Impact of Thermal Ageing Embrittlement on Westinghouse and Combustion Engineering-Designed Pressurized Water Reactor Pressurizers Based on Pressure-Temperature Limit Comparison

Abstract

The low alloy steel pressurizer (PZR) vessels in pressurized water reactor (PWR) nuclear power plants are potentially susceptible to embrittlement due to thermal ageing over the life of the plant (40–80 years). This paper determines the amount of PZR thermal ageing embrittlement, which can be accommodated based on a comparison of PZR and the U.S. NRC approved reactor pressure vessel (RPV) 10 CFR 50, Appendix G Pressure-Temperature (P-T) limit curves for the current operating U.S. PWR fleet. The maximum amount of postulated thermal ageing embrittlement, in terms of a shift in nil-ductility reference temperature (ΔRTNDT), which is permissible before the generic PZR P-T limit curves exceed the NRC-approved RPV P-T limit curves is provided in this paper. The generic P-T limit curves are determined for current operating U.S. PWR representative Westinghouse and Combustion Engineering (CE) PZR designs for various levels of postulated thermal ageing embrittlement. The locations for consideration within the PZR are the cylindrical shell to bottom head girth weld (includes consideration of the adjacent shell longitudinal seam weld), lower head region in the vicinity of the heater sleeve penetrations, and the surge nozzle corner region. The methodology to calculate the PZR P-T limit curves is per 10 CFR 50, Appendix G and the 2017 Edition of ASME Section XI, Appendix G. The PZR thermal ageing ΔRTNDT values determined in this paper could be compared to estimated or empirical values of thermal ageing embrittlement to determine if or when PZR embrittlement may impact a plant’s 10 CFR 50, Appendix G heatup and cooldown P T limit curves, and any primary loop pressure boundary design fracture mechanics evaluations.