When the core-melt accident in a light water reactor (LWR) nuclear power plant occurs, the molten core materials may be relocated into the lower plenum of reactor pressure vessel and threatens the integrity of the reactor vessel. If the lower head of the reactor vessel is breached, the molten core materials are further relocated into the lower part of the containment vessel where various phenomena potentially causing various modes of containment failure are concerned. In-vessel retention by external reactor vessel cooling (IVR-ERVC) is one of the strategies that is to mitigate or terminate the severe accident. If the feasibility of IVR-ERVC is proved with a high level of reliability, the concern on the complicated ex-vessel phenomena might be resolved or much simplified. However, even in the process of IVR-ERVC, a number of complex phenomena such as highly turbulent natural convection of the molten core with internal heat generation, crust formation, natural circulation in the cavity and so on should be considered. Failure mechanism in terms of thermal failure criterion occurs when the heat flux at the vessel wall exceeds the critical heat flux at the same place.

Figure 1. Core degradation.
Figure 2. Conventional configuration of IVR-ERVC.
The objective of this research is to evaluate the capability for the IVR-ERVC by analysis of thermal behavior of molten corium. Thermal load analysis would be conducted by lumped parametric method (LPM) and CFD method. We made a lumped parameter based computer code to evaluate the thermal load on the reactor vessel, and applied the code for a probabilistic analysis for a high power (>1000 MWe) reactor assuming the APR1400 condition. It is possible to investigate overall behavior of thermal load in RPV and the uncertainty analysis could be conducted with risk-significant variables on in-vessel corium pool coolability. The effect of Pr on the flow characteristics and thermal behavior is investigated in highly turbulent natural convection with internal heat source.

Analysis of In-vessel Retention (AIR) code is developed and verified with previous research as shown in figure 3. Uncertainty analysis is performed with major parameters for the high power reactor using the AIR code. Figure 4. shows the cumulative probability distribution of the heat flux distribution along the vessel wall. The highest value appeared in the light metal layer due to the focusing effect.

Figure 3. Comparison of the present code and probabilistic analysis method with the analysis for AP1000.
Figure 4. Cumulative probability distributions of the heat flux on the vessel outer surface for the base case.
Figure 5 shows the result of velocity vector and temperature contour at different Prandtl number in the same Rayleigh number. It can be seen that the flow characteristics and thermal behavior are affected by Pr number; the a stronger flow and more unstable temperature field is observed in all regions in the case of corium.
Figure 5. Effect of Pr on the flow characteristics and thermal behavior. Contour plots of the temperature and velocity for water and corium with algebraic heat flux model. (left) Water, Pr=4.5 (right) Corium, Pr=0.5.