Analysis of the degradation of a PWR control rod during a loss of coolant accident, using the ICARE2 code

Conditions leading to AIC control rod damage during a loss of coolant accident in a PWR geometry, even in absence of violation of the LOCA licensing criteria, are investigated using several versions of the ICARE2 code (IPSN). Before being applied to the reactor case, the code and the modelling procedure are validated against the out-of-pile severe fuel damage experiment CORA-5. Three particular initial configurations are considered for the subsequent control rod damage analysis: nominal control rod and guide tube geometry, zircaloy guide tube bowing with concurrent cladding thickness reduction and finally control rod cladding perforation. For each of these cases the thermal, mechanical and chemical behaviour is presented. Phenomena such as ballooning and cladding failure of fuel rods, guide tube failure, melt relocation and final fluid channel cross-section modification are described. Finally, the conclusions of numerous sensitivity studies are discussed and some suggestions are given for possible improvements of the ICARE2 code.