Numerical evaluation of fluid mixing phenomena in boiling water reactor using advanced interface-tracking method

改良界面追跡法を用いた沸騰水型原子炉内流体混合現象の数値解析

吉田 啓之  ; 永吉 拓至*; 高瀬 和之; 秋本 肇

Yoshida, Hiroyuki; Nagayoshi, Takuji*; Takase, Kazuyuki; Akimoto, Hajime

Thermal-hydraulic design of the current boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. Development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method using innovative two-phase flow simulation technology. In this study, a detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. In this paper, firstly, we tried to verify the TPFIT code by comparing it with the existing 2-channel air-water mixing experimental results. Secondary, the TPFIT code was applied to simulation of steam-water two-phase flow in a model of two subchannels. The fluid mixing was observed at a gap between the subchannels. The existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data. This data indicates that pressure difference between fluid channels is responsible for the fluid mixing, and thus the effects of the time averaged pressure difference and fluctuations must be incorporated in the two-phase flow correlation for fluid mixing.