Investigation of prototypical corium interaction with samples of VVER-1000 in-vessel and ex-vessel structures

Abstract

After the Fukushima Daiichi accident, many countries worldwide conducted stress tests for nuclear power plants (NPPs) in operation. These tests initiated back-fitting activities. One of the long-term back-fitting tasks for VVER-1000 reactors involves managing the ex-vessel phase of a hypothetical severe accident (SA). The challenge is to cool down the corium ejected into the reactor cavity after a reactor pressure vessel (RPV) failure by spreading it onto the dedicated area. However, to transport the corium to this area, the thermal shielding must withstand the load of poured corium. The thermal shielding ability must be investigated to maintain this strategy successfully. Moreover, the internal RPV parts may prolong the time of RPV failure and decrease the temperature of the corium ejected from the failed RPV. This phenomenon needs to be verified experimentally. Therefore, the interaction of simulated corium with the internal RPV parts and thermal shielding samples was studied experimentally. These experimental findings may help to understand the corium behavior and its cooling during severe accidents in VVER-1000 reactors.