Analysis of stress corrosion cracking in alloy 718 following commercial reactor exposure

Journal Article · Mon Aug 24 00:00:00 EDT 2015 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2015.08.039· OSTI ID:1286846

Leonard, Keith J. ^[1]; Gussev, Maxim N. ^[1]; Stevens, Jacqueline N. ^[2]; Busby, Jeremy T. ^[1]

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1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2. AREVA Inc., Lynchburg, VA (United States)

Alloy 718 is generally considered a highly corrosion-resistant material but can still be susceptible to stress corrosion cracking (SCC). The combination of factors leading to SCC susceptibility in the alloy is not always clear enough. In this paper, alloy 718 leaf spring (LS) materials that suffered stress corrosion damage during two 24-month cycles in pressurized water reactor service, operated to >45 MWd/mtU burn-up, was investigated. Compared to archival samples fabricated through the same processing conditions, little microstructural and property changes occurred in the material with in-service irradiation, contrary to high dose rate laboratory-based experiments reported in literature. Though the lack of delta phase formation along grain boundaries would suggest a more SCC resistant microstructure, grain boundary cracking in the material was extensive. Crack propagation routes were explored through focused ion beam milling of specimens near the crack tip for transmission electron microscopy as well as in polished plan view and cross-sectional samples for electron backscatter diffraction analysis. It has been shown in this study that cracks propagated mainly along random high-angle grain boundaries, with the material around cracks displaying a high local density of dislocations. The slip lines were produced through the local deformation of the leaf spring material above their yield strength. Also, the cause for local SCC appears to be related to oxidation of both slip lines and grain boundaries, which under the high in-service stresses resulted in crack development in the material.

• Research Organization: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Center for Nanophase Materials Sciences (CNMS)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE)
• Grant/Contract Number: AC05-00OR22725
• OSTI ID: 1286846
• Alternate ID(s): OSTI ID: 1252268
• Journal Information: Journal of Nuclear Materials, Vol. 466; ISSN 0022-3115
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

36 MATERIALS SCIENCE
22 GENERAL STUDIES OF NUCLEAR REACTORS
Alloy 718
Post-irradiation examination
Stress corrosion cracking