Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Wada, Yuki; Shibamoto, Yasuteru; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 249, p.127219_1 - 127219_16, 2025/10
Times Cited Count：0Wada, Yuki; Shibamoto, Yasuteru; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 239, p.126598_1 - 126598_18, 2025/04
Times Cited Count：3 Percentile：21.83(Thermodynamics)Okagaki, Yuria; Hibiki, Takashi*; Shibamoto, Yasuteru
JAEA-Review 2024-047, 58 Pages, 2025/02
JAEA-Review-2024-047.pdf:2.22MB
In PWR accident scenarios, the injection of water from the ECCS (ECC injection) might result in thermal stratification in the case of the insufficient mixing of cold and hot water and induce a PTS, affecting the RPV integrity. Therefore, PTS is a vital research issue in reactor safety, and its analysis is essential for evaluating the integrity of RPVs, which determines the reactor life. The PTS analysis comprises a coupled analysis between thermal-hydraulic and structural analysis. Especially in the thermal-hydraulic approach, reliable CFD simulations should play a vital role in the future because predicting the temperature gradient of the RPV wall requires data on the transient temperature distribution of the DC. This study reviewed from the viewpoint of the turbulence models most affecting PTS analysis based on papers published since 2010 on single- and two-phase flow CFD simulation for the experiment on PTS performed in the ROCOM, Transient TOPFLOW, UPTF, and LSTF.
Sun, Haomin; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 237, p.126445_1 - 126445_14, 2025/02
Times Cited Count：2 Percentile：36.85(Thermodynamics)Motegi, Kosuke; Shibamoto, Yasuteru; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 237, p.126406_1 - 126406_15, 2025/02
Times Cited Count：1 Percentile：36.85(Thermodynamics)Satou, Akira; Hibiki, Takashi*; Ikeda, Ryo; Shibamoto, Yasuteru
Progress in Nuclear Energy, 180, p.105593_1 - 105593_11, 2025/02
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)During a loss-of-coolant accident in a pressurized water reactor (PWR), there is a risk that pressurized thermal shock (PTS) may occur on the internal wall of the reactor pressure vessel (RPV) due to the flow of emergency core cooling (ECC) water injected into the cold leg that flows into the downcomer. PTS is caused by the rapid cooling of the downcomer wall by the ECC water and is strongly influenced by the temperature of the ECC water, the collision position and velocity of the water jet on the wall, the velocity of the liquid film on the wall, the thickness of the liquid film, and the spread of the downward flow. Therefore, the flow of ECC water discharging from the cold leg to the downcomer may strongly impact PTS events. To help understand this flow phenomenon, we reviewed studies on free outflow from a circular pipe. Experimental findings on the classification of flow conditions, transition conditions between flow conditions, end depth ratio, free surface profile of flow in the circular pipe, and shape of the nappe flowing out from the pipe have been obtained in a form that is almost consistent with each other. In contrast, when considering the flow from the cold leg to the downcomer, it is necessary to deal with the flow field in a specific situation, such as the flow into a narrow gap rather than a free space, the existence of rounded corners at the outlet of the circular pipe, and the influence of steam flow flowing from the core to the cold leg. However, few previous studies consider these factors, so we summarized them as knowledge that needs to be accumulated in the future.
Motegi, Kosuke; Shibamoto, Yasuteru; Hibiki, Takashi*; Tsukamoto, Naofumi*; Kaneko, Junichi*
JAEA-Review 2024-039, 45 Pages, 2024/09
JAEA-Review-2024-039.pdf:2.23MB
Several heat transfer correlations have been reported related to single-phase opposing flow; however, these correlations are based on experiments conducted in various channel geometries, working fluids, and thermal flow parameter ranges. Therefore, establishing a guideline for deciding which correlation should be selected based on its range of applicability and extrapolation performance is important. This study reviewed the existing heat transfer correlations for turbulent opposing-flow mixed convection. Furthermore, the authors evaluated the predictive performance of each correlation by comparing them with the experimental data obtained under various experimental conditions. The Jackson and Fewster, Churchill, and Swanson and Catton correlations can accurately predict all the experimental data. The authors confirmed that heat transfer correlations using the hydraulic-equivalent diameter as a characteristic length can be used for predictions regardless of channel-geometry differences. Furthermore, correlations described based on nondimensional dominant parameters can be used for predictions regardless of the differences in working fluids.
Matsumoto, Toshinori; Hibiki, Takashi*; Maruyama, Yu
International Journal of Energy Research, 2024(1), p.9748588_1 - 9748588_18, 2024/08
Times Cited Count：0 Percentile：0.00(Energy & Fuels)To evaluate the effectiveness of the wet cavity strategy, the authors developed a stochastic evaluation method that considers the uncertainties of the molten material conditions ejected from reactor pressure vessels. The first step was uncertainty analysis using the MELCOR code to obtain the melt condition. Five uncertain parameters related to the core degradation process were chosen. The input parameter sets were generated using Latin hypercube sampling. The second step was analyzing of the melt-behavior using the JASMINE code. The probabilistic distribution of parameters for the JASMINE analyses was determined from the MELCOR analysis results. The initial water depth was set to 0.5, 1.0, and 2.0 m. The debris height was compared with the criterion to judge its coolability. Consequently, the success probability of debris cooling was obtained through a sequence of calculations. The feasibility and technical difficulties in the MELCOR-JASMINE combined analysis were also discussed.
