Experimental study of abrupt discharge of water at supercritical conditions

• Critical flow experiments under supercritical pressures using sharp edged orifice.
• Pseudo-critical temperature estimation equation for water and carbon dioxide.
• Critical flow at supercritical pressures for water and carbon dioxide.
• Critical flow experiments up to 32 MPa and 502 °C.

Future SuperCritical Water-cooled nuclear Reactors (SCWRs) 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). Coolant pressures higher than critical values will be used to avoid boiling and eventual critical heat flux that may occur. In addition, the outlet flow enthalpy in future supercritical water-cooled nuclear reactors will be much higher than those of actual ones, which can increase overall nuclear plant efficiencies of up to 48%. However, under such flow conditions, thermal–hydraulic behaviors of supercritical water are not fully known, i.e., pressure drop, the deterioration of forced convection heat transfer, critical (choked) flow, blow-down flow rate, etc. In particular, the knowledge of critical discharge of supercritical fluids is mandatory to perform nuclear-reactor safety analyses and to design key mechanical components. Nevertheless, existing choked-flow data have been collected from experiments at atmospheric discharge pressure conditions, but in most cases using working fluids different than water. Therefore, a supercritical water facility has been built at the École Polytechnique de Montréal. In this paper, a new database containing 524 data points is obtained using this facility and compared with available information from the open literature.