Registration
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m.j.d.rushtonbangor-untf@gmail.com m.j.d.rushtonbangor-untf@gmail.com

Megan Owen – Welcome Address
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m.j.d.rushtonbangor-untf@gmail.com m.j.d.rushtonbangor-untf@gmail.com

Michael Rushton – Nuclear Futures Presentation
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m.j.d.rushtonbangor-untf@gmail.com m.j.d.rushtonbangor-untf@gmail.com

Michael Rushton – Nuclear Futures Presentation
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m.j.d.rushtonbangor-untf@gmail.com m.j.d.rushtonbangor-untf@gmail.com

Prashant Dwivedi – Study of Crystallization of Amorphous Metals through Molecular Dynamics Simulations
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The outstanding corrosion resistance that may be possible with structurally amorphous metals was recognized several years ago. For example compositions of several iron-based amorphous metals possess very good corrosion resistance and neutron absorption properties. These coatings, with further development, could be cost-effective options to enhance the corrosion resistance of drip shields and waste packages, and limit nuclear criticality in canisters for the transportation, aging, and disposal of spent nuclear fuel. Iron-based amorphous metal formulations with chromium, molybdenum, and tungsten have shown the corrosion resistance believed to be necessary for such applications. Further, the combination of crystalline and amorphous layers represents a promising route for the design of multilayered coatings with improved mechanical properties. Because glassy alloys do not exist in thermodynamic equilibrium, they undergo crystallization with the supply of thermal energy. However, despite decades of experimental and theoretical efforts, many questions regarding the details of these processes remain still open. Due to the small space scales involved, the experimental investigation of the mechanisms underlying the phase formations caused by strain/stress as well as during nanoindentation on amorphous materials is difficult. Fortunately, the fast increase in available computation power is today allowing more and more accurate and larger size molecular dynamics (MD) simulations of almost any kind of system.

James Davidson – Synthesis and Properties of W2B5 for High Energy Neutron Shielding
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Ditungsten pentaboride is a highly effective shielding material for energetic neutrons, making it an attractive candidate for many nuclear reactor applications including the spherical tokamak fusion reactor. However, there is little processing or property data available for the compound. This work investigates the synthesis of W2B5 with a particular focus on the densification kinetics of commercial W2B5 powders via hot pressing. The effect of various sintering conditions, such as sintering time, sintering pressure and sintering atmosphere, amongst other, are reported. Thermophysical and chemical properties of the monolithic material are also reported, with a focus on the stoichiometry of the material, as well as a focus on properties related to thermal stress and thermal fatigue, which will allow the lifetime of these components to be better understood. Finally, a discussion on the development of cermet composites from W2B5 and how the addition of the metallic binder may aid both sintering and the relevant thermophysical properties.

Megan Carter – Development and Characterisation of ODS Nickel-based Alloys for use in Molten Salt Reactors
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Molten salt reactors (MSRs) are a leading generation IV fission reactor design, on account of their heightened operating efficiency and improved safety in comparison to light water reactors (LWRs). Critical to the implementation of this technology is the development of structural materials capable of withstanding the high temperature, severely corrosive, and highly irradiating environment inside a MSR.

The commercially available Hastelloy family of Ni superalloys have already, historically, been proven as promising candidate materials for use in MSRs; they possess superior high temperature mechanical properties. In particular, variants of Hastelloy N (Ni-7Cr-16Mo) are shown to exhibit minimal corrosive attack in hot fluoride salts. However, Ni alloys are vulnerable to irradiation embrittlement, most significantly, that induced by helium segregation. Helium is an unavoidable product of reactor operation, produced through transmutation reactions, and thus there is a drive to develop a Hastelloy variant which is resistant to the embrittling effects of He.

Nano-sized oxide-based particles have been shown to act as effective He traps in oxide-dispersion-strengthened (ODS) steels, preventing premature component failure caused by He embrittlement. Consequently, the possibility of combining the irradiation benefits of ODS particles with the corrosion and mechanical properties of Hastelloy is a promising avenue in MSR material research.

A joint venture by the University of Oxford, North Carolina State University, the University of California-Berkeley, the University of Idaho and Idaho National Laboratory (INL), is aiming to design, test and characterise candidate ODS nickel alloys for potential use in MSRs. Candidate alloys have been fabricated used powder metallurgy, combining Hastelloy with yttria powder, before being subjected to mechanical testing and multi-level characterisation. This talk will present results from initial investigations, with a particular focus on the use of atom probe tomography, to demonstrate the extent to which yttria incorporation has been successful; a critical factor in determining the success of these alloys in trapping He, and thus preventing irradiation-induced embrittlement.

