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Kubo, Kotaro; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*
Journal of Nuclear Science and Technology, 60(4), p.359 - 373, 2023/04
Times Cited Count:5 Percentile:78.52(Nuclear Science & Technology)Probabilistic risk assessment (PRA) is an essential approach to improving the safety of nuclear power plants. However, this method includes certain difficulties, such as modeling of combinations of multiple hazards. Seismic-induced flooding scenario includes several core damage sequences, i.e., core damage caused by earthquake, flooding, and combination of earthquake and flooding. The flooding fragility is time-dependent as the flooding water propagates from the water source such as a tank to compartments. Therefore, dynamic PRA should be used to perform a realistic risk analysis and quantification. This study analyzed the risk of seismic-induced flooding events by coupling seismic, flooding, and thermal-hydraulics simulations, considering the dependency between multiple hazards explicitly. For requirements of safety improvement, especially in light of the Fukushima Daiichi Nuclear Power Plant accident, sensitivity analysis was performed on the seismic capacity of systems, and the effectiveness of alternative steam generator injection by a portable pump was estimated. We demonstrate the use of this simulation-based dynamic PRA methodology to evaluate the risk induced by a combination of hazards.
Yoshida, Hiroyuki; Horiguchi, Naoki; Ono, Ayako; Furuichi, Hajime*; Katono, Kenichi*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 7 Pages, 2022/08
Gonzalz, M. A.*; Borodin, O.*; Kofu, Maiko; Shibata, Kaoru; Yamada, Takeshi*; Yamamuro, Osamu*; Xu, K.*; Price, D. L.*; Saboungi, M.-L.*
Journal of Physical Chemistry Letters (Internet), 11(17), p.7279 - 7284, 2020/09
Times Cited Count:17 Percentile:77.33(Chemistry, Physical)Ono, Ayako; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 9 Pages, 2018/10
The mechanism of Critical Heat Flux (CHF) remains to be clarified, even though it is important to evaluate the CHF for super high heat flux components such as light water reactors (LWRs). Some theoretical models to predict the CHF is proposed so far. A macrolayer formation model which is proposed in order to predict the CHF based on the macrolayer dryout model. In this model, it is assumed that the liquid is captured inside vapor mass at coalescence. In this study, the verification of the assumption of a macrolayer formation model by the numerical simulation of CMFD code, TPFIT, from the view point of hydrodynamics.
Yamamoto, Masahiro; Sato, Tomonori; Igarashi, Takahiro; Ueno, Fumiyoshi; Soma, Yasutaka
Proceedings of European Corrosion Congress 2017 (EUROCORR 2017) and 20th ICC & Process Safety Congress 2017 (USB Flash Drive), 6 Pages, 2018/09
The authors have studied the differences between outer surface and the crevice-like portion of SUS316L in high pressurized and high temperature water containing dissolved oxygen. We have already introduced that changes in the characteristics of corrosion products along the crevice directions and gap width. It is suggested that the environmental conditions are different with the features of crevice from these results. In this report, we introduce the changes in oxide films with crevice gaps and comparison with the numerical simulation data utilizing of FEM calculation.
Takayama, Yusuke; Sato, Toshinori; Onoe, Hironori; Iwatsuki, Teruki; Saegusa, Hiromitsu; Onuki, Kenji
Dai-43-Kai Gamban Rikigaku Ni Kansuru Shimpojiumu Koenshu (CD-ROM), p.313 - 318, 2015/01
In the Mizunami Underground Research Laboratory, groundwater recovery experiment is being conducted to construct the method to understand the transition of geological environment due to groundwater recovery at the -500m access and research gallery-north. As a part of this experiment, backfill test is planned using drilling pits filled with artificial materials (clay and concrete) to evaluate the influence on the surrounding rock mass due to the interaction of rock and artificial materials. In this study, numerical simulation of the backfill test has been carried out to predict the qualitative hydro-mechanical behavior.
