Experimental study and numerical modelling of the irradiation damage recovery in zirconium alloys

Neutron irradiation damage in zirconium alloys used as fuel cladding tubes for Pressurized Water Reactors in the nuclear industry consists mainly in a high density of small prismatic dislocation loops. During post-irradiation heat treatment thermal annealing of loops occurs. This phenomenon has been investigated by transmission electron microscopy and microhardness tests. It has been shown that the loop density decreases while their mean size increases. Furthermore it was demonstrated that only vacancy loops remain present in the material after a long term annealing at high temperature. A mechanism based on vacancies diffusion has been proposed to explain the loop evolution during annealing. A cluster dynamic model, originally developed to compute the evolution of the microstructure under irradiation, has been adapted to the modelling of the annealing for zirconium alloys. This physically based model reproduces the loop size and density evolution during a large variety of heat treatments and also provides a better understanding of the mechanisms involved in the loop recovery.