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Okawa, Tomio*; Mori, Shoji*; Liu, W.*; Ose, Yasuo*; Yoshida, Hiroyuki; Ono, Ayako
Nihon Genshiryoku Gakkai-Shi ATOMO
, 63(12), p.820 - 824, 2021/12
The evaluation method of the critical heat flux based on the mechanism is needed for the efficient design and development of fuel in reactors and the appropriate safety evaluation. In this paper, the current researches relating to the mechanism of the critical heat flux are reviewed, and the issue to be considered in the future are discussed.
Yoshida, Hiroyuki; Uesawa, Shinichiro; Horiguchi, Naoki; Miyahara, Naoya; Ose, Yasuo*
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11
Yoshida, Hiroyuki; Uesawa, Shinichiro; Horiguchi, Naoki; Miyahara, Naoya; Ose, Yasuo*
Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06
no abstracts in English
Ose, Yasuo*; Yoshimori, Hajime*; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki
Nihon Kikai Gakkai Dai-26-Kai Keisan Rikigaku Koenkai Rombunshu (CD-ROM), p.701_1 - 701_2, 2013/11
no abstracts in English
Misawa, Takeharu; Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Oka, Yoshiaki*
Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo (SNA & MC 2013) (CD-ROM), 2 Pages, 2013/10
Zhang, W.; Yoshida, Hiroyuki; Ose, Yasuo*; Onuki, Akira; Akimoto, Hajime; Hotta, Akitoshi*; Fujimura, Ken*
Journal of Power and Energy Systems (Internet), 2(2), p.456 - 466, 2008/00
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Kano, Takuma; Merzari, E.*; Ninokata, Hisashi*
Annual Report of the Earth Simulator Center April 2006 - March 2007, p.223 - 228, 2007/09
no abstracts in English
Zhang, W.; Yoshida, Hiroyuki; Ose, Yasuo*; Onuki, Akira; Akimoto, Hajime
Nihon Kikai Gakkai 2007-Nendo Nenji Taikai Koen Rombunshu, Vol.3, p.145 - 146, 2007/09
In relation to the thermal-hydraulic design of FLWR, this study presents a statistical evaluation of numerical simulation results obtained by a detailed two-phase flow simulation code (named TPFIT). In order to clarify mechanisms of cross flow in such tight lattice rod bundles, the TPFIT was used to simulate water-steam two-phase flow in two modeled subchannels. Attention was focused on instantaneous fluctuation characteristics of cross flow. With the calculation of correlation coefficients between the differential pressure and gas/liquid mixing coefficients, the time scales of cross flow were evaluated, and the effects of mixing section length, flow pattern and gap spacing on correlation coefficients were investigated. The difference in mechanism between gas and liquid cross flows was pointed out.
Zhang, W.; Yoshida, Hiroyuki; Ose, Yasuo*; Onuki, Akira; Akimoto, Hajime; Hotta, Akitoshi*; Fujimura, Ken*
Dai-12-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.85 - 88, 2007/06
As a candidate for next generation reactor, the innovative FLexible-fuel-cycle Water Reactor (FLWR) adopts a remarkably tight triangular lattice arrangement with about 1 mm gap spacing between adjacent fuel rods. In relation to its design, this study presents a statistical evaluation of numerical simulation results of a detailed two-phase flow simulation code (named TPFIT). In order to make clear mechanisms of cross flow in such tight lattice rod bundles, the TPFIT is used to simulate cross flow between two modeled subchannels. Attention was focused on instantaneous fluctuation characteristics of differential pressure between two subchannels and gas/liquid mixing coefficients. With the calculation of correlation coefficients between the differential pressure and gas/liquid mixing coefficients, the time scales of cross flow, e.g. lag times were evaluated, and the effects of mixing section length, flow pattern and gap spacing on correlation coefficients were extensively investigated. The difference in mechanism between gas and liquid cross flows was pointed out.
Zhang, W.; Yoshida, Hiroyuki; Ose, Yasuo*; Onuki, Akira; Akimoto, Hajime; Hotta, Akitoshi*; Fujimura, Ken*
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04
The innovative Water Reactor for FLexible fuel cycle (FLWR) adopts a tight triangular lattice arrangement with about 1 mm gap between adjacent fuel rods. In view of the importance of accurate prediction of cross flow between subchannels in the evaluation of the boiling transition (BT) in the FLWR core, this study numerically simulated steam-water two-phase cross flow between two modeled subchannels of tight-lattice rod bundle for the FLWR by using a detailed two-phase flow simulation code with an advanced interface tracking method (named TPFIT), statistically evaluated the simulation results, and clarified mechanisms of cross flow for developing a model. The effects of flow pattern, inlet and outlet of mixing section, and gap spacing on cross flow, and the local and general characters of cross flow were extensively investigated.
