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Uchida, Mao*; Sakai, Takaaki*; Ezure, Toshiki; Tanaka, Masaaki
Mechanical Engineering Journal (Internet), 8(4), p.21-00161_1 - 21-00161_11, 2021/08
An evaluation method based on numerical analyses has been developed to predict occurrences gas entrainment phenomena at a free surface in a sodium-cooled fast reactor. In this study, experiments were conducted for gas entrainments due to drifting free surface vortexes observed in a circulating water tunnel geometry. Numerical analyses were also conducted in the same geometry using a computational fluid dynamics (CFD) code. Then, Strouhal numbers of vortex frequency and detailed flow velocity profiles were compared between experimental results and numerical results to clarify the evaluation accuracy of CFD calculation. As the results, the Strouhal numbers of the vortex frequency obtained from numerical analyses showed good agreement with the experimental data.
Matsushita, Kentaro; Fujisaki, Tatsuya*; Ezure, Toshiki; Tanaka, Masaaki; Uchida, Mao*; Sakai, Takaaki*
Keisan Kogaku Koenkai Rombunshu (CD-ROM), 26, 6 Pages, 2021/05
For the gas entrainment vortex at the free surface in sodium-cooled fast reactors, development of the numerical analysis method to evaluate amount of the gas entrainment from the free surface has been developing. In this paper, the automatic creation of analysis meshes which can suppress the calculation cost while maintaining the prediction accuracy of the vortex shape is investigated, and the adaptive mesh refinement (AMR) method is examined to the creation of analysis mesh applying to the unsteady vortex system. The refined mesh based on the criterion evaluated by vorticity, Q-value as second invariant of the velocity and the discriminant for the eigen equation of the velocity gradient tensor is considered, and it found that the AMR method based on Q-value can refine the analysis meshes most efficiently.
Igarashi, Kai*; Onuki, Ryoji*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08
Suzuki, Minoru*; Sakai, Takaaki*; Takata, Takashi; Doda, Norihiro
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05
With an aim to establish a quantitative risk assessment of accident managements (AMs) for various external hazards, the plant dynamics analyses with Continuous Markov Chain Monte Carlo (CMMC) method were carried out to assess repeatedly occurred multi-failures by volcano ash in volcanic eruption event. AM repetition of the filter exchange to recover the cooling function of the air coolers were considered. The results showed that this method can evaluate the effectiveness of AM measures against volcanic ash fall events with respect to time progress.
Hirakawa, Moe*; Kikuchi, Yuichiro*; Sakai, Takaaki*; Tanaka, Masaaki; Ohshima, Hiroyuki
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 8 Pages, 2018/07
Gas entrainment (GE) from cover gas is one of key issue for Sodium-cooled fast reactors to prevent unexpected effects to core reactivity. By using a computational fluid dynamics (CFD) code, analyses have been conducted to estimate the drifting vortexes on water experiments which were generated as wake vortexes behind a plate obstacle in the circulating water channel. In this paper, the results of comparison between experiments and analyses were discussed and the gas core lengths from the surface vortexes were evaluated by using the evaluation tool named StreamViewer developed by Japan Atomic Energy Agency.
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
Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Tanaka, Masaaki
Nuclear Engineering and Design, 312, p.30 - 41, 2017/02
Times Cited Count：6 Percentile：65.17(Nuclear Science & Technology)In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related Research and Development results on innovative technologies and lessons learned from Fukushima Dai-ichi Nuclear Power Plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V and V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.
Onoda, Yuichi; Kurisaka, Kenichi; Sakai, Takaaki
Journal of Nuclear Science and Technology, 53(11), p.1774 - 1786, 2016/11
Times Cited Count：1 Percentile：14.27(Nuclear Science & Technology)Takata, Takashi; Azuma, Emiko*; Nishino, Hiroyuki; Yamano, Hidemasa; Sakai, Takaaki*
Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 6 Pages, 2016/11
A new approach has been developed to assess event sequences under external hazard condition considering a plant status quantitatively and stochastically so as to take various scenarios into account automatically by applying a Continuous Markov Chain Monte Carlo (CMMC) method coupled with a plant dynamics analysis. In the paper, a strong wind is selected as the external hazard to assess the plant safety in a loop type sodium cooled fast reactor. As a result, it is demonstrated that the plant state is quite safe in case of the strong wind because multiple failures of the air coolers in the auxiliary cooling system (ACS) has a quite low probability. Furthermore, a weight factor is introduced so as to investigate the low failure probability events with a comparative small number of the sampling.
Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Okano, Yasushi; Sakai, Takaaki; Yamamoto, Takahiro*; Ishizuka, Yoshihiro*; Geshi, Nobuo*; Furukawa, Ryuta*; Nanayama, Futoshi*; et al.
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 12 Pages, 2016/10
This paper describes mainly volcanic margin assessment methodology development in addition to the project overview. The volcanic tephra could potentially clog filters of air-intakes that need the decay heat removal. The filter clogging can be calculated by atmospheric concentration and fallout duration of the volcanic tephra and also suction flow rate of each component. In this paper, the margin was defined as a grace period to a filter failure limit. Consideration is needed only when the grace period is shorter than the fallout duration. The margin by component was calculated using the filter failure limit and the suction flow rate of each component. The margin by sequence was evaluated based on an event tree and the margin by component. An accident management strategy was also suggested to extend the margin; for instance, manual trip of the forced circulation operation, sequential operation of three air coolers, and covering with pre-filter.
