Publications
493 results found
Fossati PCM, Mellan TA, Kuganathan N, et al., 2021, Atomistic modeling approach to the thermodynamics of sodium silicate glasses, JOURNAL OF THE AMERICAN CERAMIC SOCIETY, Vol: 104, Pages: 1331-1344, ISSN: 0002-7820
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- Citations: 2
Middleburgh SC, Lee WE, Rushton MJD, 2021, Structure and properties of amorphous uranium dioxide, ACTA MATERIALIA, Vol: 202, Pages: 366-375, ISSN: 1359-6454
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- Citations: 12
Humphry-Baker SA, Ramanujam P, Smith GDW, et al., 2020, Ablation resistance of tungsten carbide cermets under extreme conditions, International Journal of Refractory Metals and Hard Materials, Vol: 93, ISSN: 0263-4368
A cobalt-free tungsten carbide cermet (WC-FeNi) has been subjected to oxyacetylene flame tests to simulate extreme operating conditions such as a worst-case fusion reactor accident. In such an accident, air-ingress to the reactor may impinge on components operating at surface temperatures in excess of 1000 °C, leading to tungsten oxide formation and its subsequent hazardous volatilisation. Here, the most challenging accident stage has been simulated, where the initial air-ingress could lead to extremely rapid air-flow rates. These conditions were simulated using an oxidising oxyacetylene flame. The separation between flame nozzle and sample was varied to permit peak surface temperatures of ~950–1400 °C. When the peak temperature was below 1300 °C, the cermet gained mass due to the dominance of oxide scale formation. Above 1300 °C, the samples transitioned into a mass loss regime. The mass loss regime was dominated by ablation of the scale rather than its volatilisation, which was confirmed by performing a systematic thermogravimetric kinetic analysis. The result was unexpected as in other candidate shielding materials, e.g. metallic tungsten, volatilisation is considered the primary dispersion mechanism. The unusual behaviour of the cermet scale is explained by its relatively low melting point and by the lower volatility of its FeWO4 scale compared to tungsten's WO3 scale. The substantially lower volatility of the WC cermet scale compared to metallic W indicates it may have a superior accident tolerance.
Roumiguier L, Antou G, Pradeilles N, et al., 2020, Effect of powder purification by heat-treatment on the creep behaviour of dense boron carbide monoliths, JOURNAL OF THE EUROPEAN CERAMIC SOCIETY, Vol: 40, Pages: 2253-2259, ISSN: 0955-2219
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- Citations: 1
Gasparrini C, Rana D-S, Le Brun N, et al., 2020, On the stoichiometry of zirconium carbide, Scientific Reports, Vol: 10, ISSN: 2045-2322
The dependencies of the enhanced thermomechanical properties of zirconium carbide (ZrCx) with sample purity and stoichiometry are still not understood due to discrepancies in the literature. Multiple researchers have recently reported a linear relation between the carbon to zirconium atomic ratio (C/Zr) and the lattice parameter, in contrast with a more established relationship that suggests that the lattice parameter value attains a maximum value at a C/Zr ~ 0.83. In this study, the relationship between C/Zr atomic ratio and the lattice parameter is critically assessed: it is found that recent studies reporting the thermophysical properties of ZrCx have unintentionally produced and characterised samples containing zirconium oxycarbide. To avoid such erroneous characterization of ZrCx thermophysical properties in the future, we propose a method for the accurate measurement of the stoichiometry of ZrCx using three independent experimental techniques, namely: elemental analysis, thermogravimetric analysis and nuclear magnetic resonance spectroscopy. Although a large scatter in the results (ΔC/Zr = 0.07) from these different techniques was found when used independently, when combining the techniques together consistent values of x in ZrCx were obtained.
