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Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Journal of Nuclear Science and Technology, 13 Pages, 2023/00
Times Cited Count：1 Percentile：68.31(Nuclear Science & Technology)Onoda, Yuichi; John Arul, A.*; Klimonov, I.*; Danting, S.*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 13 Pages, 2022/04
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
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
Ohgama, Kazuya; Kawashima, Katsuyuki*; Oki, Shigeo
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
In order to evaluate transient behavior of Japan sodium-cooled fast reactor (JSFR) with fuel sub-assemblies with the innerduct structure (FAIDUS) precisely, a new model for a plant dynamics code HIPRAC was developed. In this new model, inner core and outer core channels can be divided into three channels, respectively, such as interior, edge and near innerduct channel, and calculate coolant redistribution and coolant temperature in each channel. Coolant temperature distribution of interior and edge channels calculated by this model was compared with previous study by the general-purpose thermal-hydraulics code 
-FLOW. Coolant temperature behavior inside the innerduct was analyzed by a commercial thermal hydraulics code STAR-CD ver. 3.26. Based on this result, horizontally-uniformed coolant temperature in the innerduct was assumed as a heat transfer model of the innderduct. Reactivity coefficients for 750 MWe JSFR with low -decontaminated transuranic (TRU) fuel were evaluated. Transient behaviors of an unprotected loss-of-flow (ULOF) accident for JSFR with 750 MWe output calculated by previous and new models were compared. The results showed that the detailed evaluation of coolant temperature improved overestimation of the coolant temperature and coolant temperature feedback reactivity of the peripheral channels including coolant inside the innerduct and in the inter-wrapper gap.
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：91.22(Nuclear Science & Technology)Okajima, Shigeaki; *; Mukaiyama, Takehiko
JAERI-M 92-031, 81 Pages, 1992/03
no abstracts in English
Onoda, Yuichi; Matsuba, Kenichi; Tobita, Yoshiharu
no journal, ,
no abstracts in English
Kawada, Kenichi; Ishida, Shinya
no journal, ,
no abstracts in English
Tobita, Yoshiharu; Suzuki, Toru; Tagami, Hirotaka
no journal, ,
The event progression in the transition phase of ULOF (Unprotected Loss of Flow), which is the representative event in ATWS (Anticipated Transient without Scram) of fast breeding reactor. The existing experimental knowledges on the important phenomena, which dominates the event progression, were adopted in the nominal case. The reactivity lowered with accordance to the fuel discharge through control rod guide tube and the event terminated without prompt criticality. If the uncertainty was considered in the analysis, the reactivity slightly exceeded the prompt criticality, but no mechanical energy was produced.
Suzuki, Toru; Tobita, Yoshiharu; Sakai, Takaaki; Nakai, Ryodai
no journal, ,
no abstracts in English
Kubota, Ryuzaburo; Suzuki, Toru; Kawada, Kenichi; Kubo, Shigenobu; Yamano, Hidemasa; Koyama, Kazuya*; Moriwaki, Hiroyuki*; Yamada, Yumi*; Shimakawa, Yoshio*
no journal, ,
A new methodology to obtain SAS4A input data of power and reactivity profile more consistent with the core design for various core states was consolidated. Using this methodology, SAS4A analyses on the initiating phase during ULOF and UTOP transients from the full power state and the low power state were performed. This analysis study suggests that the power excursion with prompt criticality leading to large mechanical energy release can be prevented in the initiating phase of the current design for the medium-scale Gen-IV loop-type SFR.
Fujimura, Koji*; Shirakura, Shota*; Oki, Shigeo; Takeda, Toshikazu*
no journal, ,
no abstracts in English
Sogabe, Joji; Wada, Yusaku*; Suzuki, Toru*; Tobita, Yoshiharu
no journal, ,
no abstracts in English
Kawada, Kenichi; Ishida, Shinya
no journal, ,
no abstracts in English
Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Murakami, Satoshi*; Tanaka, Masaaki
no journal, ,
The multi-level simulation system with 1D-CFD coupling method which enables to evaluate various phenomena from the whole plant dynamics to the local thermal hydraulics has been developed. The numerical simulation of the ULOF test in the experimental fast reactor EBR-II in the U.S. is performed for validation study of the 1D-CFD coupling method, which combines a one-dimensional plant dynamics analysis (1D) code with a computational fluid dynamics (CFD) code. Through the numerical simulation, it was shown that the whole plant response and the multi-dimensional thermal hydraulics in the core upper plenum could be simulated. And the applicability of the 1D-CFD coupling method to plant scale analysis was confirmed in comparison with the experimental results.
Sogabe, Joji; Kamiyama, Kenji; Tobita, Yoshiharu; Okano, Yasushi
no journal, ,
During severe accidents by an anticipated transient without scram, it is important to evaluate multiphase multi-component flow behavior, when a part of the disrupted core material is discharged outside the disrupted core region through control rod guide tubes. In particular, the blockage behavior of the disrupted core material in a flow path is an important phenomenon that affects the amount of relocated fuels (the fuel discharged outside the disrupted core region and the fuel remaining in the disrupted core region). A fast reactor safety analysis code, SIMMER, is currently being developed for application to the post-accident material relocation (PAMR) phase. In the paper, aiming at actual reactor analyses for the PAMR phase of the SIMMER code, a model for the blockage in the flow path for possible phenomena in the PAMR phase. The model improves the applicability of the SIMMER code to the PAMR phase on the actual reactors.