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
Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Okano, Yasushi; Sakai, Takaaki; Yamamoto, Takahiro*; Ishizuka, Yoshihiro*; Geshi, Nobuo*; Furukawa, Ryuta*; Nanayama, Futoshi*; et al.
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 10 Pages, 2016/06
This paper describes mainly volcanic probabilistic risk assessment (PRA) methodology development for sodium-cooled fast reactors in addition to the project overview. The volcanic ash could potentially clog air filters of air-intakes that are essential for the decay heat removal. The degree of filter clogging can be calculated by atmospheric concentration of ash and tephra fallout duration and also suction flow rate of each component. The atmospheric concentration can be calculated by deposited tephra layer thickness, tephra fallout duration and fallout speed. This study evaluated a volcanic hazard using a combination of tephra fragment size, layer thickness and duration. In this paper, each component functional failure probability was defined as a failure probability of filter replacement obtained by using a grace period to a filter failure limit. Finally, based on an event tree, a core damage frequency was estimated about 3
10
/year in total by multiplying discrete hazard probabilities by conditional decay heat removal failure probabilities. A dominant sequence was led by the loss of decay heat removal system due to the filter clogging after the loss of emergency power supply. A dominant volcanic hazard was 10
kg/m
of atmospheric concentration, 0.1 mm of tephra diameter, 50-75 cm of deposited tephra layer thickness, and 1-10 hr of tephra fallout duration.
Sakai, Takaaki; Ren, L.*; Tsige-Tamirat, H.*; Vasile, A.*; Kang, S.-H.*; Ashurko, Y.*; Fanning, T.*
Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 7 Pages, 2016/06
Sakai, Hiroshi*; Enami, Kazuhiro*; Furuya, Takaaki*; Kako, Eiji*; Kondo, Yoshinari*; Michizono, Shinichiro*; Miura, Takako*; Qiu, F.*; Sato, Masato*; Shinoe, Kenji*; et al.
Proceedings of 56th ICFA Advanced Beam Dynamics Workshop on Energy Recovery Linacs (ERL 2015) (Internet), p.63 - 66, 2015/12
no abstracts in English
Sawamura, Masaru; Umemori, Kensei*; Sakai, Hiroshi*; Shinoe, Kenji*; Furuya, Takaaki*; Enami, Kazuhiro*; Egi, Masato*
Proceedings of 12th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.579 - 582, 2015/09
no abstracts in English
Shinoe, Kenji*; Sakai, Hiroshi*; Umemori, Kensei*; Enami, Kazuhiro*; Sawamura, Masaru; Egi, Masato*; Furuya, Takaaki*
Proceedings of 12th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.548 - 552, 2015/09
no abstracts in English
Numata, Naoto*; Asakawa, Tomoyuki*; Sakai, Hiroshi*; Umemori, Kensei*; Furuya, Takaaki*; Shinoe, Kenji*; Enami, Kazuhiro*; Egi, Masato*; Sakanaka, Shogo*; Michizono, Shinichiro*; et al.
Proceedings of 12th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.566 - 570, 2015/09
no abstracts in English
Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Tanaka, Masaaki
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.8141 - 8155, 2015/08
In this paper, the authors focus on four kinds of thermal-hydraulic issues associated with the SDC, i.e. fuel assembly thermal-hydraulics, natural circulation decay heat removal, thermal striping phenomena, and core disruptive accidents, and provide a description of their evaluation method developments including verification and validation and necessary experimental studies for the Japan Sodium-cooled Fast Reactor (JSFR). These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all phenomena envisioned in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing down of knowledge/technologies.
Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Okano, Yasushi; Sakai, Takaaki; Yamamoto, Takahiro*; Ishizuka, Yoshihiro*; Geshi, Nobuo*; Furukawa, Ryuta*; Nanayama, Futoshi*; et al.
Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.454 - 465, 2015/05
This paper describes mainly strong wind PRA methodology development in addition to the project overview. In developing the strong wind PRA methodology, hazard curves were estimated by using Weibull and Gumbel distributions based on weather data recorded in Japan. The obtained hazard curves were divided into five discrete categories for event tree quantification. Next, failure probabilities for decay heat removal related components were calculated as a product of two probabilities: i.e., a probability for the missiles to enter the intake or outtake in the decay heat removal system, and fragility caused by the missile impacts. Finally, based on the event tree, the core damage frequency was estimated about 610
/year by multiplying the discrete hazard probabilities in the Gumbel distribution by the conditional decay heat removal failure probabilities. A dominant sequence was led by the assumption that the operators could not extinguish fuel tank fire caused by the missile impacts and the fire induced loss of the decay heat removal system.
Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Okano, Yasushi; Sakai, Takaaki; Yamamoto, Takahiro*; Ishizuka, Yoshihiro*; Geshi, Nobuo*; Furukawa, Ryuta*; Nanayama, Futoshi*; et al.
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 10 Pages, 2015/05
This paper describes mainly snow margin assessment methodology development in addition to the project overview. For the snow margin assessment, the index is a combination of a snowfall rate and duration. Since snow removal can be expected during the snowfall, the developed snow margin assessment methodology is such that the margin was regarded as the snowfall duration up to the decay heat removal failure which was defined as when the snow removal rate was smaller than the snowfall rate.
