The research area's objective is to identify the mechanisms controlling the degradation of materials present in or around nuclear reactors. More precisely, researchers study irradiation-assisted stress corrosion cracking (IASCC) in austenitic stainless steels and stress corrosion cracking (SCC) initiation and other possible long-term degradation models in nickel-based alloys.
This research will enable better lifetime performance predictions, safety assessments, and risk management during extended operation of the existing light water reactor (LWR) fleet.
Develop a mechanistic model for predicting IASCC and develop mitigation measures and remedies for IASCC for current and future LWR plants.
Identify underlying mechanisms controlling SCC crack initiation in nickel-based alloys and evaluate long-term microstructural stability and performance of these alloys under LWR conditions.
Evaluate the microstructural and property changes of an ex-service core internal component. Data and analysis of results will help develop and validate phenomenological models of irradiation damage (swelling, defect evolution, microstructure, and phase stability) under LWR conditions.