United States Nuclear Regulatory Commission - Protecting People and the Environment

Environmentally Assisted Cracking in Light Water Reactors: Semiannual Report, July 1998 – December 1998 (NUREG/CR-4667, Volume 27)

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Publication Information

Manuscript Completed: October 1999
Date Published:
October 1999

Prepared by:
0. K. Chopra, H. M. Chung, E. E. Gruber
W. E. Ruther, W. J. Shack, J. L. Smith
W. K. Soppert, 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 W6610

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Abstract

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1998 to December 1998. 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 of austenitic stainless steels (SSs), and (c) EAC of Alloys 600 and 690. Fatigue tests have been conducted to determine the crack initiation and crack growth characteristics of austenitic SSs in LWR environments. Procedures are presented for incorporating the effects of reactor coolant environments on the fatigue life of pressure vessel and piping steels. Slowstrain-rate tensile tests and posttest fractographic analyses were conducted on several model SS alloys irradiated to ≈0.3 and 0.9 x 1021 n-cm-2 (E > 1 MeV) in helium at 289°C in the Halden reactor. The results have been used to determine the influence of alloying and impurity elements on the susceptibility of these steels to irradiation-assisted stress corrosion cracking. Fracture toughness J-R curve tests were also conducted on two heats of Type 304 SS that were irradiated to ≈0.3 x 1021 n-cm-2 in the Halden reactor. Crack-growth-rate tests have been conducted on compact-tension specimens of Alloys 600 and 690 under constant load to evaluate the resistance of these alloys to stress corrosion cracking in LWR environments.

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