Microstructural characterizations of austenitic stainless steels representative of PWR internals irradiated with ions and comparison to cluster dynamic simulations

The French nuclear industry is looking into the extension of the operation time of pressurized water reactors (PWR) up to 60 years. This implies a good comprehension of the microstructural evolution under irradiation in Pressurized Water Reactors' conditions.Two representatives stainless steels from PWR's internals, 304 type steels, which differ in carbon content, has been irradiated form low to high doses (more than 100 dpa) at 450 C, irradiation at a second temperature (550 C) has also been performed at an intermediate dose. The choice of the temperature (450 C) was motivated by considering a temperature shift between neutron and ion irradiations due to their large difference in term of dose rate. The microstructural evolution has been characterized by transmitted electron microscopy on each conditions and by atom probe on highest irradiated samples. And modelling of the microstructure was performed using cluster dynamics code CRESCENDO. For both steels, at 450 C the dislocation network and Frank loops reach a saturation regime. As the cavity size and density are low the volume fraction is also low, even in the low carbon content steels, which is more favorable to swelling. The precipitation is mainly carbides. The effects of temperature, chemical composition and of ion irradiation were also investigated. Based on experimental results, a set of parameters which reproduces the evolution of the microstructure in respect to the dose and the depth of observation has been established. It has allowed to understand the effects of the irradiated surface, the injected interstitials and a possible evolution of the cascade efficiency along the damage profile. (author)