<i>2000 F.N. Speller Award Lecture:</i> Stress Corrosion Cracking in Pressurized Water Reactors—Interpretation, Modeling, and Remedies

The aging of water-cooled and moderated nuclear steam supply systems inevitably has given rise to many material corrosion problems of which stress corrosion cracking (SCC) is one of the more insidious. The goal of this paper was to review some of these examples from the author’s experience of interpreting and modeling SCC in pressurized water reactor (PWR) systems. Examples include SCC of nickel-based alloys 600 (UNS N0660), X750 (UNS N07750), and 718 (UNS N07718) and austenitic stainless steels. The last group includes the emerging issue of the effects of neutron irradiation on SCC susceptibility of nuclear core structural internals. The extent of or lack of basic understanding of these problems is described with the options for remedial measures. Analytical approaches that have been devised for component life prediction are highlighted.