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Wang, Z.; Sugiyama, Tomoyuki
Engineering Analysis with Boundary Elements, 135, p.266 - 283, 2022/02
Iwasawa, Yuzuru; Sugiyama, Tomoyuki; Abe, Yutaka*
Nuclear Engineering and Design, 386, p.111575_1 - 111575_17, 2022/01
Shiotsu, Hiroyuki; Ito, Hiroto*; Sugiyama, Tomoyuki; Maruyama, Yu
Annals of Nuclear Energy, 163, p.108587_1 - 108587_9, 2021/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Mihara, Takeshi; Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki
Journal of Nuclear Science and Technology, 58(8), p.872 - 885, 2021/08
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2279 - 2286, 2020/11
Probabilistic risk assessment (PRA) is one of the methods used to assess the risks associated with large and complex systems. When the risk of an external event is evaluated using conventional PRA, a particular limitation is the difficulty in considering the timing at which nuclear power plant structures, systems, and components fail. To overcome this limitation, we coupled thermal-hydraulic and external-event simulations using Risk Assessment with Plant Interactive Dynamics (RAPID). Internal flooding was chosen as the representative external event, and a pressurized water reactor plant model was used. Equations based on Bernoulli's theorem were applied to flooding propagation in the turbine building. In the analysis, uncertainties were taken into account, including the flow rate of the flood water source and the failure criteria for the mitigation systems. In terms of recovery action, isolation of the flood water source by the operator and drainage using a pump were modeled based on several assumptions. The results indicate that the isolation action became more effective when combined with drainage.
Zheng, X.; Mandelli, D.*; Alfonsi, A.*; Smith, C.*; Sugiyama, Tomoyuki
Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2176 - 2183, 2020/11
Tanaka, Yoichi; Tamaki, Hitoshi; Zheng, X.; Sugiyama, Tomoyuki
Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2195 - 2201, 2020/11
Kubo, Kotaro; Zheng, X.; Ishikawa, Jun; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Proceedings of Asian Symposium on Risk Assessment and Management 2020 (ASRAM 2020) (Internet), 11 Pages, 2020/11
Dynamic probabilistic risk assessment (PRA) enables a more realistic and detailed analysis than classical PRA. However, the trade-off for these improvements is the enormous computational cost associated with performing a large number of thermal-hydraulic (TH) analyses. In this study, based on machine learning (ML), we aim to reduce these costs by skipping the TH analysis. For the ML algorithm, we selected a support vector machine; we built it using a high-fidelity/high-cost detailed model and low-fidelity/low-cost simplified model. As a result, the computational costs could be reduced by approximately 80% without significantly decreasing the accuracy under the assumed conditions.
Matsumoto, Toshinori; Iwasawa, Yuzuru; Ajima, Kohei*; Sugiyama, Tomoyuki
Proceedings of Asian Symposium on Risk Assessment and Management 2020 (ASRAM 2020) (Internet), 10 Pages, 2020/11
The probability of ex-vessel debris coolability under the wet cavity strategy is analyzed. The first step is the uncertainty analyses by severe accident analysis code MELCOR to obtain the melt condition. Five uncertain parameters which are relating with the core degradation and transfer process were chosen. Input parameter sets were generated by LHS. The analyses were conducted and the conditions of the melt were obtained. The second step is the analyses for the behavior of melt under the water by JASMINE code. The probabilistic distribution of parameters are determined from the results of MELCOR analyses. Fifty-nine parameter sets were generated by LHS. The depth of water pool is set to be 0.5, 1.0 and 2.0 m. Debris height were compared with the criterion to judge the debris coolability. As the result, the success probability of debris cooling was obtained through the sequence of calculations. The technical difficulties of this evaluation method are also discussed.
Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.308 - 315, 2020/10
Dynamic probabilistic risk assessment (PRA) is a method for improving the realism and completeness of conventional PRA. However, enormous calculation costs are incurred by these improvements. One solution is to select an appropriate sampling method. In this paper, we applied the Monte Carlo, Latin hypercube, grid-point, and quasi-Monte Carlo sampling methods to the dynamic PRA of a simplified accident sequence and compared the results for each method. Quasi-Monte Carlo sampling was found to be the most effective method in this case.
