Post-CHF Heat Transfer Experiments and Modeling at Subcooled and Low-quality Conditions

During postulated large break loss of coolant accidents (LB-LOCAs) in a water-cooled nuclear reactor, the reactor is expected to shut down but the fuel cladding temperature would still increase due to the release of the stored energy in the fuel and decay energy of fission products. To prevent the cladding from overheating, the Emergency Core Cooling System will start up to inject makeup water into the reactor core, during which post-critical heat flux (post-CHF) flow regimes could dominate the heat transfer from the cladding to the coolant. The heat transfer characteristics in the post-CHF flow regimes are important for reactor design and safety. In this research, a Post-CHF Heat Transfer (PCHT) test facility was designed and constructed to perform high-pressure (up to 1,000 psi or 6.8 MPa) post-CHF heat transfer experiments at high mass fluxes (up to 2,000 kg/m2-s) with large inlet subcooling values (up to 50 ºC). A COMSOL multiphysics model was developed to inform the test section design. All the instrumentation in the test facility to measure the temperature, pressure, pressure differential, flow rate, void fraction, and flow topology have been tested to ensure their functionality. Shakedown tests have been performed in the test facility, including hydrostatic leak test, pressure control test, subcooled boiling test, hot patch test, gamma densitometer test, and X-ray radiography system test. Reflooding and dry patch tests have been performed to study the wall heat transfer characteristics in the post-CHF flow regimes. In the reflooding tests, the quench curve indicates that the wall temperature has a rapid decrease in the quenched region, during which the heat (stored energy) release from the test section plays a significant role in the total wall heat flux to the fluid. Based on the convective heat transfer coefficient profile, it is believed that inverted annular film boiling (IAFB) regime exists right downstream of the quench front. In the dry patch tests, the convective heat transfer coefficient for the dry patches is larger than those calculated by three heat transfer correlations for IAFB in the literature, especially for conditions of relatively low wall superheats, which indicates that the motion of the dry patch along the heated surface enhances the heat transfer. In addition to the experimental study, some closure models in the two-fluid model for the IAFB regime are benchmarked and improved. A correlation for the liquid-side Nusselt number is developed based on a parametric study performed in this research. A laminar vapor film model and nine existing correlations for the interfacial friction factor are evaluated based on the IAFB experimental data. The comparisons suggest that the interfacial friction factor in the smooth IAFB region is primarily dependent on the gas Reynold number and vapor film thickness and that in the wavy IAFB region, it is dominantly affected by the interfacial waves. To predict the flow regime transition from IAFB to ISFB, a correlation for the critical Weber number is proposed as a function of two non-dimensional parameters, i.e., the Reynolds number and subcooling number. In summary, this research presents a number of original contributions to the field. The PCHT test facility with its unique test capabilities brings many benefits to the community and can assist in further nuclear reactor safety improvement. The improved post-CHF closure models could be used in reactor system analysis to improve the IAFB modeling.PHDNuclear Engineering & Radiological SciencesUniversity of Michigan, Horace H. Rackham School of Graduate Studieshttp://deepblue.lib.umich.edu/bitstream/2027.42/162896/1/qingqliu_1.pd