Crystal structure, thermodynamics, magnetics and disorder properties of Be-Fe-Al intermetallics

The elastic and magnetic properties, thermodynamical stability, deviation from stoichiometry and order/disorder transformations of phases that are relevant to Be alloys were investigated using density functional theory simulations coupled with phonon density of states calculations to capture temperature effects. A novel structure and composition were identified for the Be-Fe binary {\epsilon} phase. In absence of Al, FeBe_5 is predicted to form at equilibrium above ~ 1250 K, while the {\epsilon} phase is stable only below ~ 1650 K, and FeBe_2 is stable at all temperatures below melting. Small additions of Al are found to stabilise FeBe_5 over FeBe_2 and {\epsilon}, while at high Al content, AlFeBe_4 is predicted to form. Deviations from stoichiometric compositions are also considered and found to be important in the case of FeBe_5 and {\epsilon}. The propensity for disordered vs ordered structures is also important for AlFeBe_4 (which exhibits complete Al-Fe disordered at all temperatures) and FeBe_5 (which exhibits an order-disorder transition at ~ 950 K).Comment: 14 pages, 10 figures, accepted for publication in J. Alloy Compd. on 14 March 201

The effect of cluster reconfiguration and non-stoichiometry on uranium vacancy migration in UO2

During reactor operation the release of fission gases from the fuel pellet is an important safety issue as it can lead to over-pressurization and failure of the fuel cladding. Uranium vacancy migration has been identified as the limiting step in the diffusion of fission gases through bulk UO2. The uranium vacancy migration energy is, therefore, an important parameter in this phenomenon, as well as other atomic scale processes, such as recovery from radiation damage. Chemical changes under taken by the fuel during irradiation lead to deviations from stoichiometric UO2 and the charge compensating defects that bind to the uranium vacancy also change. Therefore, we have examined the change in the migration energy for a uranium vacancy when bound to either two oxygen vacancies (Schottky defect) or to four U5+ cations (hole defects) representing UO2 and UO2+x respectively. By using empirical potentials within statics we were able to sample a large array of metastable cluster configurations to identify lower energy migration pathways that involve the reconfiguration of the cluster from the ground state configuration to metastable configurations (see UO2+x results in Figure 1). The work is published in ref [1]. Please click Additional Files below to see the full abstract

Non-stoichiometry in monoclinic zirconia and amorphous zirconia

A combination of materials modelling techniques and targeted experimental investigations have identified the manner in which non-stoichiometry is accommodated within both crystalline and amorphous ZrO2. Not only is excess oxygen possible in both crystalline and amorphous ZrO2, but it is found that there is a high propensity for significant deviations – especially in the amorphous system – forming ZrO2+x. This has clear implications to the behavior and degradation of ZrO2 as a thermal barrier coating in aerospace and energy components, but also as the boundary oxide protecting zirconium alloys in aggressive environments, including within a water cooled nuclear power reactor. The behavior was highlighted through a combination of both Raman spectroscopy and associated atomic scale predictions coupled with thermodynamic analysis of the system. As excess oxygen cannot readily oxidize Zr4+ ions beyond this charge state, the additional oxygen is accommodated instead as a peroxide ion – O22-. This peroxide specie has a distinct covalent bond not expected in the stoichiometric ionic ZrO2 system that is readily observable using Raman spectroscopy. Now that excess oxygen accommodation in ZrO2 has been highlighted, an understanding of how various dopant or alloying elements can impact its behavior can be targeted to improve component reliability. The presence of amorphous phases at grain boundaries is also discussed in terms of potential super-highways for oxygen transport through the oxide system

Computational Studies of Grain Boundary Behavior in Uranium Dioxide Nuclear Fuels

Nuclear power is responsible for the production of 380,000 Megawatts of energy worldwide, which results in over 11% of the world’s energy production [world-nuclear.org]. Pellet-cladding interactions (PCI) are a key nuclear fuel failure mechanism which presents formidable challenges to researchers due to extreme nuclear fission conditions. Although PCI interactions have been reduced due to fuel additives, understandings of PCI interactions remain elusive. We propose new approaches to increase understanding of nuclear fuel interactions; specifically, uranium dioxide and the effects of dopants. This study focuses on amorphous uranium dioxide and fission products, while benchmarking new methods with previous computational studies. Results will further research into modeling approaches and help guide future experimental design

