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Han, X.*; Shen, X.*; Yamamoto, Toshihiro*; Nakajima, Ken*; Sun, Haomin; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 178, p.121637_1 - 121637_24, 2021/10
Times Cited Count：7 Percentile：58.99(Thermodynamics)
6 rod bundleHan, X.*; Shen, X.*; Yamamoto, Toshihiro*; Nakajima, Ken*; Sun, Haomin; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 144, p.118696_1 - 118696_19, 2019/12
Times Cited Count：14 Percentile：62.14(Thermodynamics)Xiao, Y.*; Shen, X.*; Miwa, Shuichiro*; Sun, Haomin; Hibiki, Takashi*
Konsoryu Shimpojiumu 2018 Koen Rombunshu (Internet), 2 Pages, 2018/08
In order to develop constitutive equations of two-fluid model in rod bundle flow channels, experiments of adiabatic air-water upward two-phase flow in 66 rod bundle flow channel were performed. Local flow parameters such as void fraction, interfacial area concentration (IAC) and so on were measured by a double-sensor optical probe. The area-averaged void fraction and IAC data were compared with the predictions from a drift-flux model and an IAC correlation.
Yonomoto, Taisuke; Shibamoto, Yasuteru; Satou, Akira; Okagaki, Yuria
Journal of Nuclear Science and Technology, 53(9), p.1342 - 1352, 2016/09
Times Cited Count：3 Percentile：28.28(Nuclear Science & Technology)Our previous study investigated the rewetting behavior of dryout fuel surface during transients beyond anticipated operational occurrences (AOOs) for BWRs, which indicated the rewetting velocity was significantly affected by the precursory cooling defined as cooling immediately before rewetting. The present study further investigated the previous experiments by conducting additional experimental and numerical heat conduction analyses to characterize the precursory cooling. For the characterization, the precursory cooling was firstly defined quantitatively based on evaluated heat transfer rates; the rewetting velocity was investigated as a function of the cladding temperature immediately before the onset of the precursory cooling. The results indicated that the propagation velocity appeared to be limited by the maximum heat transfer rate near the rewetting front. This limitation was consistent with results of the heat conduction analysis.
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.
Liu, W.; Onuki, Akira; Tamai, Hidesada; Akimoto, Hajime
Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 5 Pages, 2005/10
In this research, the newest version of critical power correlation for tight-lattice rod bundles is proposed by using 7-rod and 37-rod bundle data derived in Japan Atomic Energy Research Institute (JAERI). For comparatively high mass velocity region, the correlation is written in local critical heat flux - critical quality type. For low mass velocity region, it is written in critical quality - annular flow length type. The correlation is verified by JAERI data and Bettis Atomic Power Laboratory data. It is confirmed the correlation is able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The correlation is further implemented into TRAC code to analyze flow decrease and power increase transients. It is confirmed transient BT can be predicted within the accuracy of the implemented critical power correlation.
Liu, W.; Kureta, Masatoshi; Akimoto, Hajime
JSME International Journal, Series B, 47(2), p.299 - 305, 2004/05
Experimental research on critical power in tight lattice bundle that simulates the Reduced-Moderation Water Reactor (RMWR) has been carried out in Japan Atomic Energy Research Institute (JAERI). The bundle consists one center rod and six peripheral rods. The 7 rods are arranged on a 14.3 mm equilateral triangular pitch. Each rod is 13 mm in outside diameter. An axial 12-step power distribution is employed to simulate the complicate heating condition in RMWR. Experiments are carried out under 
= 100-1400 kg/m
s, 
= 2-8.5 MPa. Effects of mass velocity, inlet temperature, pressure, radial peaking factor and axial peaking factor on critical power and critical quality are discussed. Compared with axial uniform heating condition, the axial non-uniform heating condition causes an obvious decrease in critical quality. Arai correlation, which is the only correlation that has been optimized for tight lattice condition, is verified with the present experimental data. The correlation is found to be able to give reasonable prediction only around RMWR nominal operating condition.
Onuki, Akira; Shibata, Mitsuhiko; Tamai, Hidesada; Akimoto, Hajime; Yamauchi, Toyoaki*; Mizokami, Shinya*
Nihon Konsoryu Gakkai Nenkai Koenkai 2003 Koen Rombunshu, p.35 - 36, 2003/07
Analytical evaluation of maximum critical power by so-called subchannnel code is indispensable for design of reduced moderation water reactor. In this study, two-phase flow distribution in a tight-lattice rod bundle is investigated using 19-rod bundle experimental rig and subchannnel analysis code NASCA. The flow distribution was measured under so-called churn flow regime and the predictive capability of NASCA was assessed. NASCA can predict the flow distribution qualitatively depending on local pressure drop. Quantitative prediction is also reasonable for liquid phase but the gas phase distribution was underestimated. Void-drift model has a dominant contribution and we should improve the model for the tight-lattice rod bundle.
Iguchi, Tadashi; Iwaki, Chikako*; Anoda, Yoshinari
JAERI-Research 2001-060, 91 Pages, 2002/02
JAERI-Research-2001-060.pdf:6.34MB
no abstracts in English
Iguchi, Tadashi; Watanabe, Hironori; Kimura, Mamoru*; Anoda, Yoshinari
JAERI-Research 2001-032, 111 Pages, 2001/05
JAERI-Research-2001-032.pdf:4.14MB
no abstracts in English
Iguchi, Tadashi; Ito, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru*; Anoda, Yoshinari
JAERI-Data/Code 2001-013, 502 Pages, 2001/03
JAERI-Data-Code-2001-013.pdf:32.38MB
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
*; *; Fujii, Terushige*; *; Matsubayashi, Masahito; Tsuruno, Akira
Nuclear Instruments and Methods in Physics Research A, 377, p.115 - 118, 1996/00
Times Cited Count：15 Percentile：76.22(Instruments & Instrumentation)no abstracts in English
Kumamaru, Hiroshige; Murata, Hideo; ; Kukita, Yutaka
The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE),Vol. 1, 0, p.217 - 222, 1995/00
no abstracts in English
Kumamaru, Hiroshige; ; Murata, Hideo; Kukita, Yutaka
Nucl. Eng. Des., 150, p.95 - 105, 1994/00
Times Cited Count：35 Percentile：92.13(Nuclear Science & Technology)no abstracts in English
Guo, Z.*; Kumamaru, Hiroshige; Kukita, Yutaka
JAERI-M 93-238, 20 Pages, 1993/12
no abstracts in English
Kumamaru, Hiroshige; Kukita, Yutaka
ANS Proc. 1991 National Heat Transfer Conf., Vol. 5, p.22 - 29, 1991/00
no abstracts in English
Anoda, Yoshinari; Kukita, Yutaka; *
Advances in Gas-liquid Flows,1990, p.283 - 289, 1990/11
no abstracts in English
; Koizumi, Yasuo; Tasaka, Kanji
Nucl.Eng.Des., 102, p.71 - 84, 1987/00
Times Cited Count：18 Percentile：83.5(Nuclear Science & Technology)no abstracts in English
; ;
Journal of Nuclear Science and Technology, 23(7), p.658 - 660, 1986/00
Times Cited Count：1 Percentile：28.17(Nuclear Science & Technology)no abstracts in English
