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Following its lifetime in-reactor, AGR fuel is discharged and stored pending a decision on the disposition route. Presently, Sellafield store AGR fuel within a caustic dosed pond until future routes, such as geological disposal, may be carried out. During storage, the 20-25Nb steel AGR cladding provides the primary containment for the radioactivity within the fuel. Intergranular attack (IGA) of the fuel cladding does not occur whilst in-reactor as it is too hot and dry, however, during storage within aqueous environments IGA may occur if chromium depletion due to irradiation and/or thermal annealing becomes severe enough. This can lead to intergranular stress corrosion cracking (IGSCC) and eventual cracking of the clad material. The aim of the proposed work is to both characterise AGR fuel cladding to support long term storage of AGR fuel, as well as to evaluate the suitability of a thermally sensitised proxy material (such as 304 steel) for the study of corrosion mechanisms and behaviours. The student would be based in the Interface Analysis Centre (IAC) at the University of Bristol, within the Materials Degradation research group of Dr Tomas Martin. Dr Martin’s group has an established track record in the use of advanced materials characterisation to analysis challenges for nuclear materials including creep cavitation[1], thermal evolution[2], carburisation[3] and corrosion. The project would build on recent work that has utilised the power of the high-speed atomic force microscope (HS-AFM) developed at Bristol IAC to observe the evolution of stress corrosion cracking of steel alloys in real time [5]. During service, stainless steel cladding within AGRs can become sensitised as a result of prolonged irradiation and elevated temperatures. This can result in susceptibility to IGA. IGA observed within sensitised AGR fuel cladding material exhibits different morphologies, dependent upon the conditions under which it occurs. In cases where there is a source of tensile stress, it can lead to IGSCC, which in turn can lead to a greater depth of corrosion. In severe cases, IGA and IGSCC can lead to failure of the cladding, which could result in significant activity release and fission gas release. As such, it is vital to understand the mechanisms occurring at the microscale and the microstructural characteristics of cladding, such that evaluations can be made into the safety of the system. The current focus of mechanistic work is to underpin the extrapolation of current knowledge and observation of fuel stored for around 25 years, up to anticipated future storage timescales. Heat treatments intended to purposefully cause thermal sensitisation in unirradiated stainless steels have been used in previous works as a method of simulating the key characteristics of irradiated stainless steels. Thermal sensitisation has been explored in previous works to render the material susceptible to IGSCC such that the failure mechanism may be studied with comparative ease [7]. Previous works have considered the variation of the severity in GB segregation by RIS to be in part due to the geometry of the GB, that is, the dimensions of the GB and the GB misorientation [6]. Different types of GBs can vary significantly in their properties and behaviours, with different misorientations leading to a different GB energy, chemistry, and mobility. The proportions and morphology of different GBs present within a material at the microscale can alter the mechanical and chemical properties on the macroscale. Similarly, the role of inclusions and precipitates in initiating corrosion is an important factor in understanding the mechanism of IGA, with niobium carbides in 20-25Nb steel often acting as initiation sites for corrosive attack. [4]. In this project, a series of measurements will be performed in order to compare irradiated ex-service AGR 20-25Nb steel cladding samples to thermally sensitised proxy samples. The microstructures in both sample types will be evaluated and compared using scanning electron microscopy (SEM) and transmission electron microscopy (TEM) techniques including electron backscatter diffraction (EBSD) and energy-dispersive X-ray spectroscopy (EDS). The project will also take advantage of the newly funded plasma focused ion beam (PFIB) at Bristol (as well as at NNL central labs in Sellafield for active material) to characterise crystallography and chemistry in 3D. Firstly, by PFIB serial-sectioning and EBSD analysis a 3D reconstruction of the grain structure may be produced, facilitating comparisons of grain boundary (GB) character distribution, preferred grain orientation and grain size. Secondly, site-specific TEM lamellae are to be produced and evaluated by transmission-EBSD and STEM EDS. EBSD analysis of the lamellae allows for identification of the GB misorientations present, and subsequent STEM EDS analysis aims to produce GB depletion profiles. These data may then be analysed to establish a relationship between elemental segregation profiles and GB misorientation. In addition to work on GB sensitisation, SCC observations will be performed by HS-AFM at the University of Bristol on the specimens produced from this proxy material and potentially also the irradiated clad using the NNUF active HS-AFM. By establishing the similarities and differences between the two materials corrosion behaviour in real-time in-situ observations, conclusions drawn from SCC tests may be placed into better context. The proposed study may also aid in the development of accurate predictive models, and the simulation of RIS effects. In addition to this, these results may be used to inform future inspection of ex-service AGR fuel cladding.Advanced characterisation of stress corrosion cracking in long-term aqueous storage of 20-25Nb AGR cladding
