抄録
This study is focused on the development of a two-fluid model gas-liquid two-phase flow simulation code (PORTHOS-MHI). This code was developed to analyze thermal-hydraulic behaviors within a steam generator (SG) tube bundle of a Pressurized Water Reactor (PWR) nuclear power plant. The developed code was verified using the interfacial-velocity and the void-fraction distributions obtained from a tube-bundle experiment of a two-dimensionally full-scale model SG with R-123 as secondary working fluid. Good agreements between the prediction of PORTHOS-MHI and the experimental results of low flow rate (30% full load) conditions were obtained.
