Thomas Bainbridge – The Drying of Spent Advanced Gas-Cooled Reactor Fuel
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In 2012 the Nuclear Decommissioning Authority took the decision to cease the reprocessing of spent advanced gas-cooled reactor fuel with any remaining fuel and future arisings to be stored pending the final decision on whether to classify it as waste for disposal. Should the spent fuel be classified as waste it will be stored until the high heat generating portion of the UK’s geological disposal facility (GDF) is ready – this is expected to be in 2075.

Currently the strategy employed is long term wet storage with work being undertaken to support the alternative of dry storage with the work also being applicable for preparing for disposal. The aim of drying the spent fuel is to prevent the radiolysis of any water left in or on the pin which could lead to hydrogen and/or hydrogen peroxide being produced. These two radiolysis products are of particular concern due to hydrogen being explosive and hydrogen peroxide being corrosive. As the primary concern with extended wet storage is that water could seep through the cladding if it has failed. Stress corrosion cracking is being investigated as the likely cause of these failures which are often highly tortuous and branched cracks making characterisation difficult.

The work being carried out as part of this PhD aims to inform the drying process by producing a process model with this being split into two aspects. The experimental side is aiming to produce representative cracks in stainless steel in order to conduct drying trials which, in turn, can be used to validate the computational side of the work. On the computational side, a code has been written to determine the length and width of a crack which is required in order to produce an accurate model of the drying process. The current progress towards these goals will be discussed here.

Emma Perry – The effect of thorium on the dissolution of mixed uranium and thorium dioxide fuels under deep geological repository conditions
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This presentation will outline the results of recent batch dissolution experiments of UO2, U0.5Th0.5O2, U0.7 Th0.3O2 and ThO2 pellets, in synthetic ground waters doped with hydrogen peroxide under anoxic conditions. The oxide powders have been fabricated by oxalic coprecipitation and thermal decomposition then pressed into 5mm pellets and sintered at 1600 degrees.

Mixed oxide fuel containing uranium and plutonium dioxide needs a disposal plan. Deep geological disposal is internationally recognised as the safest long-term solution. However, it is expected that eventually groundwater will break through the multiple barriers of a deep geological repository and contact the fuel. So, it is important to understand how mixed oxide fuel dissolution may differ from uranium-based spent nuclear fuels. Experiments with plutonium can only be conducted on licensed sites with stringent safety protocols. These experiments are limited in number, so it is necessary to perform complementary experiments on lower activity model fuels. Uranium-thorium dioxide has been selected to model Pu(IV) homogenously distributed in the uranium dioxide matrix.

The main dissolution mechanism of the uranium matrix in groundwater solutions involves the oxidation of U(IV) to the more soluble U(VI) and the subsequent removal of U(VI) by carbonate ions. There is very little oxidation in the anoxic conditions underground, particularly considering the reductants introduced by the dissolution of steel canisters. However, radiolytic oxidants will be produced by the interaction of radiation from the spent nuclear fuel and water. These radiolytic oxidants are produced close to the surface of the fuel and contribute to the oxidative dissolution of uranium dioxide. The dominant radiolytically produced oxidant is hydrogen peroxide [1]. In the presented batch experiments, hydrogen peroxide was added to the solutions initially. The consumption of hydrogen peroxide was observed in samples taken throughout the study using UV/vis spectroscopy. The uranium concentrations in the solution were followed using ICP-MS as a measure of the dissolution rate. The dissolution yield was calculated by subtracting the peroxide consumption in background effects unrelated to the pellet. The dissolution yield is independent of surface area and allows comparison to (U, Pu)O2 dissolution data.

Mixed oxide fuels are more radioactive and so will have a higher radiolytic production rate than uranium-based spent nuclear fuels. There are concerns that this will accelerate the dissolution, but these concerns are countered by evidence that the chemistry of plutonium mitigates these effects [2]. This mitigation depends on enhanced autocatalytic decomposition of the peroxide and the formation of a ‘protective’, less soluble plutonium-rich surface on the altered mixed oxide fuel as uranium is released. No thorium has been detected in the prewash, dissolution or rinse solutions of the presented batch dissolution experiments. Therefore, hydrogen peroxide consumption in the ThO2 dissolution study sets the baseline for the autocatalytic decomposition rate.

