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

英国目前运营着一支民用核舰队，主要由先进的气体反应堆（agr）、压水堆（PWR）和由压水堆组成的核能动力潜艇舰队组成。在可预见的未来，任何新的核电站建设都可能是法国电力公司/阿海珐公司的新的欧洲压水堆（EPR）和/或日立公司的ABWR。在研发投资方面，核工业对安全的承诺是最高的，牛津大学（UoO）和曼彻斯特大学（UoM）在过去几十年里与世界上几乎所有的核能参与者合作，发挥了至关重要的作用。特别是，我们投入了大量精力来了解和预测应力腐蚀开裂（SCC），这是业内最严重的问题之一，也是该项目的主要关注点。SCC是一种渐进的失效模式，它需要特定的环境（冷却水）、应力（施加或残余）和易感材料（不锈钢或镍基合金）。已经提出了几种机制来解释它在核反应堆中的发生，但不幸的是，没有一种机制能够对感兴趣的材料完全解释或预测它。一些最被接受的机制包括优先的晶间氧化，裂纹尖端周围的局部变形或氢脆[1]。50多年前，Henri Coriou[2]确定了一个“更安全”的组合物范围，其中Ni wt%在20到60之间，合金对SCC的影响较小。起初，这项研究没有受到太多关注，但后来被称为“科里欧效应”，最近它已经成为全面审查的主题。Arioka （INSS）对这种效应的有效性进行了广泛的研究，他在不同温度下对一系列具有不同Ni， Fe和Cr水平的样品进行了高压灭菌测试。他的工作得出结论，裂纹扩展速率（cgr）确实受到这些参数（化学成分和/或温度）的强烈影响，从而证实了“科里欧效应”的存在。对这种观察到的行为的机械解释尚未实现，但我们相信我们现在已经能够制定它。如果我们成功了，我们相信我们可以揭示大多数运行的SCC机制及其相互作用。我们的方法包括分离SCC裂纹萌生或扩展中单个变量的影响，这有助于揭示冷加工、水温、合金成分和应力水平对SCC的影响以及低电位条件下（压水堆和ABWR）[5]的控制机制。我们计划使用多技术表征方法，包括最先进的设备和来自牛津大学和曼彻斯特大学的联合专业知识，以更好地了解“科里欧效应”以及H是否起重要作用。拟议的项目将利用INSS.1实物提供的有史以来最雄心勃勃和最全面的一套样品。Lozano-Perez， S., Dohr， J., Meisnar， M. & Kruska， K. . PWRs中的SCC：从自下而上的方法学习。冶金与材料学报，2014,34(2):481 - 481。corou， H., Grall， L., Mahieu， C.和Pelas， M.高镍奥氏体合金的应力腐蚀敏感性和晶间攻击。腐蚀22,280-290（1966）。Feron， D.和Staehle， R. W.水冷核反应堆中镍基合金的应力腐蚀开裂。， 384（2016）。Arioka， K, Yamada， T., Miyamoto， T.和Aoki， M.压水堆一次水中Ni-Cr-Fe合金的晶间应力腐蚀裂纹扩展行为。腐蚀学报，2014,695-707。Meisnar， M., Vilalta-Clemente， A., Moody， M., Arioka， K.和Lozano-Perez， S.: SUS316不锈钢在压水堆一次水中应力腐蚀裂纹扩展速率的温度依赖性机理研究。材料学报，114,15-24（2016）。

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).