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Watanabe, Taro*; Takata, Takashi; Yamaguchi, Akira*
Nuclear Engineering and Design, 313, p.447 - 457, 2017/03
Times Cited Count：2 Percentile：19.37(Nuclear Science & Technology)Countercurrent flow limitation (CCFL) in a heat transfer tube at a steam generator (SG) of pressurized water reactor (PWR) is one of the important issues on the core cooling under a loss of coolant accident(LOCA). In order to improve the prediction accuracy of the CCFL characteristics in numerical simulations using the volume of fluid (VOF) method with less computational cost, a thin liquid film flow in a countercurrent flow is modeled independently and is coupled with the VOF method. Then, we have carried out numerical simulations of a countercurrent flow in a vertical tube so as to investigate the CCFL characteristics and compare them with the previous experimental results. As a result, it has been concluded that the effect of liquid film entrainment by upward gas flux will cause the difference in the experiments.
Ono, Ayako; Yamashita, Susumu; Sakashita, Hiroto*; Suzuki, Takayuki*; Yoshida, Hiroyuki
no journal, ,
JAEA is implementing a simulation of a two-phase flow in the reactor core by TPFIT and JUPITER which are developed by JAEA based on the surface tracking method. However, it is impossible to simulate a boiling on the heating surface in the large-scale domain by this type of simulation method since the simulation of boiling based on the surface tracking method needs the fine mesh which sufficiently resolves the initiation of boiling. Therefore, JAEA started to develop the simplified boiling model applied for the two-phase flow in the fuel assemblies. In this study, the simulation results of the convection boiling on vertical heating surface and the comparison between the simulation results and experimental results are shown.
