0613 改良界面追跡法によるBWR炉心内流体混合現象評価手法開発の現状(S45-3 原子炉システムおよびその要素技術(3),S45 原子炉システムおよびその要素技術)

概要

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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 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 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 the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) rod bundle, and the existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data.
2007-09-07