Jack Rolfe – Disposal MOX
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The UK has a substantial inventory of separated plutonium from its historical reprocessing of spent nuclear fuel. The Government’s preferred option for this is re-use as a Mixed uranium-plutonium OXide (MOX) fuel. However, 5% is not suitable for re-use and is recommended for disposal in a Geological Disposal Facility (GDF) after having first been immobilised in a ceramic wasteform. The NDA is evaluating processes for plutonium immobilisation. One option being considered is ‘Disposal MOX’ which is the manufacture of regular MOX but intended for disposal to GDF not irradiation. Regular MOX is manufactured by heterogeneous blending of uranium and plutonium feed powders and then forming fuel pellets through a technologically-mature cold-press and sinter process, yielding a heterogenous MOX product. For Disposal MOX production this may be modified by increasing the plutonium loading and introducing neutron poisons into the material for criticality safety/safeguards purposes. Alternatively, advanced manufacturing technologies such as Flash and Spark Plasma Sintering may provide routes to higher quality, denser Disposal MOX. Thus, using plutonium simulants, this project will explore using modifications to existing manufacturing routes and advanced manufacturing technologies described above in the development of a Disposal MOX, underpinning HMG’s policy on the direct disposal of plutonium.

Reece Bedford – Accommodation of Nitrogen and Chlorine in Plutonium Dioxide During Storage
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The UK possess one of the world’s largest civilian stockpiles of separated plutonium, the majority of which is stored at the Sellafield site in Cumbria. The nature of the stored plutonium can vary, but in general, the plutonium is stored as an oxide powder within a metal container, bagged out in a plastic bag, then welded shut into an outer metal container. Despite this process taking place under an argon atmosphere, species such as water, oxygen and nitrogen from the air are likely to be entrained within the packages, as well as chlorine due to the degradation of the plastic . In the near-future, this plutonium will require removal from these storage containers to either be processed into mixed-oxide (MOx) fuels or into a suitable wasteform for long-term disposal. It is, therefore, imperative that we understand how these species may interact with the material and the headspace above it and what influences they may have on the plutonium’s chemistry over long time scales in storage. This work looks to study the effects that nitrogen and chlorine have on the plutonium's defect chemistry using density functional theory.

Coffee Break
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Rachel Crawford – Investigating the Influence of Iron on the Chemical Durability of Vitrified Wasteforms
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The implementation of an engineered multi-barrier approach for nuclear waste disposal, to mitigate the release of radionuclides over the operational lifetime of a geological disposal facility (GDF), requires a detailed understanding of the interactions between steel canisters, engineered backfill and natural barriers. Accordingly, the influence of Fe species on the long-term durability of vitrified high-level waste (HLW) is an important consideration in the safety case for a GDF. The presence of Fe in many UK-specific HLW feedstocks allows for Fe-glass interactions to occur as a result of both Fe from within the glass network, and Fe from the environment. Here, short- and long-term laboratory-based studies are combined with observations from glasses exposed to complex natural environments to understand the effect of Fe on glass dissolution under a variety of geochemical conditions relevant to geological disposal.

Ashley Smith – Effects of U-content and texture on the properties of (U,Zr) alloy research reactor fuel
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Some nuclear reactors such as the Experimental Breeder Reactor II (EBR II) use Zr-rich Zr-U alloys as their nuclear fuel. Zr exhibits anisotropic behaviour due its HCP structure. The materials response to creep and hydride cracking are two examples of this anisotropy that can affect reactor operations. How the crystallites are aligned in a material (texture) will affect how the material responds to reactor conditions. This work looks to investigate how the composition of Zr-U alloys (and surrogates) relates to the texture produced through similar manufacturing methods.

