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Journal Articles

Measurement of void fraction distribution at high pressure in 4 4 simulated fuel bundle for validation of thermal-hydraulics simulation codes

Ono, Ayako; Nagatake, Taku; Uesawa, Shinichiro; Shibata, Mitsuhiko; Yoshida, Hiroyuki

Proceedings of Specialist Workshop on Advanced Instrumentation and Measurement Techniques for Nuclear Reactor Thermal-Hydraulics and Severe Accidents (SWINTH-2024) (USB Flash Drive), 7 Pages, 2024/06

Japan Atomic Energy Agency (JAEA) is developing a neutronics/thermal-hydraulics coupling simulation code for light-water reactors. Thermal-hydraulic simulation codes applied to the coupling code are expected to calculate the void fraction distribution in a rod bundle under operational conditions, which are necessary for neutron transport simulation, and need to be validated using void fraction distribution data in a rod bundle under high-temperature and high-pressure conditions. Therefore, we have conducted the measurement of the instantaneous void distribution in the 4 4 simulated fuel bundle using a developed wire mesh sensor, which is installed in the pressurized two-phase flow experimental loop of JAEA to obtain the data for code validation.

Journal Articles

Flow regime and void fraction predictions in vertical rod bundle flow channels

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

https://doi.org/10.1016/j.ijheatmasstransfer.2021.121637

Times Cited Count：19 Percentile：77.45(Thermodynamics)

Journal Articles

Experimental study on local interfacial parameters in upward air-water bubbly flow in a vertical 66 rod bundle

Han, 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

https://doi.org/10.1016/j.ijheatmasstransfer.2019.118696

Times Cited Count：23 Percentile：71.75(Thermodynamics)

Journal Articles

Local gas-liquid two-phase flow characteristics in rod bundle geometry

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.

Journal Articles

Heat conduction analyses on rewetting front propagation during transients beyond anticipated operational occurrences for BWRs

Yonomoto, Taisuke; Shibamoto, Yasuteru; Satou, Akira; Okagaki, Yuria

Journal of Nuclear Science and Technology, 53(9), p.1342 - 1352, 2016/09

AA2015-0497.pdf:1.05MB

https://doi.org/10.1080/00223131.2015.1108882

Times Cited Count：3 Percentile：26.09(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.

Journal Articles

Measurement of void fraction distribution in steam-water two-phase flow in a 44 bundle at 2 MPa

Liu, 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 $^{circ}$ C), 2.1 MPa (215 $^{circ}$ C) and 2.6 MPa (226 $^{circ}$ 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.

Journal Articles

Critical power prediction for tight lattice rod bundles

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.

Journal Articles

Critical power in 7-rod tight lattice bundle

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 $^{2}$ s, $_{ex}$ = 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.

Journal Articles

Study on gas-liquid two-phase flow distribution in a tight-lattice rod bundle

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.

JAEA Reports