Abstract
Thermal-hydraulic design of the current boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. However, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such 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 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 calculate the detailed information of the two-phase flow. We tried to verify the TPFIT code by comparing it with the 2-channel air-water and steam-water mixing experimental results. The predicted result agrees well the observed results and bubble dynamics through the gap and cross flow behavior could be effectively predicted by the TPFIT code, and pressure difference between fluid channels is responsible for the fluid mixing.
