United States Nuclear Regulatory Commission - Protecting People and the Environment

Environmentally Assisted Cracking in Light Water Reactors: Annual Report, January – December 2002 (NUREG/CR-4667, ANL-02/36, Volume 33)

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Publication Information

Manuscript Completed: March 2004
Date Published:
June 2005

Prepared by:
O.K. Chopra, H.M. Chung, R.W. Clark,
E.E. Gruber, 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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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 2002. Topics that have been investigated include: (a) environmental effects on fatigue crack initiation in carbon and low–alloy steels and austenitic stainless steels (SSs), (b) irradiation–assisted stress corrosion cracking (IASCC) of austenitic SSs in BWRs, (c) evaluation of causes and mechanisms of irradiation-assisted cracking of austenitic SS in PWRs, and (d) cracking in Ni–alloys and welds.

A critical review of the ASME Code fatigue design margins and an assessment of the conservatism in the current choice of design margins are presented. The existing fatigue ε–N data have been evaluated to define the effects of key material, loading, and environmental parameters on the fatigue lives of carbon and low–alloy steels and austenitic SSs. Experimental data are presented on the effects of surface roughness on fatigue crack initiation in these materials in air and LWR environments.

Crack growth tests were performed in BWR environments on SSs irradiated to 0.9 and 2.0 x 1021 n•cm –2. The crack growth rates (CGRs) of the irradiated steels are a factor of ≈5 higher than the disposition curve proposed in NUREG–0313 for thermally sensitized materials. The CGRs decreased by an order of magnitude in low–dissolved oxygen (DO) environments.

Slow-strain-rate tensile (SSRT) tests were conducted in high-purity 289 °C water on steels irradiated to ≈3 dpa. The bulk S content correlated well with the susceptibility to intergranular SCC in 289°C water. The IASCC susceptibility of SSs that contain >0.003 wt.% S increased drastically. Bend tests in inert environments at 23 °C were conducted on broken pieces of SSRT specimens and on unirradiated specimens of the same materials after hydrogen charging. The results of the tests and a review of other data in the literature indicate that IASCC in 289 °C water is dominated by a crack-tip grain-boundary process that involves S. An initial IASCC model has been proposed.

A crack growth test was completed on mill annealed Alloy 600 in high–purity water at 289 °C and 320 °C under various environmental and loading conditions. The results from this test are compared with data obtained earlier on several other heats of Alloy 600.

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