United States Nuclear Regulatory Commission - Protecting People and the Environment

Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels and Alloy 690 from Halden Phase-II Irradiations (NUREG/CR-6965)

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Publication Information

Manuscript Completed: February 2007
Date Published:
September 2008

Prepared by:
Y. Chen, O.K. Chopra, W.K. Soppet, N.L. Dietz Rago
W.J. Shack

Argonne National Laboratory
Argonne, IL 60439

A.S. Rao, NRC Project Manager

Prepared for:
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington DC 20555-0001

NRC Job Code Y6388

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Abstract

This work is an ongoing effort at Argonne National Laboratory on the mechanistic study of irradiation-assisted stress corrosion cracking (IASCC) in the core internals of light water reactors. Phase I1 in this effort focused on determining the influence of grain boundary engineering (GBE), alloying elements, and neutron dose on the IASCC susceptibility of austenitic stainless steels (SSs) and Alloy 690. Flat dog-bone specimens irradiated in the Halden boiling heavy water reactor to neutron doses of =2 displacements per atom (dpa) were subjected to slow strain rate tensile (SSKT) tests in high dissolved oxygen water at 28g°C. The areas of the fracture surfaces with brittle fracture morphology (intergranular and transgranular cleavage cracking) were measured with a scanning electron microscope. All materials showed significant irradiation hardening and loss of ductility. Some strain hardening was observed in the normal-Carbon (C) content SSs (Types 304 and 3 16) and Alloy 690, but not in the low-C content SSs (Types 304L and 316L). The area fraction of the intergranular cracking increased with decreasing uniform elongation and was used to rank the IASCC susceptibility of the alloys. The GBE process did not seem to have a significant effect on the IASCC behavior of Type 304 and 304L SSs, at least at a dose level of about =2 dpa, but did affect the cracking behavior of Alloy 690. A minimum-C content and a low-sulfur (S) content (<0.002 wt.%) may be required for IASCC resistance. By analysis of the results from the Halden Phase-I1 specimens along with previous results on Halden Phase-I specimens, the effect of neutron dose on IASCC behavior was determined. While irradiation hardening and embrittlement start below 0.4 dpa, IG cracking appears between 0.4 and 1.3 dpa for most commercial austenitic SSs in this study.

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