Transient thermal analysis of IVR strategy under a LB-LOCA based on ASTEC code

A Large-Break Loss of Coolant Accident (LB-LOCA) in a generic, 1000 MWe and 3-loop Chinese Pressurized Water Reactor (PWR) with the In-Vessel corium Retention (IVR) strategy was simulated by using ASTEC V2.1.1. One of the criteria of assessing the validity of IVR is that the decay heat of molten pool could be removed successfully, which means that the local heat flux will not exceed the Critical Heat Flux (CHF) of outer nucleate boiling. ASTEC determines the process of core degradation, the corium relocation and the transient heat flux to indicate whether the vessel can maintain its integrity under the LB-LOCA. From the core degradation process, for this LB-LOCA scenario, the corium with fission products slumps into the lower plenum via the downcomer, which means the sideward relocation. After the initial relocation, the mass of compositions of molten pool in lower head are almost steady except the mass of steel. The augmentation of the steel in the lower head is consistent with the melting process of the lower support plate, the fluid distribution structures and the vessel inner wall. At the azimuth from 78° to 87°, the transient heat flux is much larger compared with the other locations. The transient heat flux reaches its maximum value which is about 1.1 MW/m2 at the azimuth of 84°. And the maximal ratio of the transient local heat flux to CHF is about 0.8. For the residual thickness of the lower head, the minimum is 37.35 mm at the azimuth from 78° to 87°. The results of thermal analysis conclude that the vessel can maintain its integrity during LB-LOCA scenario because the transient heat flux from the vessel wall to the external coolant do not exceed the CHF of outer nucleate boiling