Stress Corrosion Cracking of Austenitic Stainless Steels in Simulated Nuclear Reactor Conditions

Stress corrosion cracking (SCC) susceptibility of selected austenitic stainless steels was investigated within the SAFELIFE and HPLWR-2 projects. The aim was to identify and describe the specific failure mechanisms in simulated reactor conditions, i.e., a combination of slow strain rate and high temperature in different environments. Slow J-R and standard SCC tests of AISI 304L stainless steel with different degree of cold working, were carried out under BWR condition. Slow strain-rate tensile (SSRT) tests of 316L stainless steel were carried out in ultra-pure demineralized supercritical water (SCW) solution. Tested specimens were subjected to fractographic analysis in order to characterize the failure mechanisms. The specimens failed due to a combination of transgranular SCC and transgranular ductile fracture. The ratio of SCC and ductile fracture was affected by the parameters of tests, in particular by temperature, pressure, oxygen content and strain rateJRC.DDG.F.4-Safety of future nuclear reactor