Critical heat flux and post-CHF heat-transfer: An experimental study on boiling crisis with water in the pressure range 15.5–21.0 MPa

Understanding heat transfer under near-critical pressure conditions is essential for the safe design of thermal-hydraulic systems such as future supercritical water reactors. While these systems operate under supercritical conditions during normal operation, subcritical states may occur during start-up, shutdown, or loss-of-pressure accidents. Under such conditions, a boiling crisis may develop if the critical heat flux (CHF) is exceeded, resulting in a sudden deterioration of heat transfer and a sharp increase in heating surface temperature. This behavior poses a significant safety concern, as excessive wall temperatures can result in fuel cladding degradation or failure and compromise overall system integrity. Experimental investigations on the boiling crisis, the associated critical heat flux, and heat transfer under post-CHF conditions have been conducted for several decades, leading to the development of various predictive approaches. However, most studies have focused on pressure ranges relevant to conventional pressurized water reactors. In contrast, experimental data at reduced pressures in the range 0.7 < p r < 1 remain scarce. To address this gap, the present study provides a comprehensive dataset on CHF and post-CHF heat transfer obtained from an industrial-scale test facility operating in this pressure regime. The dataset comprises 176 fully documented experiments using water as working fluid. The systematic variation of experimental parameters enables an interpretation of the contributions of the physical mechanisms governing the onset of the boiling crisis and post-CHF heat transfer with respect to pressure, mass flux, inlet temperature, and heat flux. This approach allows the identification of parameters that favorably or unfavorably influence CHF and characterize their impact on post-CHF heat transfer, thereby supporting the development and validation of improved safety-relevant prediction methods for next-generation nuclear reactor concepts. • Data gap in literature identified for CHF and post-CHF heat transfer. • Scarce data available for reduced pressures (pr > 0.7). • Water dataset generated at pr > 0.7 using an industrial-scale test-rig. • 176 CHF and post-CHF experiments with fully documented datasets. • Parametric trends identified at high subcritical pressures.