Environmentally Assisted Cracking in Light Water Reactors: Semiannual Report, January 2000 – June 2000 (NUREG/CR-4667, Volume 30)

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Publication Information

Manuscript Completed: May 2001
Date Published: June 2001

Prepared by:
O.K. Chopra, H.M. Chung, E.E. Gruber, D.R. Perkins,
W.J. Shack, W.K. Soppet, 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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This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January 2000 to June 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. Fatigue crack growth data and fracture mechanics analysis have been used to examine the fatigue S-N behavior of carbon and low-alloy steels in air and LWR environments. Fatigue life is considered to be composed of the growth of microstructurally small cracks and mechanically small cracks. The influence of reactor environments on the mechanism of fatigue crack initiation is discussed. Data from slow-strain-rate tensile tests and posttest fractographic analyses 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 have been summarized. The results have been used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. Field-cracked boiling water reactor core shroud welds have been analyzed by advanced metallographic techniques to provide a better understanding of the cracking mechanism, failure behavior, and microstructural characteristics of the material. The list of test specimens shipped to the Halden Reactor for Phase-II irradiation experiments is presented. Fatigue crack growth tests were conducted on 1/4-T compact tension specimens of cast SSs in high-purity water at 289°C to establish the test procedure and conditions that will be used for performing crack growth tests on irradiated materials. The resistance of Alloys 600 and 690 to EAC in simulated LWR environments has been evaluated. The existing crack growth data for these alloys under cyclic loads have been analyzed to establish the effects of alloy chemistry, cold work, and water chemistry. The experimental crack growth rates have been compared with growth rates that would be expected in air under the same mechanical loading conditions to obtain a qualitative understanding of the degree and range of conditions that are necessary for significant environmental enhancement of growth rates.

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