Rana D-SBK, Solvas EZ, Lee WE, et al., 2020, An investigation of the long-range and local structure of sub-stoichiometric zirconium carbide sintered at different temperatures, SCIENTIFIC REPORTS, Vol: 10, ISSN: 2045-2322
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- Citations: 7
Bowden D, Ward J, Middleburgh S, et al., 2020, The stability of irradiation-induced defects in Zr<sub>3</sub>AlC<sub>2</sub>, Nb<sub>4</sub>AlC<sub>3</sub> and (Zr<sub>0.5</sub>,Ti<sub>0.5</sub>)<sub>3</sub>AlC<sub>2</sub> MAX phase-based ceramics, ACTA MATERIALIA, Vol: 183, Pages: 24-35, ISSN: 1359-6454
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- Citations: 24
Middleburgh SC, Ipatova I, Evitts LJ, et al., 2020, Evidence of excess oxygen accommodation in yttria partially-stabilized zirconia, Scripta Materialia, Vol: 175, Pages: 7-10, ISSN: 1359-6462
Yttria partially stabilized zirconia (ZrO2)x(Y2O3)½-x has been investigated to understand accommodation of excess oxygen into its structure. ZrO2 powder with 8 wt% Y2O3 additions was treated in 30 vol% H2O2 solution to promote oxidation of the material. A new Raman peak was observed after treatment at 840 cm−1, consistent with previous reports of solid state peroxide ions (O22−). This was corroborated using atomic scale simulation based on density functional theory; these also highlighted the near-zero solution enthalpy for excess oxygen in the monoclinic structure via the formation of a peroxide ion defect.
Dahlfors M, Joannou J, Brummitt A, et al., 2020, The UK national thermal-hydraulics facility: Motivations, design and planning status, Pages: 2876-2886
A UK National Thermal-Hydraulics Facility (NTHF) dedicated to supporting new reactor and other relevant business is being developed, one of the purposes being to deliver on the government's carbon emission reduction commitments. The facility site is foreseen to be at Menai Science Park on the isle of Anglesey in North Wales, a region expected to see significant low carbon energy deployment in coming years. The UK NTHF is envisioned to cater for the needs of emerging nuclear in the UK - but also to serve as a hardware platform for international thermal-hydraulics research collaboration. Plans are to construct a platform capable of maintaining several test loops including support for the UK's on-going, conventional nuclear new build programme as well as Gen-IV systems and associated materials like molten salt and liquid metal coolant media. Motivations are given for NTHF expected capabilities and requirements, which form the basis for its current design and planning state.
Evitts LJ, Middleburgh SC, Kardoulaki E, et al., 2020, Influence of boron isotope ratio on the thermal conductivity of uranium diboride (UB<sub>2</sub>) and zirconium diboride (ZrB<sub>2</sub>), JOURNAL OF NUCLEAR MATERIALS, Vol: 528, ISSN: 0022-3115
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- Citations: 15
Hsieh Y-H, Rushton MJD, Fossati PCM, et al., 2020, Thermal footprint of a geological disposal facility containing EURO-GANEX wasteforms, PROGRESS IN NUCLEAR ENERGY, Vol: 118, ISSN: 0149-1970
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- Citations: 5
Juthapakdeeprasert J, Diaz OG, Lerdprom W, et al., 2019, Reactions and emissivity of cerium oxide with phosphate binder coating on basic refractory brick, INTERNATIONAL JOURNAL OF APPLIED CERAMIC TECHNOLOGY, Vol: 17, Pages: 668-676, ISSN: 1546-542X
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- Citations: 5
Al Nasiri N, Patra N, Pezoldt M, et al., 2019, Investigation of a single-layer EBC deposited on SiC/SiC CMCs: Processing and corrosion behaviour in high-temperature steam, Journal of the European Ceramic Society, Vol: 39, Pages: 2703-2711, ISSN: 0955-2219
Two rare earth monosilicates (Yb 2 SiO 5 , and Lu 2 SiO 5 ) were deposited using a low-cost coating application method to produce a single-layer coating. RE- oxides slurries were dip coated on oxidised CMC samples and subsequently heat treated at high temperature to ensure reaction between SiO 2 and RE-oxides to form the RE-monosilicate. A single, continuous, homogeneous thick coating of 25 μm was obtained. X-ray diffraction (XRD) confirmed formation of, RE-monosilicates in Yb and Lu silicate systems. Coated samples were exposed to 90% H 2 O static steam environment at 1350 °C for 25, 50, 100, and 150 h. Scanning Electron Microscopy (SEM) indicated that both coatings adhered strongly to the substrate. Coating thickness reduced from 22 μm to 11 μm for Yb-coating and 13 to 4 μm for Lu-coating with increasing corrosion time from 25 to 150 h, however there was no significant attack of the CMC for all steam exposure times.
