Microstructural Evolution of Neutron Irradiated Stainless Steels - Comparison between Irradiations in Experimental Reactors and in Pressurized Water Reactors

Materials of the core internals of Pressurized Water Reactors (austenitic stainless steels) are submitted to neutron irradiation. Some baffle/former bolt cracking were detected by non-destructive examinations and removed for metallurgical examination. The most likely cause to cracking was deduced to be Irradiation Assisted Stress Corrosion Cracking.To understand the ageing mechanisms associated to irradiation and propose life predictions of the component, irradiation damage is investigated experimentally by Transmission Electron Microscopy for two representative austenitic stainless steels, SA 304L and CW 316. The microstructure of these austenitic stainless steels constitutive of internals structures of PWR is characterized by TEM and compared to the microstructure obtained after irradiation for high doses up to 40 dpa at 320°C in experimental OSIRIS (mixed flux spectrum) and BOR-60 (fast breeders) reactors and at 300°C in experimental SM-2 (mixed flux spectrum) reactor. Only minor differences were detected.