Void fraction and interfacial velocity distributions of gas-liquid two-phase flow just behind an obstacle simply simulating the spacer shape measured by wire-mesh sensors

Clarifying thermal-hydraulic characteristics in a nuclear reactor core is important in particular to enhance the thermo-fluid safety of nuclear reactors. In order to investigate the spacer effect to the bubbly flow behavior in a subchannel of the nuclear reactor, bubble dynamics around the simulated spacer was visually observed and the void fraction and interfacial velocity distributions just behind the simulated spacer was measured experimentally. Bubbly flow behavior dependence upon the spacer shape was clarified quantitatively. (author)