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

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).

We will ensure that relevant groups and the nuclear industry benefit from this research by continuing and expanding our current collaborations, attending international conferences, holding targeted workshops and publishing in leading international journals. More importantly, this project will also further develop our expertise and international leadership, train of the next generation of expert scientist/technicians in a field critical to the energy security of the UK and provide the nuclear industry with a powerful research methodology.
Our research on SCC of austenitic alloy will have a direct impact on the safe operation and reliability of the new fleet of nuclear reactors in the UK. As an example, our ongoing relationship with EDF, collaborator in this project and responsible for the EPR reactor(s) at Hinkley Point C, will ensure that an efficient knowledge transfer and a two-way communication are in place.
The principal pathways to impact of this project will emerge through interactions with the project partners: the MAI/EDF (France), INSS (Japan) and other collaborating companies (already involved in collaborative projects) such as Rolls Royce (UK) and AMEC Foster Wheeler (UK). In addition, Oxford and Manchester have already strong links with the nuclear industry, which will directly benefit from any novel data or results through our frequent regular meetings, e.g. with NNL (UK), Areva (France), SENPEC (China) or EPRI (USA). In addition, we have setup a collaboration with Oxford Instruments that will give us access to state-of-the-art detectors and data analysis software which we will use in this project. These partners have expressed their commitment to this project through total in-kind contributions totaling £1.2M as outlined in the statements of support attached to this proposal.
Both the School of Materials at Manchester and the Department of Materials at Oxford University have large and successful school outreach programs. Through these programs, we will interact and engage with school children through either visiting schools or through University open days. This is a core component in the delivery of pathways to impact as it is vital that we excite the next generation of nuclear scientists at an early age.
Interactions with universities undergraduate students will also be undertaken. In particular, undergraduates will be encouraged to participate in Masters projects (at no cost to the project) in the field of environmental degradation in materials. There are already several projects proposed for this and next academic year. This is a significant pathway to impact aimed at generating more nuclear materials scientists and engineers and counteracting the general lack of specialists the UK is currently facing.
A project website will be set up, linked to the existing nuclear related projects based at Oxford University e.g. the Materials for Fusion and Fission Power website (http://mffp.materials.ox.ac.uk/), and the Bristol-Oxford Nuclear Research Centre. It will be expected that the public areas of these websites will be cross-linked highlighting the important role materials science has in the design and engineering of new nuclear builds.
The dissemination of results obtained as a direct product of this research to academia is a key factor in the success of the project. The area of this proposal - environmental degradation of nuclear materials - is an area in which the publication of results in international journals is highly desirable. Open access publication will be ensured as it is part of the university policy. This is vital in furthering public interest in work of this type as members of the general public will have access to cutting edge research that has previously been inaccessible. In addition, our research will be presented at relevant seminars, workshops and conferences.