Numerical Evaluation of Coolability Limits of External Reactor Vessel Cooling Using an Improved Thermal-Hydraulic System Analysis Code

Abstract

The external reactor vessel cooling provides a means to stabilize the molten core inside the reactor vessel during the severe accident in nuclear power plants. For a credible evaluation of the chances of success of in-vessel melt retention, one requires high-fidelity predictions on the natural circulation flow around the reactor vessel and, in particular, the corresponding critical heat flux (CHF) on a hemispherical lower head. In this study, the coolability limit of external reactor vessel cooling in the Korean APR1400 reactor was investigated numerically by improving a thermal-hydraulic system analysis code, MARS-KS1.5. Since such system analysis codes did not incorporate a CHF model for a downward-facing hemisphere, three CHF correlations were newly implemented in the MARS-KS code, two of them taking into account the effect of the mass flux. A transient analysis on the CHF and surface heat flux on the lower head was carried out while the surrounding cooling water was heated to create the two-phase natural circulation flow. The heat transfer mode on the heated surface was used as a measure of evaluating the success of the strategy, and the simulation results were compared with those obtained via the default CHF model in the code. The sensitivity analysis on the heat load from the core melt was also performed.