IRRADIATION ASSISTED STRESS CORROSION CRACKING OF AUSTENITIC STAINLESS STEEL WWER REACTOR CORE INTERNALS

摘要

This paper aims to review new results regarding Irradiation Assisted Stress Corrosion Cracking (IASCC) of neutron irradiated Ti-stabilized austenitic stainless steel 08Chl8N10T (chemically similar to AISI 321) from WWER 440 reactor's core internals of NPP Greifswald decommissioned after 15 years in service. Two components (core barrel and core shroud basket) irradiated in the LWR conditions (5×10~(-9)- 4×10~(-8) dpa/s, 260-330°C) to doses about 2-5 dpa were used for the testing. IASCC was investigated by Slow Strain Rate Tensile (SSRT) and Crack Growth Rate (CGR) tests in simulated WWER water environment at 320°C. The IASCC presence been demonstrated if detected the presence of areas of mixed intergranular (IG) and transgranular (TG) fracture on fracture surface. The two tests represent different stress strain conditions for IASCC development, namely for the crack initiation. The test results showed that plane stress condition facilitates IASCC initiation in the thick components. The fracture surface observations indicate that IASCC crack grows based on strain-controlled fracture mechanism. The results are compared with other data obtained by SSRT tests on the steel irradiated in fast reactor. Relation between the results on fast and in-service irradiated materials is mentioned, but disparateness in data not allowed any conclusions.