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Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07
Identifying accident scenarios that could lead to severe accidents and evaluating their frequency of occurrence are essential issues. This study aims to establish the methodology of the dynamic Probabilistic Risk Assessment (PRA) for sodium-cooled fast reactors that can consider the time dependency and the interdependence of each event. Specifically, the Continuous Markov chain Monte Carlo (CMMC) method is newly applied to the SPECTRA code, which analyzes the severe accident conditions of nuclear reactors, to develop an evaluation methodology for typical external hazards. Currently, a fault-tree model of air coolers of decay heat removal system is implemented as the CMMC method, and a series of preliminary analysis of the plant's transient characteristics under the scenario of volcanic ashfall has been conducted.
Jensen, C. B.*; Wachs, D. M.*; Woolstenhulme, N. E.*; Ozawa, Takayuki; Hirooka, Shun; Kato, Masato
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04
Li, C.-Y.; Uchibori, Akihiro; Takata, Takashi; Pellegrini, M.*; Erkan, N.*; Okamoto, Koji*
Dai-25-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2021/07
The capability of stable cooling and avoiding re-criticality on the debris bed are the main issues for achieving IVR (In-Vessel Retention). In the actual situation, the debris bed is composed of mixed-density debris particles. Hence, when these mixed-density debris particles were launched to re-distribute, the debris bed would possibly form a density-stratified distribution. For the proper evaluation of this scenario, the multi-physics model of CFD-DEM-Monte-Carlo based neutronics is established to investigate the coolability and re-criticality on the heterogeneous density-stratified debris bed with considering the particle relocation. The CFD-DEM model has been verified by utilizing water injection experiments on the mixed-density particle bed in the first portion of this research. In the second portion, the coupled system of the CFD-DEM-Monte-Carlo based neutronics model is applied to reactor cases. Afterward, the debris particles' movement, debris particles' and coolant's temperature, and the k-eff eigenvalue are successfully tracked. Ultimately, the relocation and stratification effects on debris bed's coolability and re-criticality had been quantitatively confirmed.
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
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.
Liu, W.; Onuki, Akira; Tamai, Hidesada; Akimoto, Hajime
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 5 Pages, 2005/10
In this research, the newest version of critical power correlation for tight-lattice rod bundles is proposed by using 7-rod and 37-rod bundle data derived in Japan Atomic Energy Research Institute (JAERI). For comparatively high mass velocity region, the correlation is written in local critical heat flux - critical quality type. For low mass velocity region, it is written in critical quality - annular flow length type. The correlation is verified by JAERI data and Bettis Atomic Power Laboratory data. It is confirmed the correlation is able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The correlation is further implemented into TRAC code to analyze flow decrease and power increase transients. It is confirmed transient BT can be predicted within the accuracy of the implemented critical power correlation.
Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo
JAERI-Data/Code 2002-015, 39 Pages, 2002/07
JAERI-Data-Code-2002-015.pdf:1.53MB
no abstracts in English
Kaminaga, Masanori; Yamamoto, Kazuyoshi
JAERI-Tech 97-016, 120 Pages, 1997/03
no abstracts in English
Kaminaga, Masanori
JAERI-Tech 97-014, 125 Pages, 1997/03
no abstracts in English
S.Navarro-Valenti*; S.H.Kim*; V.Georgevich*; R.P.Taleyarkhan*; Fuketa, Toyoshi; Soyama, Kazuhiko; Ishijima, Kiyomi; Kodaira, Tsuneo
NUREG/CP-0142 (Vol. 4), 0, p.2957 - 2976, 1996/00
no abstracts in English
Takase, Kazuyuki; Kunugi, Tomoaki; Shibata, Mitsuhiko; Seki, Yasushi
Nihon Genshiryoku Gakkai-Shi, 38(11), p.904 - 906, 1996/00
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
Yanagisawa, Kazuaki;
Journal of Nuclear Science and Technology, 32(10), p.981 - 988, 1995/10
Times Cited Count：1 Percentile：17.52(Nuclear Science & Technology)no abstracts in English
Yanagisawa, Kazuaki;
Journal of Nuclear Science and Technology, 32(9), p.889 - 897, 1995/09
Times Cited Count：2 Percentile：28.04(Nuclear Science & Technology)no abstracts in English
Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke
Journal of Nuclear Science and Technology, 32(9), p.855 - 867, 1995/09
Times Cited Count：4 Percentile：43.23(Nuclear Science & Technology)no abstracts in English
Kaminaga, Masanori; Yamamoto, Kazuyoshi; Watanabe, Shukichi; Nakano, Yoshihiro
JAERI-Tech 95-040, 79 Pages, 1995/07
no abstracts in English
Araya, Fumimasa; Murao, Yoshio
Journal of Nuclear Science and Technology, 32(4), p.339 - 350, 1995/04
Times Cited Count：3 Percentile：36.71(Nuclear Science & Technology)no abstracts in English
Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke
Transactions of the American Nuclear Society, 71, p.527 - 529, 1995/00
no abstracts in English
Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke
10th Proc. of Nuclear Thermal Hydraulics, 0, p.3 - 12, 1994/00
no abstracts in English
Yanagisawa, Kazuaki; ; Horiki, Oichiro; Soyama, Kazuhiko; Ichikawa, Hiroki; Kodaira, Tsuneo
Journal of Nuclear Science and Technology, 30(8), p.741 - 751, 1993/08
Times Cited Count：4 Percentile：44.86(Nuclear Science & Technology)no abstracts in English
