioactive wastes. To improve the compatibility of fuel processing with the environment, non-aqueous technologies are being developed, for instance, pyro-chemical methods for the reprocessing of various types of highly active fuels based on metals, oxides, carbides, or nitrides. An important scientific and technological task under investigation is transmutation of actinides. The results of international large-scale experiments on the partitioning and transmutation of fuel with various minor actinides and long-lived fission products confirm the real possibility and expediency of closing the nuclear fuel cycle.
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François, J. L., J. J. Dorantes, C. Martín-del-Campo et J. J. E. Herrera. « LWR spent fuel transmutation with fusion-fission hybrid reactors ». Progress in Nuclear Energy 65 (mai 2013) : 50–55. http://dx.doi.org/10.1016/j.pnucene.2013.02.005.
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Takeda, Toshikazu, Toshihisa Yamamoto et Maiko Miyauchi. « Interpretation of actinide transmutation in thermal and fast reactors ». Progress in Nuclear Energy 40, no 3-4 (avril 2002) : 449–56. http://dx.doi.org/10.1016/s0149-1970(02)00037-9.
Texte intégralThèses sur le sujet "Partitioning, Transmutation, Nuclear Technology, Nuclear Reactors"
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Merk, Bruno, et Varvara Glivici-Cotruta. « Studie zur Partitionierung und Transmutation (P&T) hochradioaktiver Abfälle Stand der Grundlagen- und technologischen Forschung ». Forschungszentrum Dresden, 2014. http://nbn-resolving.de/urn:nbn:de:bsz:d120-qucosa-154560.
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Das, dem Teilprojekt zu Grunde liegende, Gesamtprojekt gliederte sich in zwei Module: In Modul A (Förderung durch das BMWi, Federführung durch KIT) und Modul B (Förderung durch das BMBF, Federführung durch acatech). Projektpartner im Modul A waren DBE TECHNOLOGY GmbH, die Gesellschaft für Anlagen- und Reaktorsicherheit mbH (GRS), das Helmholtz-Zentrum Dresden-Rossendorf (HZDR), das Karlsruher Institut für Technologie (KIT) und die Rheinisch-Westfälische Technische Hochschule (RWTH) Aachen zusammen mit dem Forschungszentrum Jülich (FZJ). Modul B wurde vom Zentrum für Interdisziplinäre Risiko- und Innovationsforschung der Universität Stuttgart (ZIRIUS) bearbeitet. Die Gesamtkoordination der beidem Module erfolgte durch die Deutsche Akademie der Technikwissenschaften (acatech). Auf Grundlage einer Analyse der wissenschaftlich-technischen Aspekte durch Modul A wurden die gesellschaftlichen Implikationen bewertet und daraus in Modul B Kommunikations- und Handlungsempfehlungen für die zukünftige Positionierung von P&T formuliert. Im, vom HZDR koordinierten, Teilprojekt „Stand der Grundlagen- und technologischen Forschung“ wird eine Übersicht über den genannten Bereich gegeben. Eingeführt wird das Thema mit einer Kurzbeschreibung möglicher Reaktorsysteme für die Transmutation. Danach wird der Entwicklungsstand der Spezialbereiche Trennchemie, Sicherheitstechnologie, Beschleunigertechnologie Flüssigmetalltechnologie, Entwicklung von Spallationstargets, Transmutationsbrennstoffen und Werkstoffkonzepten sowie Konditionierung von Abfällen, beschrieben. Dies wird ergänzt durch Spezifika von Transmutationsanlagen beginnend bei physikalischen Grundlagen und Kerndesigns, über Reaktorphysik von Transmutationsanlagen, Simulationstools und die Entwicklung von Safety Approaches. Im Anschluss wird der Stand existierender Bestrahlungseinrichtungen mit schnellem Spektrum beschrieben. Nachfolgend werden basierend auf dem derzeitigen Stand von F&E die offenen Fragen und Forschungslücken in den einzelnen Teilbereichen – Wiederaufbereitung und Konditionierung, Beschleuniger und Spallationstarget, Reaktor – zusammengestellt und sowohl eine Strategie, als auch ein Fahrplan zur Schließung der Technology Gaps entwickelt. Zusätzlich werden die Hauptbeiträge, des HZDR zur Gesamtstudie beschrieben. Dies sind insbesondere die Beschreibungen der Möglichkeiten und Grenzen von P&T, die Herausforderungen an Bestrahlungseinrichtungen zur Transmutation und deren Effektivität, sowie