Environmentally Assisted Cracking in Light Water Reactors: Annual Report, January – December 2004 (NUREG/CR-4667, ANL-05/31, Volume 35)
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Manuscript Completed: September 2005
Date Published: November 2006
B. Alexandreanu, O.K. Chopra, H.M. Chung,
E.E. Gruber, W.K. Soppet, and W.J. Shack
Argonne National Laboratory
9700 South Cass Avenue
Argonne, Illinois 60439
W.H. Cullen, Jr., NRC Project Manager
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
This report summarizes work performed from January to December 2004 by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs). 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 boiling water reactors (BWRs), (c) causes and mechanisms of irradiation-assisted cracking of austenitic SS in pressurized water reactors (PWRs), and (d) cracking in Ni alloys and welds.
The existing fatigue strain vs. life data are used to establish the effects of various material and loading parameters, such as steel type, material heat treatment, strain range, strain rate, temperature, dissolved–oxygen (DO) level in water, and flow rate, on the fatigue lives of the steels. Statistical models are presented for estimating the fatigue curves as a function of material, loading, and environmental parameters. Methods are described for incorporating environmental effects into the fatigue evaluations of the American Society of Mechanical Engineers (ASME) Code.
The susceptibility of austenitic SSs and their welds to IASCC as a function of the fluence level, water chemistry, material chemistry, and fabrication history is being evaluated. The slow strain rate tensile test results obtained in the present study on steels irradiated to ≈0.45, 1.35, and 3.0 dpa are compared with data available in the literature. The bulk S content provided the only good correlation with the susceptibility to IGSCC in 289 °C water. A two-dimensional map of bulk S and C contents is presented to show the range in which austenitic SSs are either susceptible or resistant to IASCC. Based on the results of this study, an IASCC model has been proposed.
Crack growth rate (CGR) data are presented on Types 304L and 304 SS weld specimens from the heat affected zone (HAZ) before and after they were irradiated to a fluence of !0.75 dpa. Tests were conducted under cyclic loading and long hold–time trapezoidal loading in simulated 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 growth rates are discussed.
Crack growth rate data, obtained in PWR environment, are also presented for Alloy 600 removed from the control rod drive mechanism (CRDM) nozzle #3 from the Davis–Besse reactor and Alloys 182 and 82 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 Alloy 600 to determine the relative susceptibility of these alloys to EAC. A detailed characterization of the material microstructure, tensile properties, and fracture morphology is given.