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Yamano, Hidemasa; Emura, Yuki; Takai, Toshihide; Kubo, Shigenobu; Quaini, A.*; Fossati, P.*; Delacroix, J.*; Journeau, C.*
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
This report mainly introduces trends in fast reactor development in Japan in addition to introducing overseas development trends for major developing countries. The paper describes major severe accident study results focusing on kinetics of interaction in core material mixtures, physical properties of core material mixtures, high temperature thermodynamic data for the uranium oxide (UO
)-iron (Fe)-boron carbide (B
C) system, experimental studies on BC-stainless steel (SS) kinetics and BC-SS eutectic material relocation (freezing), and BC-SS eutectic and kinetics models for severe accident code systems,
Onoda, Yuichi; Ishida, Shinya; Fukano, Yoshitaka; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu; Shibata, Akihiro*; Bertrand, F.*; Seiler, N.*
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
Sogabe, Joji; Ishida, Shinya; Tagami, Hirotaka; Okano, Yasushi; Kamiyama, Kenji; Onoda, Yuichi; Matsuba, Kenichi; Yamano, Hidemasa; Kubo, Shigenobu; Kubota, Ryuzaburo*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
In the frame of France-Japan collaboration, the calculational methodologies were defined and assessed, and the phenomenology and the severe accident consequences were investigated in a pool-type sodium-cooled fast reactor.
Kamide, Hideki; Asayama, Tai; Wakai, Takashi; Ezure, Toshiki; Uchibori, Akihiro; Kubo, Shigenobu; Takeuchi, Masayuki
Nuclear Engineering and Design, 421, p.113062_1 - 113062_10, 2024/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)A sodium cooled fast reactor (SFR) is one of the most relevant and decarbonized energy supply system with higher sustainability on natural resources, footprint, and waste management. It was planned in a strategic roadmap of fast reactor decided by Inter-Ministerial Council for Nuclear Power Japan in 2022 to start a conceptual design of a demonstration reactor from 2024 with a background of accumulated knowledge and experiences of SFR development. For example, a design and lifecycle simulation/evaluation system named ARKADIA has been developed to accelerate such design works. It will enable to take into account plant lifecycle, e.g., operation and maintenance, to the plant design and optimize it based on simulations and knowledgebase. This paper shows research progresses of ARKADIA, safety design and evaluations, codes and standards, fuel cycle, and SFR development projects in Japan.
Tagami, Hirotaka; Ishida, Shinya; Okano, Yasushi; Yamano, Hidemasa; Kubo, Shigenobu; Payot, F.*; Saas, L.*; Trotignon, L.*; Gubernatis, P.*; Dufour, E.*; et al.
Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 12 Pages, 2024/05
JAEA has been developing the SIMMER-V code in collaboration with CEA to perform severe accident (SA) simulations of future sodium-cooled fast reactors (SFRs) including a unique core design with large-scale heterogeneous cores. An SA sequence in SFRs has been analyzed by: the SAS4A code for the Initiation Phase (IP), in which fuel pin disruption and vertical fuel dispersion occur in individual fuel subassemblies; and the two-dimensional SIMMER-III or three-dimensional SIMMER-IV code for the Transition Phase (TP), in which core disruption extends to the whole core. The joint development of SIMMER-V is of limited scope but aims at significantly expanding the code applicability by providing flexible interfaces to couple a SIMMER-V calculation with other computational domains or other codes, and by adding new advanced physical models such as a detailed fuel pin model and a model of flexible treatment of fuel isotopic composition. The former tasks are conducted by CEA the latter tasks by JAEA. In parallel to the code development, verification and validation of the new models and methods have been performed. This paper describes the objectives and overall framework of SIMMER-V code development program, representative new elements, and recent development progress.
Ishida, Shinya; Tagami, Hirotaka; Okano, Yasushi; Yamano, Hidemasa; Kubo, Shigenobu; Tobita, Yoshiharu
Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 10 Pages, 2024/05
Kamide, Hideki; Kawasaki, Nobuchika; Hayafune, Hiroki; Kubo, Shigenobu; Chikazawa, Yoshitaka; Maeda, Seiichiro; Sagayama, Yutaka; Nishihara, Tetsuo; Sumita, Junya; Shibata, Taiju; et al.
Jisedai Genshiro Ga Hiraku Atarashii Shijo; NSA/Commentaries, No.28, p.14 - 36, 2023/10
Developments of next generation nuclear reactors, e.g., Fast Reactor, and High Temperature Gas cooled Reactor, are in progress. They can contribute to markets of electricity and industrial heat utilization in the world including Japan. Here, current status of reactor developments in Japan and also situation in the world are summarized, especially for activities of Generation IV International Forum (GIF), developments of Fast Reactor and High Temperature Gas cooled Reactor in Japan, and SMR movements in the world.
