| description abstract | For a steam generator (SG) in a commercialized sodiumcooled fast breeder reactor (FBR), flow instability in the water side is one of the most important items needing research. As the first step of this research, thermalhydraulic experiments using water as the test fluid were performed under high pressure conditions at the Japan Atomic Energy Agency (JAEA) by using a circular tube. Void fraction, pressure drop, and heat transfer coefficient data were obtained under 15, 17, and 18 MPa. This paper discusses the steamwater pressure drop and void fraction. Using the obtained data, we evaluated existing correlations for void fraction and twophase flow multipliers under high pressure. As a result, the drift flux model implemented in the TRACBF1 code was confirmed to suitably predict the void fraction well under the present high pressure conditions. For the twophase flow multiplier, the Chisholm correlation and the homogeneous model were confirmed to be the best under the present highpressure conditions. | |
