Interactions between Stress, Environment, and Microstructure causing Irradiation-assisted Stress Corrosion Cracking in Austenitic Stainless Steels

Irradiation-assisted stress corrosion cracking is a costly degradation mechanism for austenitic stainless steel components in nuclear reactors. This cracking phenomenon involves a complex interaction between the microstructure of the irradiated steel, the high temperature water environment, and the stresses applied to the material. The exact mechanisms involved in this cracking mode are not fully understood. The series of experiments described in this document are intended to elucidate the relationship between these variables in irradiation-assisted stress corrosion cracking. The experiments are generally focused on austenitic stainless steels irradiated in pressurized water reactors allowing high-quality and relevant research to be performed. In some experiments, such materials available at Westinghouse Churchill Site were machined into specimens and sent to the University of Michigan’s Irradiated Materials Testing Laboratory for mechanical testing in simulated reactor coolant conditions. The influence of applied stress, radiation dose, and high temperature water chemistry on both crack initiation and propagation is studied. Crack initiation and growth studies will be conducted in high temperature water containing either LiOH or KOH additions (otherwise the environments will be the same, including pH) which will determine if cracking behavior differs in the presence of the two different alkali ions. The results of these experiments provide insight into the conditions required to initiate and propagate IASCC. This understanding is beneficial to the nuclear industry, where mitigating maintenance costs associated with IASCC could provide substantial economic benefits. It can also inform the design of future reactors to avoid this failure mode