Yoshida, Hiroyuki; Tamai, Hidesada; Onuki, Akira; Takase, Kazuyuki; Akimoto, Hajime
Nuclear Engineering and Technology, 38(2), p.119 - 128, 2006/04
The reduced-moderation water reactor core adopts a hexagonal tight-lattice arrangement. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of the core using an advanced numerical simulation technology. As a part of this technology development, we are developing a two-phase flow simulation code TPFIT with an advanced interface tracking method. The vector and parallelization of the code was conducted to fit the large-scale simulations. The numerical results applied to large-scale water-vapor two-phase flow in tight lattice rod bundles are shown and compared with experimental results. In the results, a tendency of the predicted void fraction distribution in horizontal plane agreed with the measured values including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Akimoto, Hajime
Proceedings of 2005 ASME International Mechanical Engineering Congress and Exposition (CD-ROM), 8 Pages, 2005/11
no abstracts in English
Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi*; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Nakatsuka, Toru; Akimoto, Hajime
Nihon Kikai Gakkai 2005-Nendo Nenji Taikai Koen Rombunshu, Vol.3, p.207 - 208, 2005/09
We started R&D project to develop the predictable technology for thermal-hydraulic performance of Reduced-Moderation Water Reactor (RMWR) in collaboration with power company/reactor vendor/university since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured BWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron energy. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight lattice configuration and the high void fraction. This presentation shows the advances of thermal/hydraulic feasibility study using large-scale test facility and advanced numerical simulation technology.
Onuki, Akira; Takase, Kazuyuki; Kureta, Masatoshi*; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, W.; Nakatsuka, Toru; Akimoto, Hajime
Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13) (CD-ROM), 8 Pages, 2005/05
R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R&D plan and describe some advances on experimental and analytical studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility and the analytical one aims to develop a predictable technology for geometry effects such as gap between rods etc. using advanced 3-D two-phase flow simulation methods. Steady-state and transient critical power experiments are conducted with the test facility (Gap width between rods: 1.3mm and 1.0mm) and the experimental data reveal the feasibility of RMWR.
Yonetani, Yoshiteru
Chemical Physics Letters, 406(1-3), p.49 - 53, 2005/04
Times Cited Count:52 Percentile:84.75(Chemistry, Physical)We report that a severe artifact appeared in molecular dynamics simulation of bulk water using the long cut-off length 18 AA ; Our result shows that increasing the cut-off length does not always improve the simulation result. Moreover, the use of the long cut-off length can lead to a spurious result. It is suggested that the simulation of solvated biomolecules using such a long cut-off length, which has been often performed, may contain an unexpected artifact.
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 14 Pages, 2004/10
no abstracts in English
Yoshida, Hiroyuki; Nagayoshi, Takuji*; Ose, Yasuo*; Takase, Kazuyuki; Akimoto, Hajime
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(3), p.233 - 241, 2004/09
When there are no experimental data such as the reduced-moderation water reactor (RMWR), therefore, it is very difficult to obtain highly precise predictions. The RMWR core adopts a hexagonal tight lattice arrangement with about 1 mm gap between adjacent fuel rods. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of RMWR core using advanced numerical simulation technology. As part of this technology development, we are developing advanced interface tracking method to improve conservation of volume of fluid. In this paper, we describe a newly developed interface tracking method and examples of the numerical results. In the present results, the error of volume conservation in the bubbly flow is within 0.6%.
Takase, Kazuyuki; Yoshida, Hiroyuki; Tamai, Hidesada; Ose, Yasuo*
Nihon Kikai Gakkai 2004-Nendo Nenji Taikai Koen Rombunshu, Vol.2 (No.04-1), p.251 - 252, 2004/09
no abstracts in English
Sakaba, Nariaki; Nakagawa, Shigeaki; Furusawa, Takayuki; Tachibana, Yukio
JAERI-Tech 2004-045, 67 Pages, 2004/04
Safety demonstration tests using the HTTR are now underway in order to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to research and development for the VHTR, which is one of the Generation IV reactors. In the safety demonstration tests, the coolant flow reduction test by tripping one or two out of three gas circulators is being performed between FY2002 and FY 2005 and by tripping all the three gas circulators will be conducted after FY2006. This paper describes the structural integrity assessment of the primary pressurised water cooler after one and two gas circulators run down. Also, the possibility of natural convection in the primary coolant after all the three gas circulator stopped was evaluated by the operation data of the reactor-scram test performed during the rise-to-power tests.
Yamazawa, Hiromi
Environmental Modelling & Software, 16(8), p.739 - 751, 2001/12
no abstracts in English
Yamazawa, Hiromi
KURRI-KR-61, p.100 - 105, 2000/00
no abstracts in English
; Nakamura, Hideo; *; Takeuchi, Hiroshi; S.Cevolani*; Martone, M.*; T.Hua*; D.Smith*; Katsuta, Hiroji
Proc. of 2nd Int. Topical Meeting on Nuclear Applications of Accelerator Technology (AccApp'98), p.541 - 547, 1998/00
no abstracts in English
Konsoryu, 10(4), p.360 - 363, 1996/00
no abstracts in English
*; ;
JAERI-M 90-082, 44 Pages, 1990/05
no abstracts in English