Study of the deformation mechanisms of zirconium alloys after and under irradiation

In Pressurized Water Reactors, the neutron flux leads to a change in the mechanical properties of the fuel cladding tubes made of zirconium alloys. Although their macroscopic behavior is well known, the microscopic deformation mechanisms of zirconium alloys still need to be characterized. In order to simulate the neutron irradiation, charged particles irradiations (ion and electron) were carried out at 400 C and 450 C on a zirconium alloy: RXA Zircaloy-4. The experimental analyses of the irradiated microstructure, performed by using a Transmission Electron Microscope (TEM), have shown some crystalline defects: dislocation loops with a Burgers vector. Their evolution (size and density) and their characteristics (nature and habit plane) have been determined and discussed based on the point defects diffusion. The results suggest a weak anisotropy in the self-interstitial diffusion. In-situ tensile tests were performed using a TEM, after ion irradiation, in order to activate the dislocation glide and to observe their interaction with the loops. Some of the experimental cases of interaction have been simulate using Dislocation Dynamics for a better understanding of the mechanisms. The simultaneous effects of the stress and of the irradiation on the deformation mechanisms have been then studied. In situ electron and ion irradiations were conducted, with and without an applied stress. Deformation mechanisms involving dislocation climb have thus been demonstrated. Through this study, models based on the identified mechanisms may be suggested, in order to predict the behavior of zirconium alloys in the reactor. (author)