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

Keith J. Leonard, Maxim N. Gussev, Jacqueline N. Stevens, Jeremy T. Busby

Research output: Contribution to journalArticlepeer-review

17 Scopus citations

Abstract

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 the present work, 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. 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.

Original languageEnglish
Pages (from-to)443-459
Number of pages17
JournalJournal of Nuclear Materials
Volume466
DOIs
StatePublished - Nov 1 2015

Funding

The authors would like to thank Dr. G.O. Ilevbare (EPRI) and Dr. J.L. Nelson (JLN Consulting) for their helpful discussion of the results. This research supported by the U.S. Department of Energy, Office of Nuclear Energy , for the Light Water Reactor Sustainability Research and Development Effort. This manuscript has been authored by the Oak Ridge National Laboratory, managed by UT-Battelle LLC under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan ( http://energy.gov/downloads/doe-public-access-plan ).

Keywords

  • Alloy 718
  • Post-irradiation examination
  • Stress corrosion cracking

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