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Ono, Ayako; Yamashita, Susumu; Suzuki, Takayuki*; Yoshida, Hiroyuki
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
JAEA is developing the methodology to predict the critical heat flux based on a mechanism in order to reduce the cost for full mock-up test. The evaluation method based on a mechanism is expected to be able to predict in the wide range of parameter under the unexpected conditions including the severe accident. In this study, the JUPITER code developed by JAEA is examined to apply for the two-phase flow simulation of LWR fuel assembly with the spacer grid. The benchmark data of single-phase flow in the bundle with the spacers by KAERI were used to validate the simulation result by JUPITER. Moreover, the single-phase flow simulation was conducted by another simulation method, STAR-CCM+, as a supplemental analysis to consider the effect of the different simulation methods. Finally, the two-phase flow simulation for the bundle with the spacer was conducted by JUPITER. The effect of the spacer with a vane on the bubble behavior is discussed.
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.
4 bundle at 2 MPaLiu, W.; Nagatake, Taku; Shibata, Mitsuhiko; Takase, Kazuyuki; Yoshida, Hiroyuki
Transactions of the American Nuclear Society, 114, p.875 - 878, 2016/06
To contribute to the clarification of the Fukushima Daiichi Accident, JAEA is working on getting instantaneous void fraction distribution data in steam water two - phase flow in rod bundle geometry under high pressure, high temperature condition, with using Wire Mesh Sensor (WMS) developed at JAEA for high pressure, high temperature condition, focusing on the low flow rate condition after the reactor scram. This paper reports the experimental results for the measured void fraction distribution in steam vapor two-phase flow in a 4 4 bundle under 1.6 MPa (202 
C), 2.1 MPa (215 C) and 2.6 MPa (226 C) conditions. The data is expected to be used in the validation of the detailed two-phase flow codes TPFIT and ACE3D developed at JAEA. The time and space averaged void fraction data is also expected being used in the validation of the drift flux models implemented in the two fluids codes, such as TRACE code.
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
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
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada
JSME International Journal, Series B, 47(2), p.323 - 331, 2004/05
no abstracts in English
Takase, Kazuyuki; Ose, Yasuo*; Akimoto, Hajime
Proceedings of the 1st International Symposium on Advanced Fluid Information (AFI-2001), p.227 - 232, 2001/10
no abstracts in English
Nakamura, Hideo
JAERI-Research 96-022, 135 Pages, 1996/05
JAERI-Research-96-022.pdf:3.86MB
no abstracts in English
; ; Fujii, Terushige*; ; Matsubayashi, Masahito; Tsuruno, Akira
Fifth World Conf. on Neutron Radiography, 0, p.118 - 125, 1996/00
no abstracts in English
Nakamura, Hideo; Kukita, Yutaka;
Journal of Nuclear Science and Technology, 32(9), p.868 - 879, 1995/09
Times Cited Count：4 Percentile：42.88(Nuclear Science & Technology)no abstracts in English
Nakamura, Hideo; Kukita, Yutaka;
Journal of Nuclear Science and Technology, 32(7), p.641 - 652, 1995/07
Times Cited Count：7 Percentile：58.73(Nuclear Science & Technology)no abstracts in English
G.R.Noghrehkar*; Kawaji, Masahiro*; A.M.C.Chan*; Nakamura, Hideo; Kukita, Yutaka
J. Fluids Eng., 117, p.1 - 9, 1995/03
no abstracts in English
26, 1993, TokaiJAERI-Conf 94-002, 154 Pages, 1994/07
no abstracts in English
Hibiki, Takashi*; Mishima, Kaichiro*; Yoneda, Kenji*; Fujine, Shigenori*; Tsuruno, Akira; Matsubayashi, Masahito
Nuclear Instruments and Methods in Physics Research A, 351, p.423 - 436, 1994/00
Times Cited Count：36 Percentile：92.62(Instruments & Instrumentation)no abstracts in English
Nakamura, Hideo; Murata, Hideo; Ito, Hideo; Anoda, Yoshinari; Kumamaru, Hiroshige; Kukita, Yutaka
Kashika Joho Gakkai-Shi, 12(47), p.47 - 56, 1992/10
no abstracts in English
Nakamura, Hideo; Murata, Hideo; Ito, Hideo; Anoda, Yoshinari; Kumamaru, Hiroshige; Kukita, Yutaka
ANS Proc. 1991 National Heat Transfer Conf., Vol. 5, p.175 - 180, 1991/00
no abstracts in English
; Murao, Yoshio
Journal of Nuclear Science and Technology, 24(10), p.785 - 797, 1987/10
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)no abstracts in English