Rushton MJD, Ipatova I, Evitts LJ, et al., 2019, Stoichiometry deviation in amorphous zirconium dioxide, RSC ADVANCES, Vol: 9, Pages: 16320-16327, ISSN: 2046-2069
Zhang H, Jayaseelan DD, Bogomol I, et al., 2019, A novel microstructural design to improve the oxidation resistance of ZrB2-SiC ultra-high temperature ceramics (UHTCs), JOURNAL OF ALLOYS AND COMPOUNDS, Vol: 785, Pages: 958-964, ISSN: 0925-8388
Gasparrini C, Podor R, Fiquet O, et al., 2019, Uranium carbide oxidation from 873 K to 1173 K, Corrosion Science, Vol: 151, Pages: 44-56, ISSN: 0010-938X
Oxidation of UC was studied from 873 to 1173 K in air and in 10 Pa oxygen using a High Temperature Environmental SEM (HT-ESEM). Conversion to U 3 O 8 improved when using 873 K as the oxide product was a fine powder. At higher temperatures (973 K to 1173 K) oxidation slowed due to a densification process with formation of coarse fragments. The oxide fragmentation at 973 K and 1073 K and oxide pulverisation at 873 K were observed in situ in a HT-ESEM. Cracking induced fragmentation and pulverisation was linked to stresses generated from the volumetric transformation from UC to U 3 O 8 .
Middleburgh SC, Lee WE, Rushton MJD, 2019, Ceramics in the nuclear fuel cycle, Advanced Ceramics for Energy Conversion and Storage, Pages: 63-87, ISBN: 9780081027264
Ceramics have played a vital role in the rise of nuclear power since its inception in the mid-20th century. Their development and improved understanding are still driving improvements in the safety, efficiency, and reliability of nuclear power (where it remains the leading source of low carbon energy) and these are having beneficial knock-on effects related to nuclear power’s economic benefits. In this chapter, we discuss the leading fuel concepts starting with uranium dioxide (discussing its manufacture and operation), as well as considering the potential advanced fuels including the so-called accident tolerant fuel candidate materials. Ceramics used to improve fuel cycle costs (e.g., burnable absorbers), ceramics formed or applied as protective barriers reducing corrosion and moderating ceramics are also discussed. Finally, a section on nuclear waste management is provided highlighting the pivotal role of ceramic and glassy waste forms.
Giorgi E, Grasso S, Zapata-Solvas E, et al., 2018, Reactive carbothermal reduction of ZrC and ZrOC using Spark Plasma Sintering, ADVANCES IN APPLIED CERAMICS, Vol: 117, Pages: S34-S47, ISSN: 1743-6753
Pletser D, Ohashi T, Yoshii Y, et al., 2018, Temperature dependent volatilisation behaviour of Cs from two commercial adsorbents used at Fukushima measured using novel experimental apparatus, PROGRESS IN NUCLEAR ENERGY, Vol: 109, Pages: 214-222, ISSN: 0149-1970
Hsieh Y-H, Humphry-Baker SA, Horlait D, et al., 2018, Durability of hot uniaxially pressed Synroc derivative wasteform for EURO-GANEX wastes, JOURNAL OF NUCLEAR MATERIALS, Vol: 509, Pages: 43-53, ISSN: 0022-3115
Humphry-Baker S, Harrison R, Greaves G, et al., 2018, A candidate fusion engineering material, WC-FeCr, Scripta Materialia, Vol: 155, Pages: 129-133, ISSN: 1359-6462
A new candidate fusion engineering material, WC-FeCr, has been irradiated with He ions at 25 and 500 °C. Ions were injected at 6 keV to a dose of ~15 dpa and 50 at. % He, simulating direct helium injection from the plasma. The microstructural evolution was continuously characterised in situ using transmission electron microscopy. In the FeCr phase, a coarse array of 3–6 nm bubbles formed. In the WC, bubbles were less prominent and smaller (~2 nm). Spherical-cap bubbles formed at hetero-phase interfaces of tertiary precipitates, indicating that enhanced processing routes to minimise precipitation could further improve irradiation tolerance.