Max Cole – Demonstration of the Disposal MOX Concept for Immobilisation of Plutonium
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Disposal mixed oxide (d-MOX) fuels are a candidate wasteform for the immobilisation of Pu. d-MOX fuels are fabricated by diluting PuO2 within a UO2 matrix and adding a neutron poison, such as Gd. The resulting wasteform will ideally adopt the cubic fluorite structure (Fm-3m) of the matrix, allowing it to retain high corrosion resistance of UO2, whilst Gd mitigates criticality in the disposal environment.

In the present work, a series of four simulant d-MOX compositions were prepared using both solid-state and wet co-precipitation fabrication routes. Th, acting as a Pu surrogate, and Gd were doped into UO2 to produce sintered pellets of simulant d-MOX. Each composition was characterised using XRD, SEM-EDS, Raman, XANES and ICP-OES, revealing the effects of composition and fabrication route on the phase formation, lattice parameter, U oxidation state and grain size.

Both solid-state and co-precipitation synthesis routes yielded dense pellets of simulant disposal-MOX with grain sizes comparable to PuO2-UO2 MOX fuel. Gd3+ and Th4+ were incorporated within the fluorite (Fm-3m) UO2 matrix, substituting directly for U4+ on the FCC cation sub-lattice. Trivalent Gd3+ was charge compensated through both the generation of oxygen vacancy defects, as observed by Raman spectroscopy analysis, and oxidation of U4+ to U5+, as determined by high energy resolution fluorescence detection x-ray absorption near edge spectroscopy (HERFD-XANES). These data demonstrate that, in principle, disposal-MOX materials are a viable candidate wasteform for Pu immobilisation.

Lunch
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Steven McGowan – Capture of uranium and other economic minerals from seawater using crop bioresidues
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Recovery of naturally occurring uranium from seawater would mean the nuclear cycle would fulfil the definition of a sustainable renewable power source, but the current best available technology is uneconomic compared with geological sourcing. However, use of naturally occurring materials to recover the uranium would offer the opportunity to significantly reduce the associated costs, if suitable materials can be identified.

Patrick Mfunyi – A Biomimetic Radiometric System for Off-shore Radioactive Particle Surveillance
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The major source of anthropogenic radionuclides in the seabed sediment in the UK can be traced to three main sources as follow: the global and local fallout due to nuclear weapon testing, the nuclear reprocessing plant at Sellafield and the fallout from accident at Chernobyl without omitting secondary main sources such as the nuclear power stations, the fuel production facilities, nuclear research facilities and dumping of low-level radioactive waste.

This project looks at the design and testing of an innovative underwater system with the intent that it will dig unaided into the sediment. A variety of design models have been produced but only two of them have been retained because of their capabilities to detect low-level, low-energy gamma ray radioactivity present under the seabed or ocean floor, autonomously. The materials to be detected are classified as alpha-rich such as americium-241 and related actinides, and beta-rich such as fission product like caesium-137 and higher energy radiation like cobalt-60. However, the complication in the design model arises from the detection of americium-241 and the actinides because they are relatively difficult to detect due to the shielding effect of the sediment. Amercium-241 arises from the beta decay of plutomium-241 and so, by detecting americium-241 under the seabed or ocean floor, plutonium abundance can also be inferred.

The seabed sediment, where the anthropogenic insoluble radioactive particles are located, is made of sand, silt or clays. The radioactive particles, to be detected by the designed and built system, are known as a localised aggregation of radioactive atoms that gives rise to inhomogeneous distribution of radionuclides significantly different from that of matrix background. The key important points in this project is the underwater system discovery and characterisation of radioactive particles. Series of analytical tools are available for identification, isolation and characterisation. Radioactive particles can be characterised by using as a minimum gamma spectrometry which determine the isotope composition of radionuclides. However, non-destructive solid-state characterisation, such as the SEM-EDX, is the best method to adopt for this project because it provides size distribution, surface morphology, elemental composition and distribution. For specific low-level, low energy gamma ray, a non-destructive technique known as LEPS can be used for activities and isotopic ratio of americium-241, plutonium-239 and plutonium-240.

Nevertheless, several digging devices have been reported in the research often inspired by how aquatic animals dig and drill on the seabed or ocean floor such as clams or worms as well as digging for the planting of trees and adapted screw burying themselves. It is inspiring to observe the performance of these designed underwater systems especially when these radiometric systems are fully submerged through their digging process.

