STRESS CORROSION CRACKING OF IRRADIATED STAINLESS STEELS IN SIMULATED PWR PRIMARY WATER

摘要

Slow Strain Rate Tensile (SSRT) tests on cold-worked 316 stainless steel (S.S.) and 304 S.S. irradiated to 3×10{sup}21 n/cm{sup}2 (E>0.1MeV) were carried out in simulated PWR primary water chemistry at 290, 325 and 340°C in order to determine Stress Corrosion Cracking (SCC) susceptibility of irradiated S.S. in PWR primary water. Segregation of major elements and impurities at a grain boundary was measured using a transmission electron microscope equipped with a field emission electron gun in order to examine the correlation between SCC susceptibility and segregation. The threshold fluence to SCC susceptibility of irradiated S.S. in PWR primary water was estimated to be order of 10{sup}21 n/cm{sup}2 (E>0.1MeV) and decreased with an increase of testing temperature. The depletion of chromium and molybdenum and the enrichment of nickel and silicon at a grain boundary of irradiated S.S. were detected in the irradiated materials. This segregation of major elements and impurities at a grain boundary was considered to be one of possible causes of SCC susceptibility of irradiated S.S. in PWR primary water.