Patra N, Lee WE, 2018, Facile Precursor Synthesis of HfC-SiC Ultra-High-Temperature Ceramic Composite Powder for Potential Hypersonic Applications, ACS APPLIED NANO MATERIALS, Vol: 1, Pages: 4502-4508, ISSN: 2574-0970
Fossati PCM, Rushton MJD, Lee WE, 2018, Atomic-scale description of interfaces in rutile / sodium silicate glass-crystal composites, Physical Chemistry Chemical Physics, Vol: 20, Pages: 17624-17636, ISSN: 1463-9076
In this work interfaces between (Na2O)x(SiO2)1−x glasses (for x=0.0, 0.1 and 0.2) and TiO2 crystals are simulated using molecular dynamics and empirical potentials. Interfaces are presented for the distinct terminat- ing surfaces of TiO2 with Miller indices ≤ 2, the properties of which have been investigated using atomistic models. Simulations showed that par- tially ordered layers had been induced in the glass close to the interfaces, with successive oxygen-rich and cation-rich planes being noted. The first silicate layer in contact with the crystal tended to be highly-structured, with Si ions occupying well-defined positions that depend on the orien- tation of the crystal at the interface, and showing 2-dimensional ordering depending on glass composition. Finally, interface energies were calcu- lated. These indicated that the interface formation may stabilise a crystal surface in comparison to maintaining a free surface. Results are pre- sented suggesting that the structural flexibility of the glass network allows it to conform to the crystal, thereby providing charge compensation and avoiding large relaxation of the crystal structure close to the interfaces. Such interfacial properties could be crucial to improving phenomenologi- cal models of glass-crystal composite properties.
Gasparrini C, Chater RJ, Horlait D, et al., 2018, Zirconium carbide oxidation: kinetics and oxygen diffusion through the intermediate layer, Journal of the American Ceramic Society, Vol: 101, Pages: 2638-2652, ISSN: 0002-7820
Oxidation of hot‐pressed ZrC was investigated in air in the 1073‐1373 K range. The kinetics were linear at 1073 K, whereas at higher temperature samples initially followed linear kinetics before undergoing rapid oxidation leading to a Maltese cross shape of the oxide. The linear kinetics at 1073 K was governed by inward oxygen diffusion through an intermediate layer of constant thickness between ZrC and ZrO2 which was comprised of amorphous carbon and ZrO2 nanocrystals. Diffusion of oxygen through the intermediate layer was measured to be 9 × 10−10 cm2 s−1 using 18O as a tracer in a double oxidation experiment in 16O/18O. Oxidation at 1073 and 1173 K produced samples made of m‐ZrO2 and either t‐ or c‐ZrO2 with an adherent intermediate layer made of amorphous carbon and ZrO2, whereas oxidation at 1273 and 1373 K produced samples with a voluminous oxide made of m‐ZrO2 showing a gap between ZrC and the oxide. A substoichiometric zirconia layer was found at the gap at 1273 K and no carbon uptake was detected in this layer when compared with the top oxide layer. The loss of the intermediate layer and the slowdown of the linear rate constant (g m−2 s−1) at 1273 K compared to 1173 K was correlated with the preferential oxidation of carbon at the intermediate layer which would leave as CO and/or CO2 leaving a gap between ZrC and substoichiometric zirconia.
