United States Nuclear Regulatory Commission - Protecting People and the Environment

Environmentally Assisted Cracking in Light Water Reactors: Semiannual Report, July 1999 – December 1999 (NUREG/CR-4667, Volume 29)

On this page:

Download complete document

Publication Information

Manuscript Completed: September 2000
Date Published:
November 2000

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

Availability Notice


This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from July 1999 to December 1999. 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), (c) EAC of Alloys 600 and 690, and (d) assessment of industry crack-growth models. The fatigue strain-vs.-life data that are available on the effects of various material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs are summarized. Effects of 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 021 n-cm-2 (E > 1 MeV) in He 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 IASCC. Fracture toughness J-R curve tests were also conducted on two heats of Type 304 SS that were irradiated to ≈0.3 and 0.9 x 021 n-cm-2 in the Halden reactor. Crack-growth-rate tests have been conducted on compact-tension specimens of Alloy 690 under cyclic loading to evaluate the enhancement of crack growth rates of these alloys in LWR environments. The existing fatigue crack growth data on Alloys 600 and 690 have been analyzed to establish the effects of temperature, load ratio, frequency, and stress intensity range AK on crack growth rates in air. Predictions of the PLEDGE code for environmentally assisted cracking in stainless steels have been compared with experimental data collected by the BWRVIP, developed at ANL, provided by P. L. Andresen of GE, used to develop the original USNRC disposition curve, and gathered from other sources in the literature. The results indicate that PLEDGE code provides conservative predictions of crack growth rates in unirradiated sensitized materials provided that an appropriate value is chosen for the parameter used to characterize the sensitization denoted by EPR.

Page Last Reviewed/Updated Thursday, August 22, 2013