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/CR-4513, 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 °C (535–625 °F). The methodology is applicable to CF-3, CF-3M, CF-8, and CF-8M materials with a ferrite content of up to 40%. The fracture toughness, tensile strength, and Charpy-impact energy of aged CASS materials are estimated from known material information. Embrittlement is characterized in terms of room-temperature (RT) Charpy-impact energy. The extent or degree of thermal embrittlement at “saturation” (i.e., the minimum impact energy that can be achieved for a material after long-term aging) is determined from the chemical composition of the material. Charpy-impact 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 J-R curve for the aged material is then obtained by correlating RT Charpy-impact energy with fracture toughness parameters. A common “predicted lower-bound” J-R 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.
Methodology for Estimating Thermal and Neutron Embrittlement of Cast Austenitic Stainless Steels During Service in Light Water Reactors
Argonne National Laboratory,
Argonne, IL 60439
U.S. Nuclear Regulatory Commission,
Washington, DC 20555
Contributed by the Pressure Vessel and Piping Division of ASME for publication in the JOURNAL OF PRESSURE VESSEL TECHNOLOGY. Manuscript received June 22, 2015; final manuscript received October 19, 2015; published online April 28, 2016. Assoc. Editor: Marina Ruggles-Wrenn.
The United States Government retains, and by accepting the article for publication, the publisher acknowledges that the United States Government retains, a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this work, or allow others to do so, for United States government purposes.
- Views Icon Views
- Share Icon Share
- Cite Icon Cite
- Search Site
Chopra, O. K., and Rao, A. S. (April 28, 2016). "Methodology for Estimating Thermal and Neutron Embrittlement of Cast Austenitic Stainless Steels During Service in Light Water Reactors." ASME. J. Pressure Vessel Technol. August 2016; 138(4): 040801. https://doi.org/10.1115/1.4031909
Download citation file:
- Ris (Zotero)
- Reference Manager