Maciej Makuch – Pourbaix diagram informed phase field model of meta-stable to stable pitting transition for austenitic stainless steel
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The 300 series austenitic stainless steel finds extensive use as structural materials in the nuclear industry wherever high strength and corrosion resistance are needed. The single order parameter phase-field model presented here shows how to utilize polarization potentials obtained from the Pourbaix diagrams to simulate the transition between the fast initial pitting and later stage slow continuous pit growth on the surface of 316L stainless steel. The ionic species concentrations at the metal surface obtained using concentration-dependent electrode polarizations agree well with the concentrations obtained using the constant polarization model. This property significantly speeds up simulation time in the stable pitting regime and thus increases the model's predictive capabilities. The second part of the work presents a multi-phase field model based on extending the Kim-Kim-Suzuki formulation. It allows to predict the behaviour of insoluble corrosion products on the corroding metal surface and gives an example of an easy to implement multi-phase field tool.

Megan Owen – Diffusion in undoped and Cr doped amorphous UO₂
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Insoluble dopants such as chromia, Cr₂O₃, and alumina, Al₂O₃, are often added to UO₂ fuel to enhance fuel properties. Such properties include larger grain sizes and increased fission gas retention. Dopants may segregate towards more favourable regions in the fuel, such as grain boundaries, creating amorphous or disordered regions. Diffusion along potentially disordered grain boundaries may differ to that observed in the crystalline bulk, which has been discussed in previous modelling work on ZrO₂. In this work, we use classical molecular dynamics to simulate amorphous undoped and Cr doped UO₂ systems. A range of temperatures and Cr concentrations have been analysed, and diffusion of all species have been computed. Comparisons have been made between crystalline and amorphous, undoped and doped, counterparts where appropriate. This will allow us to deduce whether diffusion is impacted based on the structure of the system, and how this may relate to potential amorphous grain boundary formation in doped UO₂ fuel pellets.

Kavi Sharma – Determining the Dislocation Density of Nuclear Materials Using CMWP
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In the UK, Advanced Gas-Cooled Reactors (AGRs) use a variation of 316 stainless steel as the fuel cladding due to its high temperature and corrosion resistance. The operating lives of all of the UK AGRs have been extended over recent years. As the older AGR stations reach the point where further extensions are not viable, preparation for removing the fuel and decommissioning is the next step. As a result of these extensions, and necessary lifetime management of irreplaceable core components, the UO2 fuel may have experienced increased burn-up and time in reactor, potentially impacting the cladding end of life properties (and therefore performance) during post-discharge and storage activities. Therefore, it is paramount to investigate and establish any physical and mechanical properties changes, in order to mitigate any potential future handling issues.

The aim of this PhD is to gain a fundamental understanding of the effects of radiation damage on the AGR fuel cladding material, stainless steel, by extrapolating data to predict damage for higher burn up nuclear fuel. The current focus of the project is to use a novel X-ray Diffraction (XRD) technique, Convolutional Multiple Whole Profile (CMWP) analysis, in order to determine the dislocation density and defect type within a material. This is done by analysing diffractograms, specifically parameters such as peak broadening. In order to fully understand the applicability of this technique to AGR material, a metallic zirconium system will first be investigated to establish CMWP fitting parameters. This is because zirconium alloys have been thoroughly investigated with this technique and has proven to work, therefore analysing elemental zirconium, which has less complex diffraction pattern, will be a good foundation.

In this contribution, the analysis of zirconium will be presented. Two investigations will be undertaken, the first being the zirconium as-received (Zr-AR) compared to a Zr-AR that has been heat treated (Zr-AR+HT) in order to anneal and recrystallise the sample. The second investigation was the comparison of Zr-AR+HT against zirconium that has been cold-worked to varying degrees. Future work packages, which include comparison with stainless steel materials and microscopy data will also be described.

