Materials Degradation and Aging August 2015 IN USE: BWR AND PWR IRRADIATED MATERIALS TESTING AND DEGRADATION MODELS FOR REACTOR INTERNALS ISSUE STATEMENT e performance of BWR and PWR reactor internals is affected by several irradiation-based degradation mecha- nisms: irradiation–assisted stress corrosion cracking (IASCC), irradiation embrittlement (reduction in ductility and fracture toughness), creep, stress relaxation and void swelling. Limited understanding of the factors affecting these degradation modes (such as dose, dose rate, stress intensity factors, effects of specimen size and orientation, solute additions, temperature and environmental conditions) could impact decisions related to extended operation up to 80 years. Knowledge gaps exist, for example, on the effects of high fluence on degradation of base metal, welds and heat affected zones in BWR and PWR environments. Experi- mental information is also lacking that could lead to the development of more robust, fundamentally based models and radiation resistant materials. DRIVERS Asset Management Improved understanding of irradiation effects can signifi- cantly affect decisions related to repairs, replacements and even overall unit life. Improved understanding of irradia- tion-induced degradation based on additional data and more accurate models can be used to make sound repair or replace- ment decisions and avoid unanticipated outages associated with cracking of internals. Regulatory and Industry Commitments Regulatory commitments during license renewal typically require plants to implement an aging management program that follows industry guidance. Such programs are based on the technical foundation established in the inspection and evaluation guidelines developed and maintained through EPRI research. is work also supports industry commit- ments through efforts such as the NEI-03-08 Materials Initiative. RESULTS IMPLEMENTAT ION e improved data and degradation models developed under this program will be incorporated into industry guidance for managing internals degradation in BWRs and PWRs. e goal is to provide all necessary data and models to support light water reactor operation through 80 years. Examples include: • Improved crack growth rate models and disposition curves for BWR and PWR internals based on an expanded data- base, input from an expert panel and ASME code/regula- tory approval. ese models will support utility decisions on inspection frequency, repair and replacements • Improved IASCC initiation models for PWR internals will be developed based on data from crack initiation tests on materials removed from the retired PWR plants such as Zorita and in-reactor crack initiation tests with peri- odic dynamic loading in the Halden reactor. MRP will also perform crack initiation tests in autoclaves to study the effect of lithium and investigate the effect of dynamic loading on crack initiation (with EDF). e materials include stainless steel base metals, welds and heat affected zones. • Improved fracture toughness models for BWR and PWR internals to support structural margin and integrity analyses. • Improved void swelling model for PWRs benchmarked with results from the Gondole project and further vali- dated through a comparison with other void swelling codes • Radiation-resistant materials for future replacements of internals in existing plants or for new plants (Advanced Radiation Resistant Materials Program) • Development of fundamentals models to link microstruc- ture of irradiated materials with their engineering proper- ties e.g., yield strength, ductility, fracture toughness and IASCC susceptibility PROJECT PLAN Long-term irradiation effects will be characterized by crack initiation and crack growth tests on materials removed from retired plants such as the Zorita PWR in Spain. e Zorita plant was operated for 38 years (26 Effective Full Power Years) and the highest accumulated fluence on the reactor vessel internals is approximately 50 displacements per atom