Irradiation-Assisted Stress Corrosion Cracking Initiation Screening Criteria for Stainless Steels in PWR Systems

Abstract

The Irradiation-Assisted Stress Corrosion Cracking (IASCC) initiation data for austenitic stainless steels in Pressurized Water Reactor (PWR) primary water environments were collected from available research programs and evaluated by the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The objective was to determine the relationship between applied tensile stress, neutron fluence, and initiation of IASCC at nominally constant load. Analysis of the available data shows that the applied tensile stress level for initiation of IASCC decreases with increasing neutron dose in PWR environments above the PWR threshold for IASCC of three displacements per atom (dpa). An apparent asymptotic value between approximately 30 and 35% of irradiated yield strength has been observed for neutron dose levels between approximately 10 and 100 dpa. Maximum testing times up to approximately 5000 h are now available, but these still are several orders of magnitude less than 60–80 year operating times. However, the results from this study can currently be used by the nuclear industry to assess the effects of irradiation on austenitic stainless steels in PWR systems as an indicator of the combination of stress and neutron dose at which IASCC becomes possible, particularly for subsequent license renewal (SLR) evaluations.