This PhD project funded by EDF Energy and the UKRI iCase scheme, will use advanced microscopy to understand the changes in microstructure of the steel reactor pressure vessel (RPV) at Sizewell B (SZB) under long-term thermal ageing in-service. SZB is a Light Water Reactor with aspirations to generate low carbon electricity beyond 2050. To demonstrate safe operation for Long Term Operation (LTO), degradation of primary circuit pressure boundary components must be understood and accounted for when modelling normal/fault scenarios. SZB uses surveillance programmes to track material properties of the RPV. This consists two schemes, one monitoring the effect of high temperature and neutron irradiation whilst the other monitors high temperature effects in isolation. Each scheme consists of capsules containing test specimens which are periodically extracted, tested and analysed to determine whether properties remain within acceptable limits. Detrimental changes in material properties have been observed in both schemes. Whilst property changes are within current safe operation limits, they may challenge LTO aspirations beyond the current planned end of generation (2035). Understanding the thermal aging mechanism(s) is therefore of high importance for LTO. The project will use a combination of advanced microscopy techniques to investigate the microstructure of SZB RPV steel at different thermal ageing and stress conditions. The project will aim to identify the following objectives: The student will undertake advanced microstructural characterisation at the University of Bristol, in particular scanning and transmission electron microscopy, focused ion beam microscopy and atom probe tomography. The student will be trained in the use of these techniques, and learn the skills required to analyse the resulting data. In addition, the student will be trained in the theoretical background and modelling of thermal ageing in low-alloy steels, and the underlying engineering and scientific principals behind water-cooled nuclear reactors. As part of the project, the student will be expected to undertake a three month placement with EDF at its central facilities, with potential to visit the SZB site and learn more about the operation of a nuclear reactor. The studentship will be aligned with the Centre for Doctoral Training in Nuclear Energy Futures, enabling the student to access the training modules, facilities and research visits offered by that programme. The candidate should have an undergraduate degree in materials science, physics, chemistry, engineering or an equivalent discipline. The candidate should have an enthusiasm for nuclear materials and materials science, and be comfortable with both computational modelling and experimental work. The student should be a home student as only UK fees + stipend are covered. It would be beneficial for the student to have an understanding of nuclear energy, the microstructure of metals and/or electron microscopy, but all training in these techniques will be provided. The safe storage and custodianship of spent nuclear fuel is one of the most important tasks at Sellafield. Most AGR UO2 fuels are held in storage ponds with carefully controlled water chemistry, but it is crucial to understand how behaviour is modified way from ideal conditions. If this happens and there is exposure of the fuel to this aqueous environment (following a split fuel pin, due to stress-corrosion-cracking for example), what are the consequences? Left: Plasma, forming U-based thin film surfaces at high T in our dedicated actinide facility. Right: Diamond anvil cell for high pressure synthesis and measurement (gold electrodes are visible on the diamond culet). The key factor that drives oxidation and then potential dissolution in this case is typically the production of radiolytic products due to the high radiation fields. This produces highly oxidising free radicals and hydrogen peroxide, that drive UO2 towards a UO22+ state that is soluble in water. During this oxidation process the UO2 takes on excess oxygen without much deformation, mostly retaining the fluorite structure (some subtleties can be detected with Raman, XRD or EXAFS during this stage for example), until it becomes orthorhombic U3O8, which then oxidises further to UO3. The later stages of oxidation result in a large volume expansion that could exacerbate any structural defects already present in the can. This PhD will investigate the oxidation of UO2 in liquid water, varying pH and chloride levels. The student will carry out synthesis of bulk uranium oxide powders, using our radioactive materials handling laboratory and also use the NNUF Facility for Radioactive Materials Surfaces (FaRMS), to deposit thin films of UO2 with in-situ XPS for chemical characterisation. The new XPS capability will also be used to characterise oxidised UO2 product from later experiments. A series of oxidation studies, using a range of aqueous environments will investigate the morphology, chemistry and crystal structure of the products formed, the rate of product formation, and analyse rates of volume expansion that may have consequences for current AGR fuel storage. The PhD will be part of the Nuclear Energy Futures CDT and will last a duration of four years with cohort activities from the CDT to take place alongside the PhD project. There will be monthly update meetings with the academic/industrial supervision team with more frequent interaction on a more ad-hoc basis with the academic group at the university. The PhD student will be embedded in a large and vibrant nuclear research group at the university of Bristol and there will be opportunities to explore other techniques and scientific avenues as part of this PhD, using the vast array of cutting- edge techniques at our disposal.Determination of Thermal Ageing Mechanism(s) in LWR Primary Circuit Pressure Boundary Material
Funding Notes
The oxidative dissolution of UO2 in aqueous environments
