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 that of the TPFIT, the bubbly flow data in a simulated subchannel with/without a spacer were obtained at the water-air two-phase flow condition. Bubble dynamics in the simulated subchannel were visually observed by a high speed camera. Moreover, the void fraction distributions were measured quantitatively using a wire-mesh sensor system. In addition, the spacer effect to the bubbly flow behavior was investigated. It was confirmed from the present experimental and analytical studies that the TPFIT can predict the bubble dynamics which changes from lots of small bubbles into large bubbles qualitatively.
