Understanding the mechanisms controlling low potential stress corrosion cracking in nuclear reactors

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The UK currently operates a civil nuclear fleet consisting primarily of advanced gas reactors (AGRs), a pressurized water reactor (PWR) and a nuclear-energy powered submarine fleet of PWRs. Any new nuclear build in the foreseeable future will likely be a new PWR European Pressurized Reactor (EPR) by EDF/Areva and/or an ABWR by Hitachi. The nuclear industry has one of the highest commitments to safety in terms of investment in R&D and both the Universities of Oxford (UoO) and Manchester (UoM) have played a vital role in the past decades by collaborating with almost every nuclear player in the world. In particular, we have dedicated a great effort to understand and predict stress corrosion cracking (SCC), which is one of the most serious concerns for the industry and will be the main focus of this project.SCC is a progressive failure mode which requires a specific environment (cooling water), stress (applied or residual) and a susceptible material (stainless steels or nickel base alloys). Several mechanisms have been proposed to explain its occurrence in nuclear reactors but, unfortunately, none has been capable of explaining or predicting it fully for the materials of interest. Some of the most accepted mechanisms involve preferential intergranular oxidation, local deformation around the crack tip or hydrogen embrittlement [1].Over 50 years ago, Henri Coriou [2] identified a "safer" range of compositions with Ni wt% between 20 and 60 where alloys would be less susceptible to SCC. Initially, this study did not receive much attention, but it was later known as the "Coriou effect" and most recently it has been the subject of a comprehensive review [3]. The validity of this effect has been extensively investigated by Arioka (INSS), who autoclave-tested a series of samples with varying Ni, Fe and Cr levels at different temperatures. His work led to the conclusion that crack growth rates (CGRs) are indeed strongly affected by these parameters (chemical composition and/or temperature) [4] and thus validated the existence of the "Coriou effect". A mechanistic explanation for this observed behaviour has yet to be realised, however we believe we are now in a position to formulate it. If we are successful, we believe we can unveal most opearting SCC mechanisms and their interplay. Our approach involves isolating the effects of single variables in SCC crack initiation or propagation, which has been instrumental in revealing the effect of cold-work, water temperature, alloy composition and stress level on SCC and the controlling mechanisms under low-potential conditions (PWR and ABWR) [5]. We plan to use a multi-technique characterization approach, involving state-of-the-art equipment and the combined expertise from the universities of Oxford and Manchester, to better understand the "Coriou effect" and whether H plays an important role or not. The proposed project will make use of one of the most ambitious and comprehensive set of samples ever tested, provided in kind by INSS.1. Lozano-Perez, S., Dohr, J., Meisnar, M. & Kruska, K. SCC in PWRs: Learning from a Bottom-Up Approach. Metallurgical and Materials Transactions E 1, 194-210 (2014)2. Coriou, H., Grall, L., Mahieu, C. & Pelas, M. Sensitivity to Stress Corrosion and Intergranular Attack of High-Nickel Austenitic Alloys. Corrosion 22, 280-290 (1966).3. Feron, D. & Staehle, R. W. Stress Corrosion Cracking of Nickel Based Alloys in Water-Cooled Nuclear Reactors. , 384 (2016).4. Arioka, K., Yamada, T., Miyamoto, T. & Aoki, M. Intergranular stress corrosion cracking growth behavior of Ni-Cr-Fe alloys in pressurized water reactor primary water. Corrosion 70, 695-707 (2014).5. Meisnar, M., Vilalta-Clemente, A., Moody, M., Arioka, K. & Lozano-Perez, S. A mechanistic study of the temperature dependence of the stress corrosion crack growth rate in SUS316 stainless steels exposed to PWR primary water. Acta Materialia 114, 15-24 (2016).