Sicherheitsmerkmale beschleuniger-getriebener unterkritischer Systeme inclusive grundlegender Störfallbetrachtungen und SicherheitscharakteristikThe main project, where this sub project contributed to, has been structured into two modules: module A (funded by the federal ministry of economics, managed by KIT) and module B (funded by the federal ministry of education and research, managed by acatech). Partners in module A were DBE TECHNOLOGY GmbH, the Gesellschaft für Anlagen- und Reaktorsicherheit mbH (GRS), the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the Karlsruher Institute of Technology (KIT) and the Rheinisch-Westfälische Technische Hochschule (RWTH) Aachen, in co-operation with the Forschungszentrum Jülich (FZJ). Modul B has been executed by the Zentrum für Interdisziplinäre Risiko- und Innovationsforschung der Universität Stuttgart (ZIRIUS). The overall coordination has been carried out by the Deutsche Akademie der Technikwissenschaften (acatech). The social implications have been evaluated in module B based on the analysis of the scientific and technological aspects in module A. Recommendations for communication and actions to be taken for the future positioning of P&T have been developed. In the project part, coordinated by HZDR – status of R&D – an overview on the whole topic P&T is given. The topic is opened by a short description of reactor systems possible for transmutation. In the following the R&D status of separation technologies, safety technology, accelerator technology, liquid metal technology, spallation target development, transmutation fuel and structural material development, as well as waste conditioning is described. The topic is completed by the specifics of transmutation systems, the basic physics and core designs, the reactor physics, the simulation tools and the development of Safety Approaches. Additionally, the status of existing irradiation facilities with fast neutron spectrum is described. Based on the current R&D status, the research and technology gaps in the topics: separation and conditioning, accelerator and spallation target, and reactor are characterized and a strategy as well as a roadmap for closing these gaps has been developed. In addition the major contributions of HZDR to the main project are described. The major parts are the description of the potential and the limits of P&T, the requirements and challenges for transmutation systems and the related efficiency, as well as the safety features of accelerator driven subcritical systems including the transient behavior and the safety characteristics
Livres sur le sujet "Partitioning, Transmutation, Nuclear Technology, Nuclear Reactors"
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Tej Singh, of Research Reactor Services Division, BARC. et Bhabha Atomic Research Centre, dir. Neutron transmutation doping technology of silicon and overview of trial irradiations at cirus reactor. Mumbai : Bhabha Atomic Research Centre, 2007.
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Comparison of Waste Toxicity Index and Repository Performance Assessment Approaches to Providing Guidance for R&D on Partitioning and Transmutation : Nuclear ... Nuclear Science and Technology [series]. European Communities / Union (EUR-OP/OOPEC/OPOCE), 1995.
Trouver le texte intégralChapitres de livres sur le sujet "Partitioning, Transmutation, Nuclear Technology, Nuclear Reactors"
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Tonoike, Kotaro, Hiroki Sono, Miki Umeda, Yuichi Yamane, Teruhiko Kugo et Kenya Suyama. « Options of Principles of Fuel Debris Criticality Control in Fukushima Daiichi Reactors ». Dans Nuclear Back-end and Transmutation Technology for Waste Disposal, 251–59. Tokyo : Springer Japan, 2014. http://dx.doi.org/10.1007/978-4-431-55111-9_21.
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Abderrahim, Hamid Aït. « Contribution of the European Commission to a European Strategy for HLW Management Through Partitioning & ; Transmutation ». Dans Nuclear Back-end and Transmutation Technology for Waste Disposal, 59–71. Tokyo : Springer Japan, 2014. http://dx.doi.org/10.1007/978-4-431-55111-9_7.
