Numerical study of thermal-hydraulic and structural parameters effects on CHF in subcooled forced convection flow during external reactor vessel cooling

Advanced nuclear reactors employ External Reactor Vessel Cooling (ERVC) as a crucial measure to safeguard the structural integrity of the Reactor Pressure Vessel's (RPV) lower head and to confine molten material within it during severe accidents. The ERVC technique effectively mitigates severe accidents by efficiently dissipating decay heat from the molten pool located within the lower head, utilizing coolant flow boiling. The key determinant of the upper limit of ERVC cooling capability is primarily contingent on the Critical Heat Flux (CHF). In this study, flow boiling in downward heating curved flow channel under forced convection is numerically investigated using the Eulerian-Eulerian two-fluid model and realizable k-epsilon turbulence model. Compared with the experiment, the maximum deviation of numerically predicted CHF is less than 5%, indicating that the numerical method is reliable. A thorough analysis is conducted on the impact of variable structure and thermal-hydraulic operating conditions on CHF. The results indicate that as the coolant mass flow rate, system pressure, and inlet subcooling increase, the CHF exhibits a corresponding increase. The CHF increases by 56.6% when the mass flow rate increases from 400 to 1000 kg/(m2 & sdot;s). The CHF at the pressure of 0.3 MPa is 16% higher than that of 0.1 MPa. The CHF increases from 1.14 to 1.84 MW/m2 as the inlet subcooling escalates from 10 to 20 K, with a 60% growth rate, while the CHF increases to 2.03 MW/m2 when the inlet subcooling increases from 20 to 50 K, with a 10.33% increase. The maximum difference in CHF for the three flow channel widths (120, 150, and 170 mm) is only 1.7%, indicating that the influence of flow channel widths is insignificant. This study can provide theoretical guidance for the design and optimization of ERVC in advanced nuclear reactors.