Sun, Haomin; Kunugi, Tomoaki*; Yokomine, Takehiko*; Shen, X.*; Hibiki, Takashi*
Experimental Thermal and Fluid Science, 154, p.111171_1 - 111171_24, 2024/05
Times Cited Count：3 Percentile：64.78(Thermodynamics)Okagaki, Yuria; Hibiki, Takashi*; Shibamoto, Yasuteru
International Journal of Energy Research, 2024, p.5114542_1 - 5114542_37, 2024/04
Times Cited Count：0 Percentile：0.00(Energy & Fuels)Hirose, Yoshiyasu; Abe, Satoshi; Ishigaki, Masahiro*; Shibamoto, Yasuteru; Hibiki, Takashi*
Progress in Nuclear Energy, 169, p.105085_1 - 105085_13, 2024/04
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Hirose, Yoshiyasu; Shibamoto, Yasuteru; Hibiki, Takashi*
Progress in Nuclear Energy, 168, p.105027_1 - 105027_17, 2024/03
Times Cited Count：3 Percentile：43.92(Nuclear Science & Technology)Motegi, Kosuke; Shibamoto, Yasuteru; Hibiki, Takashi*; Tsukamoto, Naofumi*; Kaneko, Junichi*
International Journal of Energy Research, 2024, p.6029412_1 - 6029412_22, 2024/01
Times Cited Count：1 Percentile：51.66(Energy & Fuels)Convection, wherein forced and natural convections are prominent, is known as mixed convection. Specifically, when a forced convection flow is downward, this flow is called opposing flow. Several heat transfer correlations have been reported related to single-phase opposing flow; however, these correlations are based on experiments conducted in various channel geometries, working fluids, and thermal flow parameter ranges. Because the definition of nondimensional parameters and their validated range confirmed by experiments differ for each correlation reported in previous studies, establishing a guideline for deciding which correlation should be selected based on its range of applicability and extrapolation performance is important. This study reviewed the existing heat transfer correlations for turbulent opposing-flow mixed convection and the single-phase heat transfer correlations implemented in the thermal-hydraulic system codes. Furthermore, we evaluated the predictive performance of each correlation by comparing them with the experimental data obtained under various experimental conditions. The Jackson and Fewster, Churchill, and Swanson and Catton correlations (Int. J Heat Mass Transf., 1987) can accurately predict all the experimental data. The effect of the difference in the thermal boundary conditions, i.e., uniform heat flux and uniform wall temperature, on the turbulent mixed-convection heat transfer coefficient is not substantial. We confirmed that heat transfer correlations using the hydraulic-equivalent diameter as a characteristic length can be used for predictions regardless of channel-geometry differences. Furthermore, correlations described based on nondimensional dominant parameters can be used for predictions regardless of the differences in working fluids.
Sun, Haomin; Kunugi, Tomoaki*; Yokomine, Takehiko*; Shen, X.*; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 211, p.124214_1 - 124214_17, 2023/09
Times Cited Count：4 Percentile：37.98(Thermodynamics)Okagaki, Yuria; Shibamoto, Yasuteru; Wada, Yuki; Abe, Satoshi; Hibiki, Takashi*
Journal of Nuclear Science and Technology, 60(8), p.955 - 968, 2023/08
Times Cited Count：3 Percentile：59.85(Nuclear Science & Technology)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：20 Percentile：76.84(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：24 Percentile：73.38(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.
Shen, X.*; Schlegel, J. P.*; Hibiki, Takashi*; Nakamura, Hideo
Nuclear Engineering and Design, 333, p.87 - 98, 2018/07
Times Cited Count：12 Percentile：30.29(Nuclear Science & Technology)Shen, X.*; Sun, Haomin; Deng, B.*; Hibiki, Takashi*; Nakamura, Hideo
International Journal of Heat and Fluid Flow, 67(Part A), p.168 - 184, 2017/10
Times Cited Count：17 Percentile：58.92(Thermodynamics)An experimental study on upward bubbly air-water flows in a vertical large-diameter square duct have been performed by mainly using four-sensor probes. Local measurements of interfacial area concentration (IAC), void fraction, 3D bubble velocity vector and bubble diameter at 3 axial positions were conducted. Although the interfacial area transport equation (IATE) and its bubble coalescence and breakup models have already played an important role in predicting the IAC in general two-phase flow fields, they are mainly developed based on the two-phase flow experimental data taken in round pipes or small diameter channels. To confirm their usability in large-diameter square duct, this study has evaluated the 1D one-group IATE with its 6 sets of bubble coalescence and breakup models with the presently-obtained database. It was found the relative error between the best prediction and the database was 25%.