Texte intégralActes de conférences sur le sujet "Partitioning, Transmutation, Nuclear Technology, Nuclear Reactors"
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Chen, Guang Jun, Yu Lin Cui, Guo Guo Zhang et Hong Jun Yao. « The Development and Innovation of Spent Fuel Reprocessing in Fuel Cycle ». Dans 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29632.
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With an increased population and an increasing demand for power, nuclear power has attracted an increasing attention and mass nuclear power plant have been built in different countries in the past several decades. At present, about ten thousands ton spent fuels are discharged from nuclear power plant every year and the estimated capacity will approximately add up to 5×105 ton. Therefore, spent fuel reprocessing, by which the co-extraction and separation as well as purification of Uranium and Plutonium could be realized and ensure the recycle of uranium resources and the management of nuclear waste, is a vital step in nuclear fuel cycle including two major strategies, i.e. once-through cycle and closed fuel cycle. It is worth noting that the utilization of MOX fuel made by plutonium mixed with uranium has been successfully achieved in thermal reactor. Fortunately, the middle experiment plant of china spent fuel reprocessing has been being debugged and will be operated completely in future two years. Various reprocessing schemes have been proposed for the extraction of actinides from fission products and other elements presented in spent nuclear fuel. However, after numerous studies of alternate reprocessing methods and intensive searches for better solvents, the PUREX process remains the prime reprocessing method for spent nuclear fuels throughout the world. High burning and strong radioactive spent fuel resulting from the evolution of various reactors drive the development of the advanced PUREX technology, which emphasizes the separation of neptunium and technetium besides the separation of the Uranium and Plutonium from the majority of highly active fission products. In addition, through Partitioning and Transmutation method, some benefits such as segregating the actinides and long life fission products from the high level waste can be obtained. The GANEX process exploited by CEA, which roots in COEX process belonged to advanced PUREX process, considers the separation of the actinides and long life fission products. The study on the pyro-chemical processing such as the method of electro-deposition from molten salts has still not replaced the traditional PUREX process due to various reasons. In conclusion, the future PUREX process will focus on the modified process including predigesting the technical flowsheets and reducing reprocessing costs and using salt-less reagent in order to minimize the waste production.
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Hyland, Bronwyn, et Brian Gihm. « Scenarios for the Transmutation of Actinides in CANDU Reactors ». Dans 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-30123.
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With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU® reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past [1–4]. The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100 to 1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.
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Balas (Ghizdeanu), Nineta, et Petre Ghitescu. « Transmutation Efficiency of Plutonium and Minor Actinides in PHWR ». Dans 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48570.
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PHWRs use natural uranium as fuel and consequently the burn-up coefficient is relatively small compared to PWRs or other existing power reactors. The small burn-up coefficient results in a high volume of irradiated fuel to be disposed, with a high concentration of plutonium and minor actinides. In Romania the irradiated fuel from the existing CANDU 6 spent fuel pool is currently transferred in the Dry Intermediate Fuel Storage Facility existing at the NPP site. Partitioning and Transmutation (P&T) techniques could contribute to reduce the radioactive inventory and its associated radio-toxicity. The use for this purpose of ADS and FBR was more studied, but HWR were not. Therefore, the paper presents different theoretical possibilities to transmute/burn the Plutonium and minor actinides in two different PHWRs — CANDU and ACR, using WIMSD code. Different types of MOX alternative fuel, with variable initial Pu content are analyzed. The results present the reactivity effects along with the isotopes concentration in spent alternative fuel and determine the optimal solution for the fuel type/composition. Thus is indicated the most suitable PHWR type of reactor for possible Plutonium and minor actinides transmutation. The simulations showed that Pu content for an irradiation period of 200 days decreases from the initial value up to 11% in a CANDU reactor and 29% in an ACR. Thus ACR can reduce the plutonium inventory from MOX fuel and could be a transmutation solution. From the economic/technical point of view this analysis also provides input for a study yet to be conducted.
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Takahashi, Minoru, Masayuki Igashira, Toru Obara, Hiroshi Sekimoto, Kenji Kikuchi, Kazumi Aoto et Teruaki Kitano. « Studies on Materials for Heavy-Liquid-Metal-Cooled Reactors in Japan ». Dans 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22166.
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Recent studies on materials for the development of lead-bismuth (Pb-Bi)-cooled fast reactors (FR) and accelerator-driven sub-critical systems (ADS) in Japan are reported. The measurement of the neutron cross section of Bi to produce 210Po, the removal experiment of Po contamination and steel corrosion test in Pb-Bi flow were performed in Tokyo Institute of Technology. A target material corrosion test was performed in the project of Transmutation Experimental Facility for ADS in Japan Atomic Energy Research Institute (JAERI). Steel corrosion test was started in Mitsui Engineering & Shipbuilding Co., LTD (MES). The feasibility study for FR cycle performed in Japan Nuclear Cycle Institute (JNC) are described.
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Andrello, Concettina, Daniel Freis, Rosa Lo Frano, Dimitri Papaioannou et Fabienne Delage. « Characterization of FUTURIX-FTA Irradiated Nuclear Fuel Samples ». Dans 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-67252.
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The amount of spent fuel and high-level waste already available, and which will be produced by the future NPPs operation, calls for the evaluation of any possible technological solution that could minimize the burden of their disposal: reduction of Minor Actinide (MA) content, in addition to the radiotoxicity and radioactivity, and of the generated thermal load (decay heat). In this context, R&D efforts currently focus on the development of methodologies and technical solutions for Partitioning and Transmutation. MAs and long-lived fission products are in fact the main contributors to the long-term radiotoxicity of spent nuclear fuel, and their transmutation to short-lived fission products, in fast spectrum nuclear reactors, in transmuters or in Accelerator Driven Systems (ADS), by neutron irradiation of dedicated fuels/targets, is a promising and widely investigated option. In order to provide substantial input for the safety assessment of innovative nuclear fuels dedicated to MA transmutation, several irradiation tests are being carried out. In some options, the investigated fuels/targets are uranium free, or of low uranium content, to improve the transmutation performance and contain high concentrations of MA and plutonium compounds. Two molybdenum based CER-MET fuels, called ITU-5 & ITU-6, were prepared at JRC Karlsruhe for the irradiation experiment FUTURIX-FTA (FUel for Transmutation of transURanium elements in phenIX/Fortes Teneurs en Actinide). The experiment performed from 2007 to 2009 in the Phénix reactor, France, in cooperation with CEA. The experiment ended after 235 equivalent full power days (EFPD) at a Linear Heat Rate of circa 130 W/cm and reached burn-ups of 18 %FIHMA and 13 %FIHMA, respectively. Afterwards, the pins were transported to the Hot Cells of JRC Karlsruhe for Post Irradiation Examination. After a short summary describing the fuel preparation and irradiation conditions of the FUTURIX FTA irradiation experiment, the paper will give an overview on the current status and further planning of the Post Irradiation Examinations of ITU-5 & ITU-6 at JRC Karlsruhe. Finally, the results of the characterisations will be discussed and conclusions on the irradiation performance will be drawn. The results of this experiment will help to increase the knowledge and understanding of the irradiation behaviour of metal based transmutation targets and the qualification and validation of models developed to predict fuel safety performance.
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Choi, Sungyeol, Il Soon Hwang, Jae Hyun Cho et Chun Bo Shim. « URANUS : Korean Lead-Bismuth Cooled Small Modular Fast Reactor Activities ». Dans ASME 2011 Small Modular Reactors Symposium. ASMEDC, 2011. http://dx.doi.org/10.1115/smr2011-6650.
