Thermal-hydraulic design of the current boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. Then, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test that simulates its design is required to confirm or modify the correlations. Development of a method that enables the thermal-hydraulic design of nuclear rectors 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 FLWRs using innovative two-phase flow simulation technology. In this study, detailed two-phase flow simulation code using advanced interface tracking method: TPFIT is developed to get the detailed information of the two-phase flow. In this paper, firstly, we tried to verify the TPFIT code comparing with the existing 2-channel air-water mixing experimental results. Secondary, the TPFIT code was applied to simulation of steamwater two-phase flow in modeled two subchannels of current BWRs rod bundle. 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. From the data, pressure difference between fluid channels is responsible for the fluid mixing, and effects of the time averaged and fluctuating pressure difference must be incorporated in the two-phase flow correlation for fluid mixing.

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