394 results found
Harrison R, 2015, Mechanism and Kinetics of Oxidation of ZrN Ceramics, Journal of the American Ceramic Society, ISSN: 1551-2916
Oxidation of ZrN ceramics from 973–1373 K under static conditions reveals parabolic rate behavior, indicative of a diffusion-controlled process. In-situ high temperature powder XRD found the oxidation mechanism begins with destabilization of ZrN through formation of a ZrN1−x phase with oxide peaks initially detected at around 773 K. The zirconium oxide layer was found to be monoclinic by in-situ XRD with no evidence of tetragonal or cubic polymorphs present to 1023 K. Bulk ceramic samples oxidized at 1173 and 1273 K underwent slower oxidation than those oxidized at 973 and 1073 K. This change in oxidation rate and hence mechanism was due to formation of a denser c-ZrO2 polymorph stabilized by nitrogen defects. This N-doped dense ZrO2 layer acts as a diffusion barrier to oxygen diffusion. However, at an oxidation temperature of 1373 K this layer is no longer protective due to increased diffusion through it resulting in grain boundary oxidation.
Quadling A, Vandeperre L, Parkes M, et al., 2015, Second Phase-Induced Degradation of Fused MgO Partially Stabilized Zirconia Aggregates, JOURNAL OF THE AMERICAN CERAMIC SOCIETY, Vol: 98, Pages: 1364-1371, ISSN: 0002-7820
Ojovan MI, Lee WE, 2015, About U(t) form of pH-dependence of glass corrosion rates at zero surface to volume ratio, Pages: 153-161, ISSN: 0272-9172
© 2015 Materials Research Society. The pH-dependence of glass corrosion rates has a well-known U-shaped form with minima for near-neutral solutions. This paper analyses the change of U-shaped form with time and reveals that the pH dependence evolves even for solutions that have pH not affected by glass corrosion mathematically corresponding to a zero surface to volume ratio. The U(t) dependence is due to changes of concentration profiles of elements in the near-surface layers of glasses in contact with water and is most evident within the initial stages of glass corrosion at relatively low temperatures. Numerical examples are given for the nuclear waste borosilicate glass K-26 which is experimentally characterised by an effective diffusion coefficient of caesium Dc 4.5 10-12 cm2/day and by a rate of glass hydrolysis in non-saturated groundwater as high as n, = 100 nm/year The changes of U-shaped form need to be accounted when assessing the performance of glasses in contact with water solutions.
Hiezl Z, Hambley DI, Padovani C, et al., 2015, Processing and microstructural characterisation of a UO2-based ceramic for disposal studies on spent AGR fuel, JOURNAL OF NUCLEAR MATERIALS, