We show with simplified numerical models, that for the kind of RBMK operated in Chernobyl: The core was unstable due to its large size and to its weak power counter-reaction coefficient, so that the power of the reactor was not easy to control even with an automatic system. Xenon oscillations could easily be activated. When there was xenon poisoning in the upper half of the core, the safety rods were designed in such a way that, at least initially, they were increasing (and not decreasing) the core reactivity. This reactivity increase has been sufficient to lead to a very high pressure increase in a significant amount of liquid water in the fuel channels thus inducing a strong propagating shock wave leading to a failure of half the pressure tubes at their junction with the drum separators. The depressurization phase (flash evaporation) following this failure has produced, after one second, a significant decrease of the water density in half the pressure tubes and then a strong reactivity accident due to the positive void effect reactivity coefficient. We evaluate the fission energy released by the accident
{"title":"A simplified analysis of the Chernobyl accident","authors":"B. Mercier, Di Yang, Ziyue Zhuang, Jiajie Liang","doi":"10.1051/EPJN/2020021","DOIUrl":"https://doi.org/10.1051/EPJN/2020021","url":null,"abstract":"We show with simplified numerical models, that for the kind of RBMK operated in Chernobyl: The core was unstable due to its large size and to its weak power counter-reaction coefficient, so that the power of the reactor was not easy to control even with an automatic system. Xenon oscillations could easily be activated. When there was xenon poisoning in the upper half of the core, the safety rods were designed in such a way that, at least initially, they were increasing (and not decreasing) the core reactivity. This reactivity increase has been sufficient to lead to a very high pressure increase in a significant amount of liquid water in the fuel channels thus inducing a strong propagating shock wave leading to a failure of half the pressure tubes at their junction with the drum separators. The depressurization phase (flash evaporation) following this failure has produced, after one second, a significant decrease of the water density in half the pressure tubes and then a strong reactivity accident due to the positive void effect reactivity coefficient. We evaluate the fission energy released by the accident","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":"1 1","pages":""},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826658","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Cross sections and fission yields can be correlated, depending on the selection of integral experimental data. To support this statement, this work presents the use of experimental isotopic compositions (both for actinides and fission products) from a sample irradiated in a reactor, to construct correlations between various cross sections and fission yields. This study is therefore complementing previous analysis demonstrating that different types of nuclear data can be correlated, based on experimental integral data.
{"title":"Fission yields and cross sections: correlated or not?","authors":"D. Rochman, E. Baugé","doi":"10.1051/EPJN/2021005","DOIUrl":"https://doi.org/10.1051/EPJN/2021005","url":null,"abstract":"Cross sections and fission yields can be correlated, depending on the selection of integral experimental data. To support this statement, this work presents the use of experimental isotopic compositions (both for actinides and fission products) from a sample irradiated in a reactor, to construct correlations between various cross sections and fission yields. This study is therefore complementing previous analysis demonstrating that different types of nuclear data can be correlated, based on experimental integral data.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":"1 1","pages":""},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826709","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
O. Pakari, T. Mager, V. Lamirand, P. Frajtag, A. Pautz
Gamma rays are an inextricable part of a nuclear reactor’s radiation field, and as such require characterization for dose rate estimations required for the radiation protection of personnel, material choices, and the design of nuclear facilities. Most commonplace radiation transport codes used for shielding calculations only included the prompt neutron induced component of the emitted gamma rays. The relative amount of gamma rays that are emitted from delayed processes – the delayed gamma fraction – amount to a significant contribution, e.g. in a typical zero power reactor at steady state is estimated to be roughly a third. Accurate predictions of gamma fields thus require an estimation of the delayed content in order to meaningfully contribute. As a consequence, recent code developments also include delayed gamma sources and require validation data. The CROCUS zero power research reactor at EPFL is part of the NEA IRPhE and has therefore been characterized for benchmark quality experiments. In order to provide the means for delayed gamma validation, a dedicated experimental campaign was conducted in the CROCUS reactor using its newly developed gamma detection capabilities based on scintillators. In this paper we present the experimental determination of the delayed gamma fraction in CROCUS using in-core neutron and gamma detectors in a benchmark reactor configuration. A consistent and flexibly applicable methodology on how to estimate the delayed gamma fraction in zero power reactors has hitherto not existed – we herein present a general experimental setup and analysis technique that can be applied to other facilities. We found that the build-up time of relevant short lived delayed gamma emitters is likely attributed to the activation of the aluminium cladding of the fuel. Using a CeBr3 scintillator in the control rod position of the CROCUS core, we determined a delayed gamma fraction of (30.6±0.6)%.
{"title":"Delayed gamma fraction determination in the zero power reactor CROCUS","authors":"O. Pakari, T. Mager, V. Lamirand, P. Frajtag, A. Pautz","doi":"10.1051/epjn/2021015","DOIUrl":"https://doi.org/10.1051/epjn/2021015","url":null,"abstract":"Gamma rays are an inextricable part of a nuclear reactor’s radiation field, and as such require characterization for dose rate estimations required for the radiation protection of personnel, material choices, and the design of nuclear facilities. Most commonplace radiation transport codes used for shielding calculations only included the prompt neutron induced component of the emitted gamma rays. The relative amount of gamma rays that are emitted from delayed processes – the delayed gamma fraction – amount to a significant contribution, e.g. in a typical zero power reactor at steady state is estimated to be roughly a third. Accurate predictions of gamma fields thus require an estimation of the delayed content in order to meaningfully contribute. As a consequence, recent code developments also include delayed gamma sources and require validation data. The CROCUS zero power research reactor at EPFL is part of the NEA IRPhE and has therefore been characterized for benchmark quality experiments. In order to provide the means for delayed gamma validation, a dedicated experimental campaign was conducted in the CROCUS reactor using its newly developed gamma detection capabilities based on scintillators. In this paper we present the experimental determination of the delayed gamma fraction in CROCUS using in-core neutron and gamma detectors in a benchmark reactor configuration. A consistent and flexibly applicable methodology on how to estimate the delayed gamma fraction in zero power reactors has hitherto not existed – we herein present a general experimental setup and analysis technique that can be applied to other facilities. We found that the build-up time of relevant short lived delayed gamma emitters is likely attributed to the activation of the aluminium cladding of the fuel. Using a CeBr3 scintillator in the control rod position of the CROCUS core, we determined a delayed gamma fraction of (30.6±0.6)%.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":"1 1","pages":""},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826873","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
E. Vandermeersch, Maxence Maillot, P. Tamagno, J. Tommasi, C. D. S. Jean
This article reviews two recently established methods to compute sensitivities of some core parameters to basic nuclear data. First, perturbation theory offers an efficient way to compute sensitivities to nuclear parameters in continuous energy transport simulations: making use of the Iterated Fission Probability method, and by coupling the Monte Carlo code TRIPOLI-4® to the nuclear evaluation code CONRAD, we were able to compute the sensitivity of core reactivity to nuclear parameters for simple ICSBEP benchmarks. Second, using a multipoint description of a nuclear system and deterministic transport calculations the sensitivity of the state eigenvector of the system to multigroup nuclear data is computed using simple and fast partial importance calculations.
{"title":"Two examples of recent advances in sensitivity calculations","authors":"E. Vandermeersch, Maxence Maillot, P. Tamagno, J. Tommasi, C. D. S. Jean","doi":"10.1051/epjn/2021012","DOIUrl":"https://doi.org/10.1051/epjn/2021012","url":null,"abstract":"This article reviews two recently established methods to compute sensitivities of some core parameters to basic nuclear data. First, perturbation theory offers an efficient way to compute sensitivities to nuclear parameters in continuous energy transport simulations: making use of the Iterated Fission Probability method, and by coupling the Monte Carlo code TRIPOLI-4® to the nuclear evaluation code CONRAD, we were able to compute the sensitivity of core reactivity to nuclear parameters for simple ICSBEP benchmarks. Second, using a multipoint description of a nuclear system and deterministic transport calculations the sensitivity of the state eigenvector of the system to multigroup nuclear data is computed using simple and fast partial importance calculations.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":"1 1","pages":""},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826849","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Amanda M. Bachmann, R. Fairhurst-Agosta, Zoë Richter, Nathan P. Ryan, Madicken Munk
Transitioning to High Assay Low Enriched Uranium-fueled reactors will alter the material requirements of the current nuclear fuel cycle, in terms of the mass of enriched uranium and Separative Work Unit capacity. This work simulates multiple fuel cycle scenarios using Cyclus to compare how the type of the advanced reactor deployed and the energy growth demand affect the material requirements of the transition to High Assay Low Enriched Uranium-fueled reactors. Fuel cycle scenarios considered include the current fleet of Light Water Reactors in the U.S. as well as a no-growth and a 1% growth transition to either the Ultra Safe Nuclear Corporation Micro Modular Reactor or the X-energy Xe-100 reactor from the current fleet of U.S. Light Water Reactors. This work explored parameters of interest including the number of advanced reactors deployed, the mass of enriched uranium sent to the reactors, and the Separative Work Unit capacity required to enrich natural uranium for the reactors. Deploying Micro Modular Reactors requires a higher average mass and Separative Work Unit capacity than deploying Xe-100 reactors, and a lower enriched uranium mass and a higher Separative Work Unity capacity than required to fuel Light Water Reactors before the transition. Fueling Xe-100 reactors requires less enriched uranium and Separative Work Unit capacity than fueling Light Water Reactors before the transition.
{"title":"Enrichment dynamics for advanced reactor HALEU support","authors":"Amanda M. Bachmann, R. Fairhurst-Agosta, Zoë Richter, Nathan P. Ryan, Madicken Munk","doi":"10.1051/epjn/2021021","DOIUrl":"https://doi.org/10.1051/epjn/2021021","url":null,"abstract":"Transitioning to High Assay Low Enriched Uranium-fueled reactors will alter the material requirements of the current nuclear fuel cycle, in terms of the mass of enriched uranium and Separative Work Unit capacity. This work simulates multiple fuel cycle scenarios using Cyclus to compare how the type of the advanced reactor deployed and the energy growth demand affect the material requirements of the transition to High Assay Low Enriched Uranium-fueled reactors. Fuel cycle scenarios considered include the current fleet of Light Water Reactors in the U.S. as well as a no-growth and a 1% growth transition to either the Ultra Safe Nuclear Corporation Micro Modular Reactor or the X-energy Xe-100 reactor from the current fleet of U.S. Light Water Reactors. This work explored parameters of interest including the number of advanced reactors deployed, the mass of enriched uranium sent to the reactors, and the Separative Work Unit capacity required to enrich natural uranium for the reactors. Deploying Micro Modular Reactors requires a higher average mass and Separative Work Unit capacity than deploying Xe-100 reactors, and a lower enriched uranium mass and a higher Separative Work Unity capacity than required to fuel Light Water Reactors before the transition. Fueling Xe-100 reactors requires less enriched uranium and Separative Work Unit capacity than fueling Light Water Reactors before the transition.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":"1 1","pages":""},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826972","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The ability to perform sensitivity analysis has been enabled for the nuclear fuel cycle simulator DYMOND through its coupling with the design and analysis toolkit Dakota. To test and demonstrate these new capabilities, a transition scenario and multi-parameter study were devised. The transition scenario represents a partial transition from the US nuclear fleet to a closed fuel cycle with small modular LWRs and fast reactors fueled by reprocessed used nuclear fuel. Four uncertain parameters in this transition were studied – start date of reprocessing, total reprocessing capacity, the nuclear energy demand growth, and the rate at which the fast reactors are deployed – with respect to their impact on four response metrics. The responses – total natural uranium consumed, maximum annual enrichment capacity required, total disposed mass, and total cost of the nuclear fuel cycle – were chosen based on measures known to be of interest in transition scenarios [2] and to be significantly impacted by the varying parameters. Analysis of this study was performed both from the direct sampling and through surrogate models developed in Dakota to calculate the global sensitivity measures Sobol’ indices. This example application of this new capability showed that the most consequential parameter to most metrics was the share of new build capacity that is fast reactors. However, for the cost metric, the scaling factor of the energy demand growth was significant and had synergistic behavior with the fast reactor new build share.
{"title":"Application of sensitivity analysis in DYMOND/Dakota to fuel cycle transition scenarios","authors":"S. Richards, B. Feng","doi":"10.1051/epjn/2021024","DOIUrl":"https://doi.org/10.1051/epjn/2021024","url":null,"abstract":"The ability to perform sensitivity analysis has been enabled for the nuclear fuel cycle simulator DYMOND through its coupling with the design and analysis toolkit Dakota. To test and demonstrate these new capabilities, a transition scenario and multi-parameter study were devised. The transition scenario represents a partial transition from the US nuclear fleet to a closed fuel cycle with small modular LWRs and fast reactors fueled by reprocessed used nuclear fuel. Four uncertain parameters in this transition were studied – start date of reprocessing, total reprocessing capacity, the nuclear energy demand growth, and the rate at which the fast reactors are deployed – with respect to their impact on four response metrics. The responses – total natural uranium consumed, maximum annual enrichment capacity required, total disposed mass, and total cost of the nuclear fuel cycle – were chosen based on measures known to be of interest in transition scenarios [2] and to be significantly impacted by the varying parameters. Analysis of this study was performed both from the direct sampling and through surrogate models developed in Dakota to calculate the global sensitivity measures Sobol’ indices. This example application of this new capability showed that the most consequential parameter to most metrics was the share of new build capacity that is fast reactors. However, for the cost metric, the scaling factor of the energy demand growth was significant and had synergistic behavior with the fast reactor new build share.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":"1 1","pages":""},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826997","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Massimo SALVATORES was not the man of a country, an organization, or a team. Certainly because of his origin, his education, and his culture, Massimo has always favored a broader and more open collaboration instead of a bureaucratic and shortsighted approach to the research, keeping the achievements only to a restricted inner circle. He was convinced that disinterested sharing makes one stronger and Massimo is one of the few nuclear reactor physicists who elevated international collaboration to its highest level. A short history of the major contributions that Massimo made to his dear discipline, Neutronics, will emphasize this peculiar aspect of his career.
{"title":"A journey into Massimo Salvatores scientific work","authors":"A. Zaetta, C. D. S. Jean, R. Jacqmin","doi":"10.1051/EPJN/2021009","DOIUrl":"https://doi.org/10.1051/EPJN/2021009","url":null,"abstract":"Massimo SALVATORES was not the man of a country, an organization, or a team. Certainly because of his origin, his education, and his culture, Massimo has always favored a broader and more open collaboration instead of a bureaucratic and shortsighted approach to the research, keeping the achievements only to a restricted inner circle. He was convinced that disinterested sharing makes one stronger and Massimo is one of the few nuclear reactor physicists who elevated international collaboration to its highest level. A short history of the major contributions that Massimo made to his dear discipline, Neutronics, will emphasize this peculiar aspect of his career.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":"1 1","pages":""},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826785","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
C. De Saint Jean, P. Tamagno, P. Archier, G. Noguere
The CONRAD code is an object-oriented software tool developed at CEA since 2005. It aims at providing nuclear reaction model calculations, data assimilation procedures based on Bayesian inference and a proper framework to treat all uncertainties involved in the nuclear data evaluation process: experimental uncertainties (statistical and systematic) as well as model parameter uncertainties. This paper will present the status of CONRAD-V1 developments concerning the theoretical and evaluation aspects. Each development is illustrated with examples and calculations were validated by comparison with existing codes (SAMMY, REFIT, ECIS, TALYS) or by comparison with experiment. At the end of this paper, a general perspective for CONRAD (concerning the evaluation and theoretical modules) and actual developments will be presented.
{"title":"CONRAD – a code for nuclear data modeling and evaluation","authors":"C. De Saint Jean, P. Tamagno, P. Archier, G. Noguere","doi":"10.1051/EPJN/2021011","DOIUrl":"https://doi.org/10.1051/EPJN/2021011","url":null,"abstract":"The CONRAD code is an object-oriented software tool developed at CEA since 2005. It aims at providing nuclear reaction model calculations, data assimilation procedures based on Bayesian inference and a proper framework to treat all uncertainties involved in the nuclear data evaluation process: experimental uncertainties (statistical and systematic) as well as model parameter uncertainties. This paper will present the status of CONRAD-V1 developments concerning the theoretical and evaluation aspects. Each development is illustrated with examples and calculations were validated by comparison with existing codes (SAMMY, REFIT, ECIS, TALYS) or by comparison with experiment. At the end of this paper, a general perspective for CONRAD (concerning the evaluation and theoretical modules) and actual developments will be presented.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":"1 1","pages":""},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826836","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
G. Rodriguez, F. Varaine, L. Costes, C. Venard, F. Serre, F. Chanteclair, Marie-Sophie Chenaud, F. Dechelette, E. Hourcade, D. Plancq, Jean Hamy, Jean-François Dirat, B. Carluec, B. Perrin, D. Verrier, S. Kubo, N. Ishikawa, Masaaki Tanaka, K. Takano, S. Ohki, T. Ozawa, H. Yamano, Yoshio Shimakawa, H. Mochida, R. Shimizu, S. Kosaka, Yumi Yamada, K. Koyama, Hisatomo Murakami, T. Iitsuka, K. Oyama, Fumiaki Kaneko, Koichi Higurashi, K. Kurita
In the frame of the France-Japan agreement on nuclear collaboration, a bilateral collaboration agreement on nuclear energy was signed on March 21st, 2017, including a topic dedicated to Sodium-cooled Fast Reactor (SFR). This agreement has set the framework to start a bilateral discussion on a joint view of an SFR concept. France (CEA and FRAMATOME) and Japan (JAEA, MHI and MFBR) have carried out these studies from 2017 to 2019. Based on the beginning of the basic design phase of ASTRID project − ASTRID − 600 MWe (ASTRID for Advanced Sodium Technological Reactor for Industrial Demonstration), the two countries performed a common work to examine ways to develop a feasible common design concept, which could be realized both in France and in Japan. The subject was then extended and extrapolated with the ASTRID − 150 MWe data (reduced power reactor and enhanced experimental capabilities) in a second phase of this study. France and Japan first focused on design requirements. Common requirements were identified, as well as differences in the safety approach and the structural design requirements, according to national standards and respective site conditions, in particular considering seismic hazards. The teams developed common Top-Level design Requirements (TLRs) to allow common specification data, then joint design. This collaborative work was carried out through the implementation of twelve France-Japan Working Groups, working jointly. This paper is providing a review of this joint synthesis on Sodium Fast Reactor design concept. It is summarizing the context and objectives, then the definition and approaches of the Top Level Requirements. This paper is then dealing with the major design features: the core design and their related safety aspects, and the nuclear island design. Thus, this paper is providing a comprehensive review of this joint work gathering French and Japan nuclear design teams during two full years.
{"title":"France–Japan synthesis concept on sodium-cooled fast reactor review of a joint collaborative work","authors":"G. Rodriguez, F. Varaine, L. Costes, C. Venard, F. Serre, F. Chanteclair, Marie-Sophie Chenaud, F. Dechelette, E. Hourcade, D. Plancq, Jean Hamy, Jean-François Dirat, B. Carluec, B. Perrin, D. Verrier, S. Kubo, N. Ishikawa, Masaaki Tanaka, K. Takano, S. Ohki, T. Ozawa, H. Yamano, Yoshio Shimakawa, H. Mochida, R. Shimizu, S. Kosaka, Yumi Yamada, K. Koyama, Hisatomo Murakami, T. Iitsuka, K. Oyama, Fumiaki Kaneko, Koichi Higurashi, K. Kurita","doi":"10.1051/epjn/2021014","DOIUrl":"https://doi.org/10.1051/epjn/2021014","url":null,"abstract":"In the frame of the France-Japan agreement on nuclear collaboration, a bilateral collaboration agreement on nuclear energy was signed on March 21st, 2017, including a topic dedicated to Sodium-cooled Fast Reactor (SFR). This agreement has set the framework to start a bilateral discussion on a joint view of an SFR concept. France (CEA and FRAMATOME) and Japan (JAEA, MHI and MFBR) have carried out these