Reactor Core and Primary Systems
Goal
This 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 initiation and other possible long-term degradation models in nickel-based alloys.
Outcome
This research will enable better lifetime performance predictions, safety assessments and risk management during extended operation of the existing light water reactor fleet.
Planned Major Accomplishments
- 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 stress corrosion cracking initiation in nickel-based alloys and evaluate long-term microstructural stability and performance of these alloys under light water reactor 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 light water reactor conditions.
