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

英国目前运营着一支民用核电船队，主要由先进气体反应堆(AGR)、压水堆(PWR)和核动力潜艇压水堆组成。在可预见的未来，任何新的核建设都可能是EDF/阿海珐的新压水堆欧洲加压反应堆(EPR)和/或日立的ABWR。就研发投资而言，核工业对安全的承诺是最高的之一，牛津大学(UOO)和曼彻斯特大学(UOM)在过去几十年里都发挥了至关重要的作用，与世界上几乎所有的核参与者进行了合作。特别是，我们致力于了解和预测应力腐蚀破裂(SCC)，这是行业最严重的问题之一，也将是本项目的主要焦点。应力腐蚀破裂是一种渐进破坏模式，需要特定的环境(冷却水)、应力(施加或残余)和敏感材料(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，没有一种机制能够完全解释或预测感兴趣的材料中的它。一些最被接受的机制包括优先晶间氧化、裂纹尖端周围的局部变形或氢脆[1]。50多年前，Henri Coriou[2]发现了一种“更安全”的成分范围，即Niwt%在20到60之间，在这些成分中，合金不太容易发生应力腐蚀。起初，这项研究没有得到太多的关注，但后来它被称为“科里奥效应”，最近它已经成为全面审查的主题[3]。这一效应的有效性已经被Arioka(INSS)广泛研究，他在不同的温度下用高压灭菌器测试了一系列具有不同镍、铁和铬水平的样品。他的工作得出的结论是，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[4]，从而证实了“科里奥效应”的存在。对这种观察到的行为的机械解释还没有实现，但我们相信我们现在有能力对其进行阐述。如果我们成功了，我们相信我们可以揭开最具操作性的SCC机制及其相互作用的面纱。我们的方法包括隔离SCC裂纹萌生或扩展的单变量影响，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(PWR和ABWR)的控制机制[5]。我们计划使用一种多技术表征方法，包括最先进的设备和牛津大学和曼彻斯特大学的综合专业知识，以更好地了解“科里奥效应”以及H是否扮演重要角色。拟议的项目将利用INSS.1以实物形式提供的有史以来最雄心勃勃和最全面的一套样本。Lozano-Perez，S.，Dohr，J.，Meisnar，M.&Kruska，K.SCC in PWR：从自下而上的方法中学习。冶金与材料学报E1,194-210(2014)2.高镍奥氏体合金的应力腐蚀敏感性和晶间腐蚀敏感性。腐蚀22,280-290(1966)水冷核反应堆中镍基合金的应力腐蚀破裂。，384(2016)。[2]Arioka，K.，Yamada，T.，Miyamoto，T.和Aoki，M..镍铬铁合金在压水堆原水中的晶间应力腐蚀裂纹生长行为.腐蚀70,695-707(2014)。Meisnar，M.，Vilalta-Clemente，A.，Moody，M.，Arioka，K.&Lozano-Perez，S.在压水堆原水中暴露的SUS316不锈钢应力腐蚀裂纹扩展速率的温度依赖性的机理研究。《材料学报》第114期，第15-24期(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).