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Since 1994, Seoul National University (SNU) has developed an innovative future nuclear power based on LBE cooling advanced Partitioning and Transmutation (P&T) approach that leaves no high-level waste (HLW) behind with transmutation reactor named as Proliferation-resistant, Environment-friendly, Accident-tolerant, Continual, and Economical Reactor (PEACER). A small modular lead-bismuth cooled reactor has been designated as Ubiquitous, Robust, Accident-forgiving, Nonproliferating and Ultra-lasting Sustainer (URANUS-40) with a nominal electric power rating of 40 MW (100 MW thermal) that is well suited to be used as a distributed power source in either a single unit or a cluster for electricity, heat supply, and desalination. URANUS-40 is a pool type fast reactor with and an array of heterogeneous hexagonal core, fueled by proven low-enriched uranium dioxide fuels. The primary cooling system is designed to be operated by natural circulation. 3D seismic base isolation system is introduced underneath the entire reactor building allowing an earthquake of 0.5g zero period acceleration (ZPA) for the Safe Shutdown Earthquake (SSE). Also, the proliferation risk can be effectively managed by capsulized core design and a long refueling period (25yr).
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Takahashi, Minoru, Hiroshi Sekimoto, Kotaro Ishikawa, Naoki Sawada, Tadashi Suzuki, Koji Hata, Susumu Yoshida, Suizheng Qiu, Toyohiko Yano et Masamitsu Imai. « Experimental Study on Flow Technology and Steel Corrosion of Lead-Bismuth ». Dans 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22226.
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For the feasibility study of Pb-Bi-cooled fast reactors (FR) and the Pb-Bi target of accelerator-driven nuclear transmutation systems , Pb-Bi flow technologies were developed and steel corrosion behavior in a Pb-Bi flow was investigated using a Pb-Bi circulation loop. The performance of an electro-magnetic flow meter with electrically insulated electrodes plated with Rh was better than those of conventional and tubular types. Oxygen concentration was controlled by continuous injection of Ar, H2 and H2O mixture gas into the Pb-Bi flow. In order to have desired oxygen potential, the partial pressure ratio of PH2/PH2O was chosen in the range from 0.12 to 2.2 by bubbling the mixture of Ar and H2 in water columns at the room temperature. By injecting the mixture gas into the loop for sufficient time, the oxygen potentials measured by the oxygen sensor made of solid electrolyte ZrO2-Y2O3 agreed well with those in the injected gas mixture. In the first corrosion test, steels were exposed to a Pb-Bi flow at the temperature of 550 °C, the velocity of 2 m/s and the oxygen concentration of ∼5.0×10−7 wt.% for 959 hours. It was found that the weight loss was larger in the order of SS316, low Cr steel (SCM420) and high Cr steels (STBA26, SUS405, SUS430). Corrosion was suppressed by a Cr oxide layer for high Cr steels. A porous layer was formed on SS316 surface due to high solubility of Ni in Pb-Bi. In the second corrosion test, the oxygen concentration was kept at 3.6×10−7 wt.% by injecting Ar, H2 and H2O mixture gas into a Pb-Bi flow, and steels were exposed to a Pb-Bi flow at the temperature of 550 °C, the velocity of 2 m/s for 1000 hours. Serious erosion damage was observed in SCM420 at the entrance, and some erosion damages appeared in low Cr steels: SCM420, F82H, STBA26 and HCM12 downstream. Crack type damage was observed on the surface of HCM12, and pitting-type damage was observed on the surface of 2 1/4Cr-1Mo steel. Some penetration of Pb-Bi into the materials appeared in some of the erosion-damaged steels.
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Tikku, Sunil, Gilbert Raiskums, John Harber et Phil Foster. « Safety System and Control System Separation Requirements for ACR-1000™ and Operating CANDU® Reactors ». Dans 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-30320.
Texte intégralRésumé :
Digital control and safety plus the complete functional and physical separation between control and safety and also between the safety systems have been key long standing principles of CANDU® nuclear reactor technology. This paper presents a historical evolution of these principles that make CANDU reactors one of the safest technologies in the world today. The original Generation II CANDU 6 reactors started with complete separation of control from safety and the division of safety systems into two groups having strong physical separation such as opposite sides of the reactor or reactor building. Within each group a more moderate distance separation was employed. With the advent of distributed computer technology for control and display functions, key processing equipment is now moved out remote from the control rooms and distributed into channelized field equipment rooms around the reactor building as in the Four-Quadrant concept for ACR-1000™. This new approach is immune to total unavailability of any control room or equipment room due to events such as fire with minimal impact to any of the safety systems regardless of their grouping. In addition to physical separation, appropriate functional partitioning, design rules to avoid communication cross links, and diversity principles are applied to computer based I&C systems as defences against common cause faults.
