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Austenitic Stainless Steels (ASS) are widespread in primary and auxiliary circuits of Pressurized Water Reactors (PWRs). Moreover, some components suffer stress corrosion cracking (SCC) under neutron irradiation. This degradation could be the result of the increase of hardness and / or the modification of chemical composition at the grain boundary by irradiation. In order to avoid complex and costly corrosion facilities, the effects of radiation hardening on the material are commonly simulated by applying a pre-strain on non-irradiated material prior to stress corrosion cracking tests. The typical features of the cracking process in primary environment at 360°C during CERTs included an initiation stage (composed of a true initiation time and a slow propagation regime leading to a crack depth lower than 50 μm), then a “rapid” propagation stage before mechanical failure. Pre-straining increased significantly CGRs and the mode of pre-straining could strongly modify the crack path. No significant cracking (< 50 μm) was obtained under a pure static loading. A dynamic loading (CERT or cyclic) was required and various thresholds (hardness, elongation, stress) for the occurrence of SCC were determined. An important R&D program is in progress to develop initiation and propagation models for SCC of austenitic SS in primary environment.

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