Ciara Fox – Moltex Energy Tackling Corrosion in Molten Salt Nuclear Reactors for the Stable Salt Reactor – Uranium (SSR-U)
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Corrosion of metals by molten salts is a long-standing challenge in the nuclear industry for molten salt Generation IV reactor designs. Molten salts strip the oxide layers from metal alloys, exposing a bare metal surface to the salts that contain oxidising species, the unprotected metal surfaces corrode and leach alloying elements into the salts. At Moltex Energy for the SSR-U, corrosion is prevented by stabilisation of the redox potential. The SSR-U is a fluoride salt reactor with separate fuel and coolant salts. The reactor core comprises an array of fuel tubes that are maintained at a low pressure in a graphite matrix, which fills most of the tank. Each tube sits in a separate channel, within which a molten salt primary coolant circulates.

In the low enriched uranium fuel salt, control of the redox potential is achieved by controlling the oxidation states of the uranium salts which act as a redox buffer in a eutectic salt mixture with a sodium fluoride diluent. The redox buffer enables maintenance of redox potential and neutralisation of potentially corrosive fission products generated throughout the reactor’s lifespan. In the primary and secondary coolant salts, the redox potential is controlled by the addition of a proprietary redox control agent that scavenges any oxygen that leaks into the reactor before it can attack the metal.

Gareth Stephens – Identifying the most stable accommodation mechanisms for Li in ZrO2 fuel cladding
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With efforts being made to prolong the burnup of nuclear fuels and increase efficiency of pressurised water reactors (PWRs), there is a focus on extended residence times of fuel within reactors. In addition, there are potentially significant cost benefits through plant simplification if a soluble boron-free lithiated primary water chemistry can be demonstrated to be a viable route for PWR-based small modular reactor operation. However, the corrosion behaviour of the zirconium alloy clad material under lithiated conditions remains a concern as the mechanisms that underpin this have yet to be fully identified. The mechanism by which Li accelerates zirconium alloy corrosion will allow new alloying additions to be considered and new water chemistry regimes to be investigated, improving the efficiency and performance of future nuclear power reactors.

Density functional theory (DFT) and targeted experiments were used to identify the most stable accommodation mechanisms for Li in ZrO2. Atomic scale modelling was used to produce Brouwer diagrams that predict the nature of the defect structures and their competing species concentrations as Li is accommodated in ZrO2. This was then combined with experimental data to corroborate the predictions.

The solubility of Li in bulk ZrO2 is predicted to be low, however, solubility in amorphous structures has been found which comprise complex grain boundaries. An increase of Li results in an increase in vacant oxygen defects which may aid transport of oxygen to the metal surface, accelerating corrosion. Further experiments using differential scanning calorimetry have shown that in Li stabilises the amorphous structure to higher temperatures before crystallisation compared to the undoped amorphous ZrO2 samples.

Susannah Lea – Peridynamic Modelling Method of the Corrosion of Zirconium Alloy
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Peridynamics is a non-local continuum mechanics modelling technique that can capture discontinuities, in particular cracking, where discrete finite element methods often cannot. Cracks can evolve naturally within the model, without an initiation point, making it ideally suited to modelling cracking in the oxide on zirconium alloys and naturally capturing phenomena such as buckling of the oxide and the oxide-metal interface. This can provide insight into the mechanisms driving the formation of predominantly lateral cracks seen periodically through the oxide thickness as it grows. These cracks occur in relation to transitions, where the passivating nature of the oxide layer breaks down and allows oxygen transport to the metal surface before the passivating oxide forms again.

This work uses bond-based peridynamics implemented in the finite element software Abaqus, where truss elements represent forces carried between material points (nodes) and is a study into the various model parameters that affect the development of the oxidation front and its progression. This includes the thickness of the transition region, the amount of metal initially oxidised and the rate the front moves through. Furthermore, the effect of boundary conditions and constraints are investigated in an attempt to capture oxidation behaviour observed experimentally and to understand the mechanisms behind it.

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

Phil Monks – Careers Presentation
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m.j.d.rushtonbangor-untf@gmail.com m.j.d.rushtonbangor-untf@gmail.com

Nick Smith – Careers Presentation
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m.j.d.rushtonbangor-untf@gmail.com m.j.d.rushtonbangor-untf@gmail.com

Janet Wilson – Careers Presentation
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m.j.d.rushtonbangor-untf@gmail.com m.j.d.rushtonbangor-untf@gmail.com

Careers, Networking, Drinks & Posters Session
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m.j.d.rushtonbangor-untf@gmail.com m.j.d.rushtonbangor-untf@gmail.com