改良界面追跡法によるBWR炉心内流体混合現象の数値解析

抄録

Thermal-hydraulic design of the current BWR is performed by correlations with empirical results of actual-size tests. Then, when the reactor of new design is developed, an actual size test is required to confirm or modify the correlations. 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 for BWRs using innovative two-phase flow simulation technology. In this study, detailed two-phase flow simulation code using advanced interface tracking method : TPFIT is developed. In this paper, the TPFIT code was applied to simulation of two-phase flow in modeled 2 subchannels of BWRs rod bundle, and the existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data.