Ishida, Shinya; Tagami, Hirotaka; Tobita, Yoshiharu; Okano, Yasushi; Yamano, Hidemasa; Kubo, Shigenobu
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
no abstracts in English
Kubo, Shigenobu; Chikazawa, Yoshitaka; Ohshima, Hiroyuki; Kamide, Hideki
Handbook of Generation IV Nuclear Reactors, Second Edition, p.173 - 194, 2023/03
Handbook of Generation IV Nuclear Reactors, Second Edition is a fully revised and updated comprehensive resource on the latest research and advances in generation IV nuclear reactor concepts. Editor Igor Pioro and his team of expert contributors have updated every chapter to reflect advances in the field since the first edition published in 2016. JAEA contributes to Chapter 5; Sodium-cooled Fast Reactors (SFRs) and Chapter 12; Generation-IV Sodium-cooled Fast Reactor (SFR) concepts in Japan. Major characteristics and current technology developments including safety enhancement were described in Chapter 5. Chapter 12 shows design activities of SFR. Innovative technology developments, and update of the Japan sodium-cooled fast reactor design with lessons learned from the TEPCO Fukushima Daiichi NPP accident.
Takano, Kazuya; Oki, Shigeo; Ozawa, Takayuki; Yamano, Hidemasa; Kubo, Shigenobu; Ogura, Masashi*; Yamada, Yumi*; Koyama, Kazuya*; Kurita, Koichi*; Costes, L.*; et al.
EPJ Nuclear Sciences & Technologies (Internet), 8, p.35_1 - 35_9, 2022/12
The France and Japan teams have carried out collaborative works to have common technical views regarding a sodium-cooled fast reactor concept. Japan has studied the feasibility of an enhanced high burnup low-void effect (CFV) core and fuel using oxide dispersion-strengthened steel cladding in ASTRID 600. Regarding passive shutdown capabilities, Japan team has performed a preliminary numerical analysis for ASTRID 600 using a complementary safety device, called a self-actuated shutdown system (SASS), one of the safety approaches of Japan. The mitigation measures of ASTRID 600 against a severe accident, such as a core catcher, molten corium discharge assembly, and the sodium void reactivity features of the CFV core, are promising to achieve in-vessel retention for both countries. The common design concept based on ASTRID 600 is feasible to demonstrate the SFR core and safety technologies for both countries.
Nakamura, Hironori*; Hayakawa, Satoshi*; Shibata, Akihiro*; Sasa, Kyohei*; Yamano, Hidemasa; Kubo, Shigenobu
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 7 Pages, 2022/10
In order to evaluate long-term coolablity of the debris-bed with decay heat, a three-dimensional calculation method coupled with the debris bed module was developed in this study. The coupled code calculation results show that natural circulation of the coolant between the hot pool and the cold pool is established through the four intermediate heat exchangers after the activation of the dipped direct heat exchangers. The cold pool with the debris-bed is continually cooled not only by the natural circulation flow, but also by heat transfer to the hot pool through the plenum separation plate between the hot pool and the cold pool. The effect of the three-dimensional flow field around the core catcher on the temperature in the debris-bed is about 20K under the current calculation condition.
Futagami, Satoshi; Kubo, Shigenobu; Sofu, T.*; Ammirabile, L.*; Gauthe, P.*
Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10
Yamano, Hidemasa; Kubo, Shigenobu; Tokizaki, Minako*; Nakamura, Hironori*
Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 12 Pages, 2022/10
Specific design features of advanced sodium-cooled fast reactors (SFRs) designed in Japan are a passive reactor shutdown system, a passive decay heat removal system (DHRS), and an in-vessel retention (IVR) concept against an anticipated transients without scram (ATWS) in design extension condition (DECs). The present paper describes numerical analysis methodologies for event sequences studied in Japan and some numerical analyses of DECs to show the effectiveness of the passive shutdown system against a typical ATWS and severe accident mitigation measures for the IVR of molten core. For the passive shutdown capability, the numerical analysis has demonstrated the effectiveness of a self-actuated shutdown system against a severe ATWS event, for which the temperature response time was separately evaluated by a computational fluid dynamics (CFD) code. A recently developed debris-bed cooling analysis methodology coupled with a CFD code and a debris-bed module has successfully simulated a three-dimensional coolant flow field near the debris bed with the passive DHRS capability in order to demonstrate the debris-bed coolability on a core catcher.
