document: Failure of structural components is a major concern in the nuclear power industry and represents not only a safety issue, but also a hazard to economic performance. Stress corrosion cracking (SCC), and especially intergranular stress corrosion cracking (IGSCC), have proved to be a significant potential cause of failures in the nuclear industry in materials such as Alloy 600 (74% Ni, 16% Cr and 8% Fe) and stainless steels, especially in Pressurised Water Reactors (PWR) [1–5]. Stress corrosion cracking in pressurized water reactors (PWSCC) occurs in Alloy 600 in safety critical components, such as steam generator tubes, heater sleeves, pressurized instrument penetrations and control rod drive mechanisms [2,6,7]. Understanding the mechanisms that control SCC in this alloy will allow for continued extensions of life in current plant as well as safer designs of future nuclear reactors.
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