Environmentally Assisted Cracking in Light WaterReactors: SemiannualReport, July 2000 – December 2000 (NUREG/CR-4667, ANL-01/09, Volume 31)

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Publication Information

Manuscript Completed: October 2001
Date Published:
April 2002

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

Argonne National Laboratory
9700 South Cass Avenue
Argonne, IL 60439

M.B. McNeil, 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 W6610M

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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 July 2000 to December 2000. Topics that have been investigated include (a) environmental effects on fatigue S–N behavior of primary pressure boundary materials, (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690.

The fatigue strain-vs.-life data are summarized for the effects of various material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Effects of the reactor coolant environment on the mechanism of fatigue crack initiation are discussed. Two methods for incorporating the effects of LWR coolant environments into the ASME Code fatigue evaluations are presented.

Slow-strain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to ≈0.9 x 1021 n-cm-2 (E > 1 MeV) in He at 289°C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. A fracture toughness J-R curve test was conducted on a commercial heat of Type 304 SS that was irradiated to ≈2.0 x 1021 n-cm-2 in the Halden reactor. The results were compared with the data obtained earlier on steels irradiated to 0.3 and 0.9 x 1021 n-cm-2 (E > 1 MeV) (0.45 and 1.35 dpa). Neutron irradiation at 288°C was found to decrease the fracture toughness of austenitic SSs.

Tests were conducted on compact-tension specimens of Alloy 600 under cyclic loading to evaluate the enhancement of crack growth rates in LWR environments. Then, the existing fatigue crack growth data on Alloys 600 and 690 were analyzed to establish the effects of temperature, load ratio, frequency, and stress intensity range on crack growth rates in air.

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