燃料集合体内のスペーサによって生じる縮拡流の影響を受ける気泡流に関する数値解析

抄録

Clarifying thermal-hydraulic characteristics in a nuclear reactor core is important in particular to enhance the thermo-fluid safety of nuclear reactors. With progress of a technology on Computational Fluid Dynamics (CFD), development of analysis methods to predict numerically complicated phenomena between liquid and gas phases is performed by atomic research institutes and universities in each country. The objective of this study is to clarify the prediction performance of a two-phase flow analysis code TPFIT which was developed by JAEA. In order to validate predicted results, the bubbly flow in a subchannel with/without a grid spacer were measured under the water-air two-phase flow condition. Bubble dynamics in the simulated subchannel with an obstacle were investigated experimentally and numerically and the spacer effect to the bubbly flow behavior was clarified.