Thermal-hydraulic design of a boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. Then, when the reactor of a new design is developed, 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 BWRs using an innovative two-phase flow simulation technology. For this design method, we are developed an advanced interface tracking method that improves fluid volume conservation, to enable high accuracy prediction of two-phase flow fluid mixing phenomena in the fuel bundles. It was incorporated in the detailed two-phase flow simulation code: TPFIT. And the vectorization and parallelization of TPFIT code was conducted to analyze enormous amounts of data. In this study, to verify the TPFIT code performance, the TPFIT code was applied to the air-water and steam-water bubbly two-phase flow in various flow channels and the numerical results were compared with experimental results. Furthermore, the numerical results applied to the fluid mixing phenomena in boiling water reactor rod bundles are shown, and the existing correlations for the fluid mixing phenomena are evaluated by use of these results.
Development of Two-Phase Flow Correlation for Fluid Mixing Phenomena in Boiling Water Reactor