Stoichiometry deviation in amorphous zirconium dioxide

Amorphous zirconia (a-ZrO(2)) has been simulated using a synergistic combination of state-of-the-art methods: employing reverse Monte-Carlo, molecular dynamics and density functional theory together. This combination has enabled the complex chemistry of the amorphous system to be efficiently investigated. Notably, the a-ZrO(2) system was observed to accommodate excess oxygen readily – through the formation of neutral peroxide (O(2)(2−)) defects – a result that has implications not only in the a-ZrO(2) system, but also in other systems employing network formers, intermediates and modifiers. The structure of the a-ZrO(2) system was also determined to have edge-sharing characteristics similar to structures reported in the amorphous TeO(2) system and other chalcogenide-containing glasses

Coated ZrN sphere-UO2 composites as surrogates for UN-UO2 accident tolerant fuels

Uranium nitride (UN) spheres embedded in uranium dioxide (UO2) matrix is considered an innovative accident tolerant fuel (ATF). However, the interaction between UN and UO2 restricts the applicability of such composite in light water reactors. A possibility to limit this interaction is to separate the two materials with a diffusion barrier that has a high melting point, high thermal conductivity, and reasonably low neutron cross-section. Recent density functional theory calculations and experimental results on interface interactions in UN-X-UO2 systems (X = V, Nb, Ta, Cr, Mo, W) concluded that Mo and W are promising coating candidates. In this work, we develop and study different methods of coating ZrN spheres, used as a surrogate material for UN spheres: first, using Mo or W nanopowders (wet and binder); and second, using chemical vapour deposition (CVD) of W. ZrN-UO2 composites containing 15 wt% of coated ZrN spheres were consolidated by spark plasma sintering (1773 K, 80 MPa) and characterised by SEM/FIB-EDS and EBSD. The results show dense Mo and W layers without interaction with UO2. Wet and binder Mo methods provided coating layers of about 20 mu m and 65 mu m, respectively, while the binder and CVD of W methods layers of about 12 mu m and 3 mu m, respectively

High temperature, low neutron cross-section highentropy alloys in the Nb-Ti-V-Zr system

High-entropy alloys (HEAs) with high melting points and low thermal neutron cross-section are promising new cladding materials for generation III+ and IV power reactors. In this study a recently developed high throughput computational screening tool Alloy Search and Predict (ASAP) has been used to identify the most likely candidate single-phase HEAs with low thermal neutron cross-section, from over a million four-element equimolar combinations. The selected NbTiVZr HEA was further studied by density functional theory (DFT) for moduli and lattice parameter, and by CALPHAD to predict phase formation with temperature. HEAs of NbTiVZrx (x = 0.5, 1, 2) were produced experimentally, with Zr varied as the dominant cross-section modifier. Contrary to previous experimental work, these HEAs were demonstrated to constitute a single-phase HEA system; a result obtained using a faster cooling rate following annealing at 1200 °C. However, the beta (BCC) matrix decomposed following aging at 700 °C, into a combination of nano-scale beta, alpha (HCP) and C15 Laves phases

Uranium nitride-silicide advanced nuclear fuel: Higher efficiency and greater safety

The development of new nuclear fuel compositions is being driven by an interest in improving efficiency/lowering cost and increasing safety margins. Nuclear fuel efficiency is in large measure a function of the atomic density of the uranium, that is, the more fissionable uranium available per unit volume the less fuel volume that is required. Proliferation concerns limit the concentration of fissile 235U, and thus attention is directed to higher overall uranium content fuel. Among the options are the high temperature phases U3Si2 and composite UN- U3Si2 where the design would have the more water-stable U3Si2 surround the more soluble, but higher uranium density UN grains. (Uranium metal of course has the highest atomic density, however its low melting point, high degree of swelling under irradiation, and chemical reactivity eliminate it from consideration.) Another advantage of the nitride and silicide phases are their high thermal conductivity, greatly exceeding the current standard UO2 fuel, with the high conductivity potentially allowing the fuel to operate at a higher power density. Please click Additional Files below to see the full abstract