Patra N, Al Nasiri N, Jayaseelan DD, et al., 2018, Thermal properties of C-f/HfC and C-f/HfC-SiC composites prepared by precursor infiltration and pyrolysis, JOURNAL OF THE EUROPEAN CERAMIC SOCIETY, Vol: 38, Pages: 2297-2303, ISSN: 0955-2219
Chinnam RK, Fossati PCM, Lee WE, 2018, Degradation of partially immersed glass: A new perspective, Journal of Nuclear Materials, Vol: 503, Pages: 56-65, ISSN: 0022-3115
The International Simple Glass (ISG) is a six-component borosilicate glass which was developed as a reference for international collaborative studies on high level nuclear waste encapsulation. Its corrosion behaviour is typically examined when it is immersed in a leaching solution, or when it is exposed to water vapour. In this study, an alternative situation is considered in which the glass is only partially immersed for 7 weeks at a temperature of 90ºC. In this case, half of the glass sample is directly in the solution itself, and the other half is in contact with a water film formed by condensation of water vapour that evaporated from the solution. This results in a different degradation behaviour compared to standard tests in which the material is fully immersed. In particular, whilst in standard tests the system reaches a steady state with a very low alteration rate thanks to the formation of a protective gel layer, in partially-immersed tests this steady state could not be reached because of the continuous alteration from the condensate water film. The constant input of ions from the emerged part of the sample caused a supersaturation of the solution, which resulted in early precipitation of secondary crystalline phases.This setup mimics storage conditions once small amounts of water have entered a glass waste form containing canister. It offers a more realistic outlook of corrosion mechanisms happening in such situations than standard fully-immersed corrosion tests.
Ojovan MI, Burakov BE, Lee WE, 2018, Radiation-induced microcrystal shape change as a mechanism of wasteform degradation, Journal of Nuclear Materials, Vol: 501, Pages: 162-171, ISSN: 0022-3115
Experiments with actinide-containing insulating wasteforms such as devitrified glasses containing 244 Cm, Ti-pyrochlore, single-phase La-monazite, Pu-monazite ceramics, Eu-monazite and zircon single crystals containing 238 Pu indicate that mechanical self-irradiation-induced destruction may not reveal itself for many years (even decades). The mechanisms causing these slowly-occurring changes remain unknown therefore in addition to known mechanisms of wasteform degradation such as matrix swelling and loss of solid solution we have modelled the damaging effects of electrical fields induced by the decay of radionuclides in clusters embedded in a non-conducting matrix. Three effects were important: (i) electric breakdown; (ii) cluster shape change due to dipole interaction, and (iii) cluster shape change due to polarisation interaction. We reveal a critical size of radioactive clusters in non-conducting matrices so that the matrix material can be damaged if clusters are larger than this critical size. The most important parameters that control the matrix integrity are the radioactive cluster (inhomogeneity) size, specific radioactivity, and effective matrix electrical conductivity. We conclude that the wasteform should be as homogeneous as possible and even electrically conductive to avoid potential damage caused by electrical charges induced by radioactive decay.
Lerdprom W, Bhowmik A, Grasso S, et al., 2018, Impact of spark plasma sintering (SPS) on mullite formation in porcelains, JOURNAL OF THE AMERICAN CERAMIC SOCIETY, Vol: 101, Pages: 525-535, ISSN: 0002-7820
Kota S, Zapata-Solvas E, Chen Y, et al., 2017, Isothermal and Cyclic Oxidation of MoAlB in Air from 1100 degrees C to 1400 degrees C, JOURNAL OF THE ELECTROCHEMICAL SOCIETY, Vol: 164, Pages: C930-C938, ISSN: 0013-4651
Lerdprom W, Zapata-Solvas E, Jayaseelan DD, et al., 2017, Impact of microwave processing on porcelain microstructure, CERAMICS INTERNATIONAL, Vol: 43, Pages: 13765-13771, ISSN: 0272-8842
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