This is key area of our research, which focuses more specifically on two aspects of degradation: Environmentally Assisted Cracking(EAC), and water chemistry effects on structural materials.
Key research areas
- Corrosion fatigue
- Hydrogen embrittlement
- Structural integrity
- Weld performance
This research area underpins the continued safe operation of existing current and next-generation reactor systems. Our research provides a mechanistic understanding of environment-material interactions in a broad range of core, plant and advanced materials.
Over the past five years, this research area has greatly benefitted from the addition of advanced microstructural analysis expertise and instrumentation in order to extend our understanding of the nanoscale reactions that impact the performance of austenitic stainless steels and Ni-base alloys in light water reactors. Overall, Plant Materials research in the MPC has doubled in size over the past five years, growing from ten (approximately four academics, two post-docs and four PhD students) to twenty-two, including seven academics, five post-docs and ten PhD students.
The group maintains a strong presence at key international conferences, including the International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, the Nuclear Plant Chemistry Conference, and the four-yearly Fontevraud symposium. The MPC is also an active member of the International Cooperative Group on Environmentally-Assisted Cracking (ICG-EAC), and co-organised the 2017 ICG-EAC symposium.
Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 600 is one of the major challenges for nuclear power plant operation. While extensive research has focused on PWSCC crack growth rates, the initiation stages, from precursor events to embryonic cracks, are undoubtedly most important to study because SCC can be undetected for several decades before a failure occurs during operation. Low pressure high temperature hydrogenated steam exposure technique has been developed to accelerate the oxidation that occurs in high temperature water under conditions where this alloy is known to be susceptible to SCC. Furthermore, this system was successfully employed for the first time in-situ for the characterization and evolution of the preferential intergranular oxidation of Alloy 600. The combination of bulk and in-situ experiments have demonstrated and independently confirmed the observation of Ti and Al oxide enrichment ahead the Cr-rich oxide were the precursor phases for the subsequent formation of intergranular oxide penetration. Oxidation studies in primary water and in steam (both in-situ and ex-situ) demonstrates that the above observations were relevant to initial stages of preferential, intergranular oxidation relevant to nuclear power systems.
High Sulphur (S) content in austenitic stainless steels is intuitively expected to promote more enhanced corrosion fatigue crack propagation in high temperature water. However, it has been shown that in PWR primary coolant environments, high S content in austenitic stainless steels has the effect of reducing corrosion fatigue crack propagation at low crack growth rates via heavy oxidation, homogeneous strain distribution and blunting, as opposed to the general enhancement in crack growth due to hydrogen-induced localised shear.
Extensive effort has been invested in studying the effect of machining on the microstructure of annealed and/or cold-worked stainless steels (304 and 316) as well as correlating SCC initiation in PWR water environment with surface condition. During slow-strain rate tensile (SSRT) tests, initiation occurred in the machined surface and some cracks penetrated into the deformed region. Furthermore, the machining mark orientation appeared to affect SCC initiation while results suggest that post-machining residual stress does not have a measurable effect on SCC initiation susceptibility. Moreover, it was found that a nanocrystalline layer which forms the top of the machined samples reduced the SCC initiation resistance of the annealed material but enhance the SCC initiation resistance of the cold-worked material.
This work, part of NNUMAN, and NUGENIA projects MICRIN and MICRIN+, will be continued in the new Horizon 2020 MEACTOS program.
Understanding of the initiation process in Stress Corrosion Cracking (SCC) is important for the accurate prediction of component lifetime. Surface oxide fracture in 316L grade stainless steel in both the annealed and 1dpa proton-irradiated conditions has been observed during straining in a Scanning Electron Microscope (SEM) and during in-situ straining in a high temperature water environment. Digital Image Correlation (DIC) was used to determine macro-strain and localized strains and related to grain orientation data collected by Electron Back Scattered Diffraction (EBSD) prior to oxidation. At 2.5% applied strain the annealed 316 sample showed only a small number of oxide fractures at grain boundaries, while the proton irradiated sample showed extensive oxide fracture along dislocation channels across grains, with transition to adjacent grains also observed.
The new Horizon 2020 SOTERIA project will build upon these initial results to examine IASCC in 5 to 10 dpa proton-irradiated Type 316L stainless steel.
The major contributor to the radiation dose received by workers within nuclear power plants (NPPs) is the radiation field caused by the highly radioactive 60Co. While there are continued efforts to replace Co-Cr-based StelliteTM alloys, 60Co deposition still dominates the out-of-core radiation fields. Therefore, understanding the oxide growth mechanism under high-temperature aqueous conditions is fundamental to the phenomenon of activity build-up associated with 60Co and to optimizing radiation field mitigation technologies, such as Zn injection. Surface analytical and morphological analysis combined with advanced analytical transmission electron microscopy have been used to study the formation of oxide films on Type 316 SS exposed to normal water chemistry and hydrogen water chemistry BWR coolant.
