United States Nuclear Regulatory Commission - Protecting People and the Environment

Environmentally Assisted Cracking in Light Water Reactors: Annual Report, January – December 2005 (NUREG/CR-4667, ANL-06/33, Volume 36)

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Publication Information

Manuscript Completed: August 2006
Date Published:
August 2007

Prepared by:
B. Alexandreanu, Y. Chen, O.K. Chopra,
H.M. Chung, E.E. Gruber,
W.J. Shack, and W.K. Soppet

Argonne National Laboratory
9700 South Cass Avenue
Argonne, Illinois 60439

S. Crane, NRC Project Manager

Prepared for:
Division of Fuel, Engineering and Radiological Research
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 from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs).

Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed.

The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45–3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed.

The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa.

Crack growth rate data, obtained in the pressurized water reactor environment, are presented on Ni-alloy welds prepared in the laboratory or obtained from the nozzle-to-pipe weld of the V.C. Summer reactor. The experimental CGRs under cyclic and constant load are compared with the existing CGR data for Ni-alloy welds to determine the relative susceptibility of these materials to environmentally enhanced cracking under a variety of loading conditions.

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