This project main objective is to build an autonomous underwater system that can dig itself into the seabed for the purpose of low-level, low-energy radioactivity characterisation. This has required the development of two innovative design models that would allow crossing to a desired underwater location, drilling the seabed or ocean floor and remaining stable while buried with option to reposition to different location.

Harvey Plows – Wire-Mesh and Optical Fibre sensing for Thermal Hydraulic Measurements
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The study of thermal hydraulic phenomena is essential to support the design and safe operation of nuclear power stations. To help tackle the thermal hydraulic research challenges facing the nuclear industry, Bangor University’s Nuclear Futures Institute has constructed the Thermal Hydraulics Open-access Research (THOR) facility at Anglesey’s Menai Science Park. This is a highly modular water thermal hydraulic loop designed to support a wide range of industrial and academic projects. The work that will be conducted during this PhD will focus primarily on developing new instrumentation techniques, initially using wire-mesh sensor (WMS) and fibre Bragg grating (FBG) sensor principles, to design improved instruments for the measurement of the properties of multiphase flow. These instruments will be installed within the THOR facility to conduct experiments to assess their accuracy, precision and reliability. In particular, novel FBG sensor designs will be investigated due to their durability, geometric flexibility and high data capture rates. Computational fluid dynamics and system simulations, using OpenFOAM and RELAP5 respectively, are also being developed to assist in the design of sensor benchmarking experiments.

Enrique Casañas – Low-temperature fabrication of ceramic tritium breeder materials, for enhanced control of microstructure and phase formation
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Lithium-containing ceramics are one of the tritium breeding concepts to be tested in ITER. Due to the specific needs of a breeder blanket and the wide range of properties offered by the materials considered, no single lithium ceramic has been singled out as the best compromise between physical properties and lithium density. Two main ceramics have been identified as potential lithium sources, however: lithium metatitanate (Li2TiO3) and lithium orthosilicate (Li4SiO4). Conventional high temperature fabrication techniques can lead to lithium loss due to volatilisation, decreasing the lithium density in the final product. Here we apply low temperature fabrication techniques to the formation of dense, lithium-containing ceramics, including the newly developed reactive cold sintering (RCS) method. This novel fabrication method enables the formation of dense pellets at reduced temperatures, addressing lithium volatilisation concerns, and also allows the incorporation of chemically and/or morphologically distinct phases, enhancing control over the composition and microstructure. This has ultimately enabled the formation of composite ceramics more specifically tailored to the needs of a breeder blanket. In this contribution we demonstrate the flexibility of the RCS method which was used to produce dense Li2TiO3 with bi-modal grain size distribution, and a dense Li2TiO3-Li4SiO4 composite material. As well as microstructural data, thermal and mechanical data will be presented and compared to ceramics produced via more conventional techniques. Our results suggest that reactive cold sintering could provide a pathway to produce lithium ceramics with little concern for volatilisation as well as a simpler way to introduce enriching phases and allow for retention of specific morphologies from initial powder to final pellet.

Huw Jones – A Surrogate Machine Learning Model for Advanced Gas-cooled Reactor Graphite Core
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A surrogate machine learning model was developed with the aim of predicting seismic graphite core displacements from crack configurations for the advanced gas-cooled reactor. The model was trained on a dataset generated by a software package which simulates the behaviour of the graphite core during a severe earthquake. Several machine learning techniques, such as the use of convolutional neural networks, were identified as highly applicable to this particular problem. Through the development of the model, several observations and insights were garnered which may be of interest from a graphite core analysis and safety perspective. The best performing model was capable of making 95% of test set predictions within a 20 percentage point margin of the ground truth.

Coffee Break
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m.j.d.rushtonbangor-untf@gmail.com m.j.d.rushtonbangor-untf@gmail.com

Networking
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Free Time
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Grand Dinner – Bar Uno
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m.j.d.rushtonbangor-untf@gmail.com m.j.d.rushtonbangor-untf@gmail.com