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Muhammad, I.; Nagatake, Taku; Uesawa, Shinichiro; Ono, Ayako
Proceedings of 21st International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-21) (Internet), 12 Pages, 2025/08
This research aims to validate ACE-3D using data from a two-phase flow experiment. For this purpose, a two-phase flow experiment was conducted in a four-by-four unheated fuel assembly. In the experiment, the time-averaged void fraction distribution was measured using a wire mesh sensor system under high temperatures and high-pressure conditions. The experimental results were analyzed, and the data were visualized to understand better the behavior and characteristics of the two-phase flow in the fuel assembly. A two-phase flow data set is being developed, covering a wide range of experimental conditions, including higher-pressure regions, which can be used for validating thermal-hydraulic codes. Finally, the ACE-3D code was applied to the two-phase flow experiment. The calculation results were then compared to the experimental ones, and the issues were identified for improving ACE-3D in future simulations.
Shibamoto, Yasuteru; Sun, Haomin; Yonomoto, Taisuke
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 10 Pages, 2014/12
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.
Yamamoto, Kazuyoshi; Watanabe, Shukichi; Nagatomi, Hideki; Kaminaga, Masanori; Funayama, Yoshiro
JAERI-Tech 2002-034, 40 Pages, 2002/03
JAERI-Tech-2002-034.pdf:1.97MB
JRR-4, a swimming-pool type research reactor with a thermal power of 3.5MW, attained criticality in July 1998, after replacing its 90% enrichment fuel with a 20% enrichment fuel under the Reduced Enrichment Program. As a part of the program, safety analysis on thermo-hydraulics of the reactor core was conducted on cases including single channel blockage accident. With the conclusion that a certain margin on thermo-hydraulics was necessary, investigation and experiments were carried out with an aim to increase the core flow rate. To increase the core flow, it was carried out to reduce the bypass flow in the core and to increase the primary coolant flow rate from 7m
/min to 8m/min. After flow measurements using a mock-up fuel element, flow velocity of the fuel channel was determined as 1.45m/s as opposed to the designed value of 1.44m/s, and the ratio of core flow to total flow was 0.88, exceeding the value 0.86 used for the safety analysis.This report describes the JRR-4 core flow increase plan as well as the results of the channel flow rate measurement
Onuki, Akira; Akimoto, Hajime
Proceedings of the 8th International Symposium on Flow Modeling and Turbulence Measurements (FMTM2001) (CD-ROM), 7 Pages, 2001/12
Multi-dimensional analyses have been expected with expanding computation resources for gas-liquid two-phase flow. We recently developed models for bubble turbulent diffusion and bubble diameter to predict the phase distribution by a multi-dimensional two-fluid model. This study was performed to verify our model. The verification was performed using databases under diameter; 9 mm to 155 mm, pressure; atmospheric to 4.9 MPa, flow rate; superficial gas velocity = 0.01 to 5.5 m/s and superficial liquid one = 0.0 to 4.3 m/s, fluid combination; air-water or steam-water. Through the assessments, our model was found to be applicable to the wide range of flow conditions including the effect of pipe diameter. The shape of phase distribution and the average void fraction are predicted well qualitatively and quantitatively. Since the model is established using the ratio of bubble diameter to eddy size as a key-parameter, the ratio is one of important parameters to develop the constitutive equations in the multi-dimensional two-fluid model.
Onuki, Akira; Akamatsu, Mikio*; Akimoto, Hajime
Nihon Konsoryu Gakkai Dai-5-Kai Oganaizudo Konsoryu Foramu Hobunshu, p.87 - 92, 2001/09
Multi-dimensional analyses have been expected with expanding computation resources for gas-liquid two-phase flow in a complex geometry such as fuel rod bundles. Japan Atomic Energy Research Institute is developing a numerical analytical method for the geometry effect, which is based on three-dimensional two-fluid model. In this study, a general curvilinear coordinate system was introduced to the two-fluid model code ACE-3D and air-water two-phase flow around a circular cylinder was analyzed. The present method predicts an air concentration to vortex regions behind the cylinder and a temporal fluctuation of vortex intensity; these two phenomena have been observed in experiments. It is clarified that the phenomena depend on a relative relationship between the drag force and the inertia of bubbles due to pressure fields.
Onuki, Akira; Akimoto, Hajime
International Journal of Multiphase Flow, 26(3), p.367 - 386, 2000/03
Times Cited Count：144 Percentile：96.10(Mechanics)no abstracts in English
Kondo, Masaya; Otani, Etsuo*; Nakamura, Hideo; Asaka, Hideaki*; Anoda, Yoshinari
Proceedings of 2nd Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-2), p.344 - 350, 2000/00
no abstracts in English
Takase, Kazuyuki; Kunugi, Tomoaki; ; Seki, Yasushi
Fusion Engineering and Design, 42, p.83 - 88, 1998/00
Times Cited Count：13 Percentile：69.83(Nuclear Science & Technology)no abstracts in English
Onuki, Akira; ; Akimoto, Hajime
Konsoryu Shimpojiumu '98 Koen Rombunshu, p.221 - 222, 1998/00
no abstracts in English
Onuki, Akira; Akimoto, Hajime
Proc. of 1st European-Japanese Two-phase Flow Group Meeting, p.1 - 8, 1998/00
no abstracts in English
Onuki, Akira; Akimoto, Hajime
Proc. of 3rd Int. Conf. on Multiphase Flow (ICMF'98), p.1 - 6, 1998/00
no abstracts in English
Kaminaga, Masanori; Murayama, Yoji; ;
JAERI-Tech 97-043, 63 Pages, 1997/09
no abstracts in English
Kunii, Katsuhiko; Iwamura, Takamichi; Murao, Yoshio
Journal of Nuclear Science and Technology, 34(1), p.21 - 29, 1997/01
Times Cited Count：1 Percentile：14.48(Nuclear Science & Technology)no abstracts in English
Onuki, Akira; Kamo, Hideki*; Akimoto, Hajime
Eighth Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 3, p.1670 - 1676, 1997/00
no abstracts in English
Takase, Kazuyuki; Kunugi, Tomoaki; Fujii, Sadao*; Shibazaki, Hiroaki*
Flow Visualization Image Process. 1997, 1(00), p.185 - 190, 1997/00
no abstracts in English
Takase, Kazuyuki; Kunugi, Tomoaki; Seki, Yasushi
Proc. of Int. Topical Meetig on Advanced Reactors Safety, 2, p.1268 - 1275, 1997/00
no abstracts in English
Onuki, Akira;
Int. J. Multiph. Flow, 22(6), p.1143 - 1154, 1996/00
Times Cited Count：45 Percentile：84.07(Mechanics)no abstracts in English
; ; Fujii, Terushige*; ; ; ; Matsubayashi, Masahito; Tsuruno, Akira
Nuclear Instruments and Methods in Physics Research A, 377, p.153 - 155, 1996/00
Times Cited Count：8 Percentile：58.07(Instruments & Instrumentation)no abstracts in English
in La
CuO
and La
Sr
CuO pellets and their superconducting propertiesSasaki, Yuji
Analytical Sciences, 11, p.1005 - 1008, 1995/12
Times Cited Count：1 Percentile：7.25(Chemistry, Analytical)no abstracts in English