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Kano, Takuma; Akimoto, Hajime
Annual Report of the Earth Simulator Center April 2005 - March 2006, p.261 - 265, 2007/01
no abstracts in English
Satake, Shinichi*; Kunugi, Tomoaki*; Takase, Kazuyuki; Ose, Yasuo*
Physics of Fluids, 18(12), p.125106_1 - 125106_8, 2006/12
Times Cited Count：34 Percentile：74.54(Mechanics)no abstracts in English
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Akimoto, Hajime
Proceedings of 2006 ASME International Mechanical Engineering Congress & Exposition (IMECE 2006) (CD-ROM), 8 Pages, 2006/11
Water-vapor two-phase flow structure in a fuel bundle of an advanced light water reactor was analyzed numerically by large-scale direct simulations. A newly developed two-phase flow analysis code was used. It can precisely predict the interface behavior between the liquid and gas phase by using the interface tracking method. The present analytical geometry simulates a tight-lattice fuel bundle with 37 fuel rods and four spacers. The fuel rod outer diameter is 13 mm and gap spacing between each rod is 1.3 mm. Each spacer is installed in an arbitrary axial position in order to keeping the gap width. Water flows upward from the bottom of the fuel bundle. The inlet conditions of water are as follows: temperature 283
C, pressure 7.2 MPa, flow rate 400 kg/m
s, and the Reynolds number 40,000. In the present study three-dimensional computations were carried out under the non-heated isothermal flow condition in order to remove the effect of heat transfer by the fuel rods. The average mesh size in the present numerical study was 0.15 mm. From results of a series of the numerical simulations, the following consideration was derived: (1) The fuel rod surface is encircled with thin water film; (2) The bridge phenomenon by the water film appears in the region where the spacing between fuel rods is narrow; (3) Vapor flows downward the triangular region where the spacing between fuel rods is large; and, (4) A flow configuration of vapor shows the streak structure in the vertical direction.
Kunugi, Tomoaki*; Kawara, Zensaku*; Ose, Yasuo*; Ito, Kei; Sakai, Takaaki
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.385 - 390, 2006/11
In order to design the compact FBR with higher coolant velocity compared to the conventional reactor designs, it is necessary to clarify a criterion of a cover gas entrainment (GE) from the free-surface of the coolant pool in the reactor vessel to the heat exchanger through the hot leg. Three flow regimes are considered as the GE phenomena: a vortex dimple, a waterfall and a surface disturbance. In this study, to evaluate the GE vortex phenomena: especially for the vortex dimple, the DNS based on the MARS (Multi-interfaces Advection and Reconstruction Solver (Kunugi, 2001)) were performed for simulating the unsteady vortex-shedding experiment accompanied with the GE phenomena (Okamoto et al., 2004). The applicability of the present DNS to predict the onset of the GE vortex phenomena is discussed.
Takase, Kazuyuki; Yoshida, Hiroyuki; Tamai, Hidesada; Ose, Yasuo*; Aoki, Takayuki*; Xu, Z.*
Nihon Kikai Gakkai 2006-Nendo Nenji Taikai Koen Rombunshu, Vol.2, p.39 - 40, 2006/09
Three-dimensional large-scale numerical simulations were carried out to predict the complicated water-vapor two-phase flow characteristics in a fuel bundle of an advanced light water reactor. Conventional analysis methods with a two-fluid model need composition equations and empirical correlations based on the experimental data. Therefore, it is difficult to obtain high prediction accuracy when experimental data are nothing. Then, a new two-phase flow analysis method was proposed and the TPFIT code was developed. This paper describes the predicted liquid film, bubbly and droplet flow behavior in the simulated fuel channels with the TPFIT code, and the predicted two-phase flow behavior around a curved fuel rod with the FLUENT code which is one of the most famous commercial code. From the present results, the high prospect was acquired on the possibility of development of the thermal design procedure of the advanced nuclear reactors by large-scale simulations.
Takase, Kazuyuki; Ose, Yasuo*; Yoshida, Hiroyuki; Kano, Takuma; Aoki, Takayuki*
Nihon Kikai Gakkai Kanto Shibu Dai-12-Ki Sogo Koenkai Koen Rombunshu, p.229 - 230, 2006/03
no abstracts in English
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Akimoto, Hajime
Computational Fluid Dynamics 2004, p.649 - 654, 2006/00
no abstracts in English
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
Takase, Kazuyuki; Ose, Yasuo*; Yoshida, Hiroyuki; Akimoto, Hajime; Satake, Shinichi*
Proceedings of International Conference on Jets, Wakes and Separated Flows (ICJWSF 2005), p.137 - 144, 2005/11
no abstracts in English
Kume, Etsuo; Kitamura, Tatsuaki*; Takase, Kazuyuki; Ose, Yasuo*
Kashika Joho Gakkai-Shi, 25(Suppl.2), p.369 - 370, 2005/10
no abstracts in English