studies from 2017 to 2019. Based on the beginning of the basic design phase of ASTRID project − ASTRID − 600 MWe (ASTRID for Advanced Sodium Technological Reactor for Industrial Demonstration), the two countries performed a common work to examine ways to develop a feasible common design concept, which could be realized both in France and in Japan. The subject was then extended and extrapolated with the ASTRID − 150 MWe data (reduced power reactor and enhanced experimental capabilities) in a second phase of this study. France and Japan first focused on design requirements. Common requirements were identified, as well as differences in the safety approach and the structural design requirements, according to national standards and respective site conditions, in particular considering seismic hazards. The teams developed common Top-Level design Requirements (TLRs) to allow common specification data, then joint design. This collaborative work was carried out through the implementation of twelve France-Japan Working Groups, working jointly. This paper is providing a review of this joint synthesis on Sodium Fast Reactor design concept. It is summarizing the context and objectives, then the definition and approaches of the Top Level Requirements. This paper is then dealing with the major design features: the core design and their related safety aspects, and the nuclear island design. Thus, this paper is providing a comprehensive review of this joint work gathering French and Japan nuclear design teams during two full years.","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":"1 1","pages":""},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826858","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
B. Pérot, C. Carasco, C. Eleon, S. Bernard, A. Sardet, W. E. Kanawati, C. Deyglun, G. Perret, G. Sannie, V. Valković, D. Sudac, J. Obhodas, S. Moretto, G. Nebbia, C. Fontana, F. Pino, A. Donzella, A. Zenoni, A. Iovene, C. Tintori, M. Moszynski, M. Gierlik
Neutron inspection of sea-going cargo containers has been widely studied in the past 20 yr to non-intrusively detect terrorist threats, like explosives or Special Nuclear Materials (SNM), and illicit goods, like narcotics or smuggling materials. Fast 14 MeV neutrons are produced by a portable generator with the t(d, n)α fusion reaction, and tagged in both direction and time thanks to the alpha particle detection. This Associated Particle Technique (APT) allows focusing inspection on specific areas of interest in the containers, previously identified as containing suspicious items with X-ray radiographic scanners or radiation portal monitors. We describe the principle of APT for non-nuclear material identification, and for nuclear material detection, then we provide illustrations of the performances for 10 min inspections with significant quantities (kilograms) of explosives, illicit drugs, or SNM, in different cargo cover loads (e.g. metallic, organic, or ceramic matrices).
{"title":"Sea container inspection with tagged neutrons","authors":"B. Pérot, C. Carasco, C. Eleon, S. Bernard, A. Sardet, W. E. Kanawati, C. Deyglun, G. Perret, G. Sannie, V. Valković, D. Sudac, J. Obhodas, S. Moretto, G. Nebbia, C. Fontana, F. Pino, A. Donzella, A. Zenoni, A. Iovene, C. Tintori, M. Moszynski, M. Gierlik","doi":"10.1051/EPJN/2021004","DOIUrl":"https://doi.org/10.1051/EPJN/2021004","url":null,"abstract":"Neutron inspection of sea-going cargo containers has been widely studied in the past 20 yr to non-intrusively detect terrorist threats, like explosives or Special Nuclear Materials (SNM), and illicit goods, like narcotics or smuggling materials. Fast 14 MeV neutrons are produced by a portable generator with the t(d, n)α fusion reaction, and tagged in both direction and time thanks to the alpha particle detection. This Associated Particle Technique (APT) allows focusing inspection on specific areas of interest in the containers, previously identified as containing suspicious items with X-ray radiographic scanners or radiation portal monitors. We describe the principle of APT for non-nuclear material identification, and for nuclear material detection, then we provide illustrations of the performances for 10 min inspections with significant quantities (kilograms) of explosives, illicit drugs, or SNM, in different cargo cover loads (e.g. metallic, organic, or ceramic matrices).","PeriodicalId":44454,"journal":{"name":"EPJ Nuclear Sciences & Technologies","volume":"1 1","pages":""},"PeriodicalIF":0.5,"publicationDate":"2021-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"57826702","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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