United States Nuclear Regulatory Commission - Protecting People and the Environment

Environmentally Assisted Cracking in Light Water Reactors: Annual Report, January – December 2003 (NUREG/CR-4667, ANL-05/17, Volume 34)

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Publication Information

Manuscript Completed: May 2005
Date Published:
May 2006

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

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 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 by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2003. 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) evaluation of causes and mechanisms of irradiation-assisted cracking of austenitic SS in pressurized water reactors (PWRs), and (d) cracking in Ni alloys and welds.

Fatigue tests have been conducted on two heats of Type 304 stainless steel (SS) under various material conditions to determine the effect of heat treatment on fatigue crack initiation in these steels in air and LWR environments. Heat treatment has little or no effect on the fatigue life in air and low dissolved oxygen (DO) environment, whereas in a high–DO environment, fatigue life is lower for sensitized SSs.

Crack growth rate (CGR) data were obtained on Type 304L SS (Heat C3) irradiated to 0.3 x 1021 n/cm2, nonirradiated Type 304 L SS submerged–arc weld heat affected zone (HAZ) specimens from the Grand Gulf (GG) reactor core shroud, and a Type 304 SS laboratory–prepared shielded metal arc weld. The irradiated specimen of Heat C3 showed very little enhancement of CGRs in high–DO water. The results for the weld HAZ material indicate that under predominantly mechanical fatigue conditions, the CGRs for the GG Type 304L weld HAZ are lower than those for shielded metal arc (SMA) weld HAZ prepared in the laboratory with Type 304 SS.

Slow-strain-rate tensile (SSRT) tests have been completed in high-purity 289 °C water on steels irradiated to ≈3 dpa. The bulk sulfur (S) content correlated well with the susceptibility to intergranular stress corrosion cracking (IGSCC) in 289 °C water. The irradiation-assisted stress corrosion cracking (IASCC) susceptibility of SSs that contain >0.003 wt.% S increased drastically. These results 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. The IASCC–resistant or susceptible behavior of austenitic SSs in BWR-like oxidizing environment is described in terms of a two–dimensional map of bulk S and carbon (C) contents of the steels.

Crack growth tests were completed on a Alloy 600 round robin specimen and a Alloy 182 weld specimen in simulated PWR water at 320 °C. Under cyclic loading, the CGRs for the weld specimen were a factor of ≈5 higher than those for Alloy 600 under the same loading conditions in air; little or no environmental enhancement was observed. The CGRs obtained with a trapezoidal waveform (i.e., a constant load with periodic unload/reload) were comparable to the average behavior of Alloy 600 in a PWR environment. The cyclic CGRs for the Alloy 600 round-robin specimen show significant environmental enhancement. However, the crack front was U-shaped, indicating that the growth rates were significantly higher near the edge of the specimen than the center.

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