Methodology for Estimating Thermal and Neutron Embrittlement of Cast Austenitic Stainless Steels During Service in Light Water Reactors

Abstract

Cast austenitic stainless steel (CASS) materials, which have a duplex structure consisting of austenite and ferrite phases, are susceptible to thermal embrittlement during reactor service. In addition, the prolonged exposure of these materials, which are used in reactor core internals, to neutron irradiation changes their microstructure and microchemistry, and these changes degrade their fracture properties even further. This paper presents a revision of the procedure and correlations presented in NUREG/CR4513, Rev. 1 (Aug. 1994) for predicting the change in fracture toughness and tensile properties of CASS components due to thermal aging during service in light water reactors (LWRs) at 280–330 آ°C (535–625 آ°F). The methodology is applicable to CF3, CF3M, CF8, and CF8M materials with a ferrite content of up to 40%. The fracture toughness, tensile strength, and Charpyimpact energy of aged CASS materials are estimated from known material information. Embrittlement is characterized in terms of roomtemperature (RT) Charpyimpact energy. The extent or degree of thermal embrittlement at “saturationâ€‌ (i.e., the minimum impact energy that can be achieved for a material after longterm aging) is determined from the chemical composition of the material. Charpyimpact energy as a function of the time and temperature of reactor service is estimated from the kinetics of thermal embrittlement, which are also determined from the chemical composition. The fracture toughness JR curve for the aged material is then obtained by correlating RT Charpyimpact energy with fracture toughness parameters. A common “predicted lowerboundâ€‌ JR curve for CASS materials of unknown chemical composition is also defined for a given grade of material, range of ferrite content, and temperature. In addition, guidance is provided for evaluating the combined effects of thermal and neutron embrittlement of CASS materials used in the reactor core internal components. The correlations for estimating the change in tensile strength, including the Ramberg/Osgood parameters for strain hardening, are also described.