Abstract
Supercritical pressure water cooled reactor (SCWR), which is operated at supercritical pressure, is one of the next generation reactors for the purpose of improving economic efficiency and safety. In the SCWR, water pressure passes the critical pressure during startup, shutdown or in case of loss of coolant accident (LOCA). In the near-critical pressure region which the pressure slightly below the critical pressure, critical heat flux (CHF) condition tends to occur at relatively low heat flux and then there is a risk of serious damage of fuel rod due to abrupt rise of surface temperature. In this study, experiments on CHF at near-critical pressure in vertical upward flow inside a tube were conducted to clarify characteristics of CHF at near-critical pressure.
