Experimental Study of Water Flow at Supercritical Pressures

摘要

Future supercritical water cooled nuclear reactors will operate at a coolant pressure close to 25 MPa and at outlet temperatures ranging from 500°C to 625°C, i.e., above the critical pressure and temperature of the water (22.06 MPa and 373.95 °C, respectively). Using coolant pressures higher than critical values, avoid boiling and eventual critical heat flux to occur. In addition, the outlet flow enthalpy in these reactors will be much higher than those of actual ones. Consequently, the increase of this thermodynamic property should permit overall nuclear plant efficiencies of up to 48% to be achieved. However, under such flow conditions, thermalhydraulics behaviors of supercritical water are not fully known, i.e., pressure drop, forced convection and heat transfer deterioration, critical and blow down flow rate, etc. Up to now, only a very limited number of studies in these areas have been performed. In particular, the knowledge of critical (choked) discharge of supercritical fluids is mandatory to perform nuclear reactor safety analyses and to design key mechanical components (e.g., control and safety relief valves, etc.). Nevertheless, existing choked flow data have been collected from experiments at atmospheric discharge pressure conditions and in most cases by using working fluids different than water. For this reason, an experimental supercritical water facility has been built at Ecole Polytechnique de Montreal. In this paper, preliminary results obtained using this facility are presented and discussed.