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Song, Tae Yung, Choonho Cho et Chungho Cho. « The Lead-Alloy Corrosion Study at KAERI ». Dans 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89600.
Texte intégralRésumé :
KAERI (Korea Atomic Energy Research Institute) is developing an accelerator-driven transmutation system called HYPER (HYbrid Power Extraction Reactor). HYPER is the 1000MWth system designed to transmute the long-lived TRU (Transuranic Elements) and FP (Fission Product) included in the PWR spent fuel. LBE (Lead-Bismuth Eutectic) is used as the spallation target and coolant material in HYPER. KAERI has also investigated the conceptual design of a lead-cooled fast reactor. Lead (Pb) is used as the coolant material in that reactor. The most significant problem is a corrosion when using the lead-alloy liquid metal in those reactors. Therefore, it is necessary to study the corrosion characteristics and develop the technology to protect the steel structure materials against a corrosion. KAERI has been developing the facilities needed to study the corrosion of lead-alloy. KAERI fabricated a static corrosion test facility in 2003. The static corrosion tests of HT-9, 316L and T91 have been performed at 600 °C and 650 °C since 2003. The Pb-Bi loop was constructed in 2006. The Pb-Bi loop is an isothermal loop which can be operated at temperatures up to 550 °C. The Pb loop is designed to be operated with ΔT = 150 °C (Tmin = 450 °C and Tmax = 600 °C). The first stage of the Pb loop construction was finished and operations began in 2006. We will complete the second stage of the Pb loop construction after testing the first stage Pb loop.
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Cui, Deyang, Xiangzhou Cai, Jingen Chen et Chenggang Yu. « Analysis of Sustainable Thorium Fuel Utilization in Molten Salt Reactors Starting From Enriched Uranium ». Dans 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-67177.
Texte intégralRésumé :
Molten salt reactor (MSR), as one of the six systems selected by the Generation IV International Forum (GIF) for future advantaged reactors research and development (R&D), has excellent performances such as high inherent safety, desirable breeding capacity, low radioactive waste production, flexible fuel cycle and non-proliferation. Meanwhile, thorium, as an appealing alternative nuclear fuel to uranium, is more abundant than uranium in the earth’s crust. Realization of thorium fuel cycle in MSRs will greatly contribute to sustainable energy supply for global development. The objective of this paper is to analyze and evaluate thorium fuel utilization in a program in which MSRs are expected to be developed step by step. The program can be described as follows: 1 The first stage is a converter reactor fueled with low enriched uranium. With limited processing based on current chemical partitioning technology and fuel-feeding techniques in the generation-I MSR; 2 The second stage is a 233U production reactor. By using the enriched uranium, it can produce 233U which does not exist in nature; 3 The third stage is a thorium breeding reactor. It is a breeder reactor with Th/233U fuel cycle, and sustainable thorium utilization for energy production is expected to be eventually realized. By employing an in-house developed tool based on SCALE6.1, the performance of MSR fueled with low enriched uranium is firstly assessed. It is found that MSR is attractive regarding conversion ratio when compared with light water reactors. Then we illustrate the feasibility of 233U production in MSR. Enriched uranium with two enrichments are used as driver fuels to start MSR and produce 233U. The results show that 233U production can be achieved and the double time is about 79.1 years for 20% enriched uranium and 28.3 years for 60% enriched uranium. Finally, the performance of MSR based on pure Th/233U fuel cycle is evaluated. It is found that breeding fissile material is possible in MSR and the breeding ratio is desirable (1.049). Comparison of the three-stage MSRs is also conducted and the results indicate that the resource utilization efficiency is much higher in stage-III than that in the first two stages and much less minor actinides is produced in MSR operating on Th/233U fuel cycle than that in traditional light water reactor.