Experimental Study of Stainless Steel 304l Interaction With Zircaloy at High Temperatures

One of the critical factors in the analysis of the in-vessel retention of corium during a postulated severe accident in a nuclear power plant is the interaction of corium constituents with the reactor vessel wall material. In a CANDU© reactor, after fuel channel disassembly, corium constituents would come into contact with the calandria vessel wall. Ablation of the wall due to physicochemical interaction with corium at high temperature could cause vessel failure, increasing the likelihood of radioactive release to the environment. Therefore, the interaction of the calandria vessel wall material (stainless steel 304L) with each of the main corium constituents (Zircaloy-4, Zr-2.5%Nb, Zircaloy-2, and UO2) was studied experimentally. Arrhenius equations that can be used for analysis of in-vessel corium retention were derived from experimental results. For alloys of Zr, measured rates of interaction were comparable with those reported for similar light-water reactor materials. For UO2, the rate of interaction at 1200 °C was negligible. Experiments performed with a zirconium oxide layer of 10 μm on the Zircaloy samples showed that the oxide acts as a protective barrier against high-temperature interactions with the vessel material. In all cases, if the temperature of the corium–vessel interface remains below the lowest Fe–Zr eutectic temperature of 928 °C, no significant ablation is observed after 24 h.