United States Nuclear Regulatory Commission - Protecting People and the Environment

Crack Growth Rates of Irradiated Austenitic Stainless Steel Weld Heat Affected Zone in BWR Environments (NUREG/CR-6891, ANL-04/20)

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Publication Information

Manuscript Completed: August 2004
Date Published:
January 2006

Prepared by:
O.K. Chopra, B. Alexandreanu, E.E. Gruber,
R.S. Daum, W.J. Shack

Argonne National Laboratory
9700 South Cass Avenue
Argonne, Illinois 60439

W.H. Cullen, Jr., NRC Project Manager

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

Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of reactor pressure vessels because of their superior fracture toughness. However, exposure to high levels of neutron irradiation for extended periods can exacerbate the corrosion fatigue and stress corrosion cracking (SCC) behavior of these steels by affecting the material microchemistry, material microstructure, and water chemistry. Experimental data are presented on crack growth rates (CGRs) of the heat affected zone (HAZ) in Types 304L and 304 SS weld specimens before and after they were irradiated to a fluence of 5.0 x 1020 n/cm2 (E > 1 MeV) (≈0.75 dpa) at ≈288°C. Crack growth tests were conducted under cycling loading and long hold time trapezoidal loading in simulated boiling water reactor (BWR) environments on Type 304L SS HAZ of the H5 weld from the Grand Gulf reactor core shroud and on Type 304 SS HAZ of a laboratory–prepared weld. The effects of material composition, irradiation, and water chemistry on CGRs are discussed.

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