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Uchibori, Akihiro; Doda, Norihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*; Tanaka, Masaaki; Enuma, Yasuhiro; Wakai, Takashi; et al.
Nuclear Engineering and Design, 413, p.112492_1 - 112492_10, 2023/11
Times Cited Count:1 Percentile:72.91(Nuclear Science & Technology)The ARKAIDA has been developed to realize automatic optimization of plant design from safety evaluation for the advanced reactors represented by a sodium-cooled fast reactor. ARKADIA-Design offers functions to support design optimization both in normal operating conditions and design basis events. The multi-level simulation approach by the coupled analysis such as neutronics, core deformation, core thermal hydraulics was developed as one of the main technologies. On the other hand, ARKAIDA-Safety aims for safety evaluation considering severe accidents. As a key technology, the numerical methods for in- and ex-vessel coupled phenomena during severe accidents in sodium-cooled fast reactors were tested through a hypothetical severe accident event. Improvement of the ex-vessel model and development of the AI technology to find best design solution have been started.
Uchibori, Akihiro; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*; Doda, Norihiro; Tanaka, Masaaki; Enuma, Yasuhiro; Wakai, Takashi; Asayama, Tai; Ohshima, Hiroyuki
Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 10 Pages, 2022/09
The ARKAIDA has been developed to realize automatic optimization of plant design from safety evaluation for the advanced reactors represented by a sodium-cooled fast reactor. ARKADIA-Design offers functions to support design optimization both in normal operating conditions and design basis events. The multi-level simulation approach by the coupled analysis such as neutronics, core deformation, core thermal hydraulics was developed as one of the main technologies of the ARKADIA-Design. On the other hand, ARKAIDA-Safety aims for safety evaluation considering severe accidents. As a key technology, the numerical methods for in- and ex-vessel coupled phenomena during severe accidents in sodium-cooled fast reactors were tested through a hypothetical severe accident event.
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.
Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu
Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00393_1 - 16-00393_10, 2017/04
The achievement of In-Vessel Retention (IVR) of the accident consequences in an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, is effective and rational approach in enhancing the safety characteristics of sodium-cooled fast reactor. The objective of the present study is to show that the decay heat generated from the relocated fuels would be stably removed in post-accident-material-relocation/post-accident-heat-removal (PAMR/PAHR) phase, where the relocated fuels mean fuel discharged from the core into the low-pressure plenum through control-rod guide tubes, and fuel remnant in the disrupted core region (non-discharged fuel). As a result of the present assessments, it should be concluded that the stable cooling of the relocated fuels was confirmed and the prospect of IVR was obtained.
Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai
Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00395_1 - 16-00395_9, 2017/04
no abstracts in English
Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11
Sogabe, Joji; Suzuki, Toru; Wada, Yusaku; Tobita, Yoshiharu
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 3 Pages, 2016/06
no abstracts in English
Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06
Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Matsuba, Kenichi; Ito, Kei; Ohshima, Hiroyuki
Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04
Times Cited Count:27 Percentile:91.4(Nuclear Science & Technology)Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Ito, Kei
Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07
Sogabe, Joji; Takata, Takashi*; Yamaguchi, Akira*; Kikuchi, Shin; Ohshima, Hiroyuki
Journal of Nuclear Science and Technology, 49(11), p.1067 - 1077, 2012/11
Times Cited Count:1 Percentile:10.14(Nuclear Science & Technology)Sodium water reaction (SWR) is a design basis accident of a sodium-cooled fast reactor (SFR). In a steam generator (SG) of the SFR, when a heat transfer tube fails, highly pressurized water or water vapor will leak into liquid sodium resulting in a chemical reaction between sodium and water or water vapor. The mechanisms of the SWR are complicated and have not been fully elucidated. The authors have developed a numerical code and have carried out SWR experiments of a counter-flow diffusion flame (in gas phase). In this paper, the authors perform numerical simulations based on the experimental conditions to validate two chemical reaction models. In addition, sensitivity analyses are performed for various hydration numbers and water vapor flow velocities. It is founded that hydration reaction occurs somewhat in the gas-phase reaction and that influences of the water vapor flow velocity are not negligible mainly from the viewpoint of the reaction surface location.
Yamaguchi, Akira*; Takata, Takashi*; Ohshima, Hiroyuki; Sogabe, Joji*; Deguchi, Yoshihiro*; Kikuchi, Shin
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 8 Pages, 2010/05
Sodium-water reaction (SWR) is a design basis accident of a sodium fast reactor (SFR). A breach of the heat transfer tube in a steam generator (SG) results in contact of liquid sodium with water. The purpose of the present paper is to delineate the mechanism and process of the SWR by a counter-flow diffusion flame experiment and a numerical simulation.
Sogabe, Joji*; Yamaguchi, Akira*; Takata, Takashi*; Ohshima, Hiroyuki; Kikuchi, Shin; Deguchi, Yoshihiro*
no journal, ,
no abstracts in English
Sogabe, Joji*; Yamaguchi, Akira*; Takata, Takashi*; Ohshima, Hiroyuki; Kikuchi, Shin
no journal, ,
Evaluation of sodium-water reaction phenomena is one of important safety issues in sodium-cooled fast reactors. In this study, we have considered sodium dynamics near reaction surface based on the experimental results, including concentration distributions of sodium, water vapor, and aerosol measured in counter-flow diffusion flame experiment of sodium and water.
Sogabe, Joji; Koyama, Kazuya*; Tobita, Yoshiharu
no journal, ,
no abstracts in English
Sogabe, Joji; Suzuki, Toru; Tobita, Yoshiharu
no journal, ,
no abstracts in English
Sogabe, Joji; Wada, Yusaku; Suzuki, Toru; Tobita, Yoshiharu
no journal, ,
no abstracts in English
Sogabe, Joji; Wada, Yusaku*; Suzuki, Toru*; Tobita, Yoshiharu
no journal, ,
no abstracts in English
Uchibori, Akihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*
no journal, ,
The ARKADIA-Safety for safety evaluation and design optimization considering severe accident has been developed. This paper presents validation of sodium fire analysis model and integration of ex-vessel analysis model for improvement of the evaluation method.
Okano, Yasushi; Uchibori, Akihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Ishida, Shinya; Sogabe, Joji; Takata, Takashi*
no journal, ,
no abstracts in English