Abstract

Although a complex degradation mechanism, the precursors for Stress Corrosion Cracking (SCC) are simply tensile stress, susceptible material and a corrosive environment as expressed by the textbook Venn Diagram. The majority of in-service failures in civil nuclear reactors relating to SCC of stainless steel components have occurred in sensitized material located in occluded and/or stagnant regions exposed to contaminants or oxygenated conditions etc. However, recent operating events have highlighted the risk of SCC of cold worked austenitic stainless steels exposed to good quality primary coolant. Good quality PWR primary coolant is typically high purity water that is pH controlled via lithium hydroxide dosing, containing sufficient dissolved hydrogen to suppress water decomposition by radiolysis and thereby maintain deoxygenated conditions. This paper explores the thresholds for initiation and growth of SCC in good quality water and outlines a methodology for the management of plant risks through:

a) Review of plant components versus the key risk indicators — fabrication processes that introduce cold work, elevated surface hardness, high operational loads, chemical composition and primary circuit operating temperatures.

b) SCC growth rate calculations for high-risk sites to determine if SCC initiation could pose a threat to integrity over the plant lifetime.

c) Review of the adequacy of existing In-Service Inspection (ISI) regimes for defense-in-depth monitoring for signs of SCC.

d) Deployment of supplementary optimized inspection techniques for high risk sites.

e) Factoring of operational experience and examination of decommissioned plant components.

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