Wang, Z.; Duan, G.*; Matsunaga, Takuya*; Sugiyama, Tomoyuki
International Journal of Heat and Mass Transfer, 157, p.119919_1 - 119919_20, 2020/08
Times Cited Count：3 Percentile：65.8(Thermodynamics)Wang, Z.; Iwasawa, Yuzuru; Sugiyama, Tomoyuki
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 12 Pages, 2020/08
Trianti, N.; Motegi, Kosuke; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 9 Pages, 2020/08
Hotta, Akitoshi*; Akiba, Miyuki*; Morita, Akinobu*; Konovalenko, A.*; Vilanueva, W.*; Bechta, S.*; Komlev, A.*; Thakre, S.*; Hoseyni, S. M.*; Sk
ld, P.*; et al.
Journal of Nuclear Science and Technology, 57(4), p.353 - 369, 2020/04
Times Cited Count：3 Percentile：41.31(Nuclear Science & Technology)Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki
Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12
Times Cited Count：3 Percentile：57.07(Nuclear Science & Technology)no abstracts in English
Motegi, Kosuke; Trianti, N.; Matsumoto, Toshinori; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.4324 - 4335, 2019/08
Tamaki, Hitoshi; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.72 - 82, 2019/08
JAEA participated in the OECD/NEA BSAF2 project with our integrated severe accident analysis code, THALES2/KICHE, in order to analyze and discuss the accident progression and source term of the accident at the Fukushima Daiichi NPS. One of important characteristics of THALES2/KICHE code is that it has the capability of predicting iodine chemistry based on reaction kinetics in the aqueous phase. JAEA performed the three week analysis for the accident at unit 1 on the basis of the boundary conditions and assumptions proposed by the BSAF2 project and our own assumptions. In addition to the failure of the drywell, it was assumed in the present analysis that continuous leakage occurred through the containment venting line due to incomplete closing of valves in the line. The releases of fission products, especially for iodine and cesium, within three weeks after the earthquake were estimated to be approximately 6% and 1% of the initial inventory, respectively.
Tamaki, Hitoshi; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.100 - 111, 2019/08
JAEA participated in the OECD/NEA BSAF2 project with our integrated severe accident analysis code, THALES2/KICHE, in order to analyze and discuss the accident progression and source term of the accident at the Fukushima Daiichi NPS. One of important characteristics of THALES2/KICHE code is that it has the capability of predicting iodine chemistry based on reaction kinetics in the aqueous phase. JAEA performed the three week analysis for the accident at unit 2 on the basis of the boundary conditions and assumptions proposed by the BSAF2 project and our own assumptions. One of focusing points in the BSAF2 project was the trend of measured data of reactor vessel from 20:00 March 14 to 02:00 March 15. An assumption was made that the lower part of the suppression chamber failed to form a water leakage path. The released iodine and cesium within three weeks after the earthquake were predicted to be approximately 3% and 0.1% of the initial inventory, respectively.
Ishikawa, Jun; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.536 - 547, 2019/08
Nakatsuka, Toru; Maeda, Toshikatsu; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.1650 - 1656, 2019/08
The OECD/NEA is launching a new project named "Analysis of Information from Reactor Buildings and Containment Vessels of Fukushima Daiichi Nuclear Power Station (ARC-F)" Project. This project will serve as the successor to the precedent NEA project, "Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) Phase II" which investigated the accident scenarios, associated fission products behavior in the damaged units and source term to the environment. The ARC-F project comprises three tasks: Task 1: Refinement of analysis for accident scenarios and associated fission product transportation and dispersion; Task 2: Compilation and management of data and information; and Task 3: Discussion for future long-term project. Japan Atomic Energy Agency is the operating agent, responsible to lead all the tasks. Duration of the project is from January 2019 to December 2021 and the final report is planned to be published in 2022.
