Environmentally Assisted Cracking in Light Water Reactors: Annual Report, January – December 2001 (NUREG/CR-4667, ANL-02/33, Volume 32)

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Publication Information

Manuscript Completed: August 2002
Date Published: November 2002

Prepared by:
O.K. Chopra, H.M. Chung, R.W. Clark, E.E. Gruber,
R.W. Hiller, W.J. Shack, W.K. Soppet, R.V. Strain

Argonne National Laboratory
9700 South Cass Avenue
Argonne, Illinois 60439

W.H. Cullen, Jr., and C.E. Moyer, NRC Project Managers

Prepared for:
Division of Engineering Technology
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

NRC Job Code Y6388

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Abstract

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2001. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs, and (c) EAC of Alloy 600.

The effects of key material and loading variables, such as strain amplitude, strain rate, temperature, dissolved oxygen (DO) level in water, and material heat treatment, on the fatigue lives of wrought and cast austenitic SSs in air and LWR environments have been evaluated. The mechanism of fatigue crack initiation in austenitic SSs in LWR environments has also been examined. The results indicate that the presence of a surface oxide film or difference in the characteristics of the oxide film has no effect on fatigue crack initiation in austenitic SSs in LWR environments.

Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to ≈2 x 10²¹ n·cm^-2 (E > 1 MeV) in He at 289 °C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. Corrosion fatigue tests were conducted on nonirradiated austenitic SSs in high-purity water at 289 °C to establish the test procedure and conditions that will be used for the tests on irradiated materials. A comprehensive irradiation experiment was initiated to obtain many tensile and disk specimens irradiated under simulated pressurized water reactor conditions at ≈325°C to 5, 10, 20, and 40 dpa.

A crack growth test was completed on 30% cold-worked Alloy 600 in high-purity water under various environmental and loading conditions. The results are compared with data obtained earlier on several heats of Alloy 600 tested in high-DO water under several heat treatment conditions.