The deposition of loosely adherent corrosion product (CRUD) in regions of accelerated flow in reactor water circuits has implications including: heat transfer efficiency, incorporation of radioactive species, the axial offset anomaly, and the discrete deposition of corrosion product onto the steam generator tubing of PWRs. An experimental system has been developed to deposit CRUD on discs with micro-orifices (Figure 7) at plant relevant water chemistries, temperatures, and water velocities and accelerations. The modelling of electrokinetic effects in primary water using a new method to simulate the potential distribution has led to a new understanding of the mechanism by which features that disrupt the flow of the solution can give rise to polarisation of the metal surfaces.
Modelling the distribution of potential allows the location of predicted anodic and cathodic regions where CRUD deposition may be promoted or retarded to be mapped and correlated with experimental results.
Since the first detection of irradiation-induced solute clusters using atom probe field-ion microscopy in neutron-irradiated PWR RPV welds in 1984-5, extensive work around the world has been performed to understand the formation of these clusters and how they lead to the degradation in fracture toughness in RPV materials. Over the past 35 years, analytical techniques have improved tremendously – first for atom probe analysis, and more recently for advanced analytical electron microscopy. Thus, it is now possible to generate microstructural data via advanced STEM-EDX spectrum imaging techniques that previously could only be obtained using atom probe analysis/atom probe tomography.
As a potential means of rapidly developing high dose irradiation damage in a variety of structural alloys considered for use in future reactors, a DOE NEUP Integrated Research Program based at the University of Michigan was developed to explore ion irradiation damage and compare the form and extent of damage with neutron-irradiated alloys, and to extend the understanding of ion damage using computational materials modelling. The MPC, via an NEUP-EPSRC programme, applied advanced analytical electron microscopy to assess the development of damage/defects and the solute segregation/clustering and precipitation than can develop during during irradiation.
Burke and Lim led the Alloy 800H characterisation effort, demonstrating that significant solute segregation and defect formation occurred during Fe+ irradiation. Additionally, specimens irradiated in BOR 60 were also characterised, further demonstrating the ability of the MPC to analyse and study neutron-irradiated (active) materials. These observations were corroborated by U Michigan using atom probe tomography.
The evolution of irradiation damage in high Ni (~3.3 wt.%) and conventional (0.8 wt.%) A508 forging steels is of increasing interest due to the impressive fracture toughness afforded by the martensitic/bainitic high Ni A508 Gr4N steels compared to the bainitic A508 Gr3 steels used for RPV applications. Based on the expertise and knowledge of neutron irradiation damage in the Gr4N steels, the use of proton irradiation is being explored to determine whether similar solute-related hardening and defects to those produced via neutron-irradiation are generated. Thus, it is essential to characterise the proton-irradiated microstructures over a range of doses in order to measure the detailed nanoscale solute segregation and clustering that may occur.
This work has required extensive baseline studies and experimental technique development using the Dalton Cumbrian Facility. The results obtained to-date clearly show that proton irradiation to 0.4 and 1.1 dpa is inducing significant Ni redistribution as well as some Mn and Si segregation. This work is continuing utilising the FEI Talos advanced analytical FEG-S/TEM to examine the effect of dose. In addition, nanoindentation is being used to assess the irradiation-induced hardening in the narrow near-surface proton-irradiated zones.
Vanadium-based alloys constitute advanced structural material candidates for the first wall of future magnetically-confined fusion reactors, due to their relatively low cross section for neutron activation. In addition, the body-centred cubic (bcc) nature of the V matrix, with additions of Ti, provides these materials with good resistance to radiation-induced void swelling. These considerations have led to the V-4Cr-4Ti alloy being identified as the prime V-based candidate material for fusion reactor applications.
Using the DCF proton irradiation facility and the advanced analytical TEM capabilities in the EM Centre, we have characterised to atomic resolution the mono-layer thick TiO-type precipitates induced by proton irradiation in a V-4Cr-4Ti alloy at a dose of 0.3 dpa and a temperature of 350°C. Precipitate formation coincided with the coarsening of radiation-induced interstitial a/2〈111〉dislocation loops that were already present at 300°C. The dislocation network induced by prior cold work was mostly recovered at 300°C and 0.3 dpa, and is therefore expected to exert a minimal effect on the precipitate formation. This monolayer-thick precipitate forms at an early stage in the radiation-induced ageing process for V-4Cr-4Ti at low temperatures.