REFLOOD THERMAL-HYDRAULICS TESTING USING THE NRC-PSU ROD BUNDLE HEAT TRANSFER (RBHT) TEST FACILITY

Evaluation of thermal-hydraulics codes against experimental data for nuclear reactor safety is an integral step in verification of code accuracy. An international study on reflood thermal-hydraulics, sponsored by the Nuclear Energy Agency (NEA) Working Group on Accident Management and Analysis (WGAMA), has been conducted, with data collected in the Rod Bundle Heat Transfer (RBHT) test facility, located at the Pennsylvania State University. A series of sixteen benchmark tests were conducted, with conditions covering a carefully selected range of oscillatory, variable stepped, and constant rate reflood injection velocities. These unique conditions are useful for code validation and model improvement. Many of the test conditions were chosen to investigate the separate effects of important parameters, such as reflood liquid subcooling, reflood rate, and system pressure. A simulated decay power profile was tested in two test conditions and two tests examined an increase and decrease in system pressure. The effect of 2.5 cm/s ± 2.5 cm/s oscillatory reflood frequencies of 0.25 and 0.5 Hz was also examined. It was observed that for a given reflood velocity, lower liquid subcooling extends overall bundle quench time and increases steam generation rate within the bundle. Lower subcooling also decreases bundle liquid inventory and reduces peak rod temperatures. The addition of flow oscillations increased carryover and extended overall quench times. An increase in system pressure of 137 kPa resulted in suppressed steam flow, increased bundle liquid inventory and decreased overall quench time. In contrast, when the system pressure was decreased by 137 kPa, steam flow increased, bundle liquid inventory decreased and overall quench time increased. It is anticipated that additional analysis of the data and test results will lead to improved reflood thermal-hydraulics models and correlations.