The UK has a strong commitment to reduce its reliance on fossil fuels which requires increasing and diversifying the contribution from clean energy sources. As a mature and clean energy production method, nuclear power will no doubt make a significant contribution to the UKs energy portfolio. The UK currently operates a civil nuclear fleet consisting primarily of advanced gas reactors (AGRs) and a nuclear-energy powered submarine fleet of pressurized water reactors (PWRs). Two European Pressurized Reactor (EPR) are currently under construction. To ensure a safe and efficient operation, problems affecting the reactor structural components need to be addressed and understood. The nuclear industry has one of the highest commitments to safety in terms of investment in R&D and Oxford has played a vital role in the last decades by collaborating with almost any nuclear player in the world. In particular we have dedicated a great effort to understand and predict stress corrosion cracking (SCC), which will be the main focus of this project.
SCC is a progressive failure mode which requires a corrosive environment (cooling water), stress (applied or residual) and a susceptible material (stainless steels or Nickel alloys). All can be found in the primary and secondary cooling circuit in PWRs where several metallic components have been reported to fail by SCC. Over the last 60 years, 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 oxidation, local deformation around the crack tip or Hydrogen embrittlement.
Over the last decade, we have concentrated our efforts in isolating the effects of single variables in SCC crack initiation or propagation. This, in our opinion, is crucial, since the phenomena is already complex enough on its own, due to the presence of a complex microstructure, water chemistry and stress. This endeavour has been carefully planned and performed through collaborations with INNS (Japan), EDF (France), Areva (France), Framatome (UK), AECL (Canada), EPRI (USA) and PNNL (USA). We now have a much more reliable and validated understanding on the effect of cold-work, water temperature, alloy composition and stress level on SCC than before and the controlling mechanisms under PWR conditions are finally emerging.
Over the years, our approach has provided with the much needed high-resolution characterization from crack tips, grain boundaries and surface oxides. For that purpose, we have developed new sample preparation and characterization techniques that are now widely used by the community. We have optimized a range of techniques that allows a truly multi-scale characterization of SCC. They include atom-probe tomography (APT), Scanning and Transmission Electron Microscopy [(S)TEM], Scanning Auger Microscopy (SAM), Secondary Ion Mass Spectroscopy (SIMS), Focussed Ion Beam (FIB) 3D tomography, electron tomography, low-voltage scanning electron microscopy (SEM), electron back-scattered diffraction (EBSD), both in reflective and transmission mode, and thermal desorption spectroscopy (TDS) .
We have now greatly narrowed down the list of candidate mechanisms for SCC in PWRs and, through an on-going collaborative project with INSS (Japan), have access to a key matrix of samples that promise to deliver the most conclusive results to date. INSS has autoclave tested a series of samples with varying Ni, Fe and Cr levels at different temperatures. Crack growth rate has been found to be strongly affected by these parameters (chemical composition and/or temperature) and preliminary results suggest that we have a well-established range of techniques to identify the operating mechanisms.
This project aims to identify the operating mechanisms for SCC as well as its impact in the different commercial alloys tested and will run between 2018 and 2021.