Mechanism and Estimation of Fatigue Crack Initiation in Austenitic Stainless Steels in LWR Environments (NUREG/CR-6787, ANL-01/25)
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Manuscript Completed: September 2001
Date Published: August 2002
Argonne National Laboratory
9700 South Cass Avenue
Argonne, Illinois 60439
J. Muscara, NRC Project Manager
Division of Engineering Technology
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
NRC Job Code Y6388
The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life (ε-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of fatigue crack initiation in austenitic stainless steels in LWR coolant environments. The existing fatigue ε-N data have been evaluated to establish the effects of key material, loading, and environmental parameters (such as steel type, strain range, strain rate, temperature, dissolved-oxygen level in water, and flow rate) on the fatigue lives of these steels. Statistical models are presented for estimating the fatigue ε-N curves for austenitic stainless steels as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are presented. The influence of reactor environments on the mechanism of fatigue crack initiation in these steels is also discussed.