Analysis of critical points of the In-Vessel Retention safety evaluation

In-Vessel Retention (IVR) strategy for nuclear reactors in case of a Severe Accident (SA) intends to stabilize and retain the corium in the vessel by using the vessel wall as a heat exchanger with an external water loop. This strategy relies on simple actions to be passively taken as soon as SA signal is raised: vessel depressurization and reactor pit flooding. Then, the strategy is successful if the vessel keeps it integrity, which means that the heat flux coming from the corium pool does not exceed the cooling capacity of the External Reactor Vessel Cooling (ERVC) at each location along the vessel wall (no vessel melt-through) and the ablated vessel wall is mechanically resistant. The main uncertainties in this IVR safety evaluation are associated to the thermal load applied from the corium pool to the vessel wall and the resulting minimum vessel thickness after ablation. Indeed, the heat fluxes distribution along the vessel wall is directly dependent on the corium stratification which occurs as aresult of thermochemical interactions in the pool: when liquid steel is mixed with UO2 and partially oxidized Zr coming from the degradation of the fuel and claddings, there is a phase separation between oxide and metal phases due to a gap of miscibility. The primordial impact of the corium behaviour in the lower plenum of the reactor vessel on the IVR safety evaluation was clearly highlighted in the Phenomena Identification Ranking Table (PIRT) on IVR performed in the frame of the European IVMR (In-Vessel Melt Retention) project (Fichot et al., 2019). As a result, the focus is made in this paper on the critical points which impact the value of the minimum vessel thickness orequivalently the maximum heat flux reached at the outer surface of the vessel wall. Efficiency of the ERVC and mechanical resistance of the vessel wall are consequently not discussed here.The main objective is to identify the generic critical situations leading to an excessive heat flux to the vessel wall and the investigation of possible means to avoid them. In this perspective, the calculations of IVR strategy done by the project partners for different reactor designs and accident scenarios were used as a database to identify and understand the occurrence of critical configurations with excessive heat flux to the vessel wall. The results of 25 sequences are used, which correspond to 9 different reactor designs: a generic PWR 900MWe, a PWR 1100MWe with heavy reflector, a generic PWR 1300MWe, a generic Konvoi 1300MWe, a generic German BWR69, aNordic BWR, a BWR-5 Mark II, a VVER1000 and a VVER440/v213. In addition, different SA integral codes (ASTEC, ATHLET-CD, MAAP -combined with MAAP_EDF and PROCOR codes for simulation of lower plenum behavior-, MELCOR and RELAP/SCDAPSIM codes) are used.