Post-irradiation SCC investigations on highly-irradiated core internals component materials

After the shutdown of the Chooz A Power Plant, core internals material samples have been made available, particularly a highly irradiated AISI 304 core baffle corner material with a fluence corresponding to 20 years of operation of a PWR. SCC tests (Constant Load Tests and Slow Strain Rate Tensile Tests) have been performed and will be presented in this paper as well as some complementary fractographic investigations. The effects of stress level, initial loading rate, test temperature and test environment will be discussed. Following these results a full time to failure versus stress curve can be proposed for this irradiated material. Intergranular cracking was observed after constant load and SSRT tests at 340°C in simulated PWR water but also in argon, although to a lesser extent. Complementary investigations to clarify the mechanical behaviour of the irradiated steel were also performed and the results will be discussed.