Yamano, Hidemasa; Kubo, Shigenobu; Kan, Taro*; Shibata, Akihiro*; Hourcade, E.*; Dirat, J. F.*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 10 Pages, 2022/08
In this paper, the approach to event tree development and the scope of the event tree analysis were described with key points on core catcher loading. For the analytical conditions, two core catcher loading conditions were given as bounding and conservative cases. For important heading of the event tree, key important phenomena were included: strong back design, fuel-coolant interaction and quench in the sodium plenum design, jet attack, criticality and coolability on the core catcher. In this paper, preliminary trial quantification was attempted using a probability ranking table which is based on engineering judgement. This event tree analysis has identified the dominant sequence, and clarified the effect of the core catcher loading and effectiveness of design measures. This study suggests that the criticality measure is very important for the core catcher study.
Yamano, Hidemasa; Kubo, Shigenobu; Sasa, Kyohei*; Shibata, Akihiro*; Hourcade, E.*; Dirat, J. F.*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 9 Pages, 2022/08
This paper describes coolability evaluations of a debris bed with a variety of decay heat removal system (DHRS) operating conditions with a whole vessel model assuming fuel accumulation on the core catcher in a short term. The evaluation tool is a one-dimensional plant dynamics code, Super-COPD, with a debris bed module. The coolability evaluations have indicated that the current core catcher design secures sufficient natural circulation flows around the core catcher to ensure the debris bed cooling when at least one circuit of DHRS was activated. Sensitivity analyses under a pessimistic condition have shown that the debris bed is coolable with at least one circuit of improved DHRS even if most of fuel accumulates on the core catcher in a short term.
Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
Bachrata, A.*; Gentet, D.*; Bertrand, F.*; Marie, N.*; Kubota, Ryuzaburo*; Sogabe, Joji; Sasaki, Keisuke; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04
In the frame of France-Japan collaboration, one of the objectives is to define and assess the calculation methodologies, and to investigate the phenomenology and the consequences of severe accident scenarios in sodium fast reactors (SFRs). A methodology whose purpose is to assess the loadings of the structures induced by a Fuel Coolant Interaction (FCI) taking place in the sodium plenum of SFR has been defined in the frame of the collaboration between France and Japan during 2014-2019. The work progress will be spread over the period 2020-2024 and the main objectives and milestones will be introduced in the paper. The objective of studies is to comprehensively address the margin between the limit of integrity of the main vessel structures and the loadings resulting from severe accidents. For this purpose, the SIMMER mechanistic calculation code simulates core disruptive accident sequences in SFRs. A fluid structure dynamics tool evaluates this interaction i.e. EUROPLEXUS is used in CEA studies and AUTODYN tool is used in JAEA studies. In the paper, a benchmark study is described in order to illustrate the evaluation of vapour expansion phase in the hot plenum. To do that, joint input data are used on the basis of an ASTRID 1500 MWth core degraded state after the power excursion which leads to vapour expansion. The most penalizing case was evidenced in this study by suppressing the action of transfer tube in-core mitigation devices in SIMMER input deck and thus privileging the upward molten core ejection. Even if the most penalizing case was evidenced in this paper, no significant RV deformation was observed in both EUROPLEXUS and AUTODYN calculation results. The assumed mechanical energy was small for the core expansion phase.
Kubo, Shigenobu; Payot, F.*; Yamano, Hidemasa; Bertrand, F.*; Bachrata, A.*; Saas, L.*; Journeau, C.*; Gosse, S.*; Quaini, A.*; Shibata, Akihiro*; et al.
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04
Kato, Atsushi; Kubo, Shigenobu; Chikazawa, Yoshitaka; Miyagawa, Takayuki*; Uchita, Masato*; Suzuno, Tetsuji*; Endo, Junji*; Kubo, Koji*; Murakami, Hisatomo*; Uzawa, Masayuki*; et al.
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 11 Pages, 2022/04
The authors are carrying out conceptual design studies for a pool-type sodium-cooled fast reactor. There are main challenges such as measures against severe earthquake in Japan, thermal hydraulic in a reactor vessel (RV), a decay heat removal system design. When the JP-pool SFR of 650 MWe is installed in Japan, it shall be designed against the severe seismic conditions. Additionally, a newly three-dimensional seismic isolation system is under development.
Rodriguez, G.*; Varaine, F.*; Costes, L.*; Venard, C.*; Serre, F.*; Chanteclair, F.*; Chenaud, M.-S.*; Dechelette, F.*; Hourcade, E.*; Plancq, D.*; et al.
EPJ Nuclear Sciences & Technologies (Internet), 7, p.15_1 - 15_8, 2021/00
France (CEA and FRAMATOME) and Japan (JAEA, MHI and MFBR) have carried out studies to establish a common technical view regarding sodium-cooled fast reactor concept. Japan and France performed a common work to examine ways to develop a feasible common design concept, which could be built both in France and/or in Japan. This paper is providing a review of this joint synthesis on Sodium Fast Reactor design concept.
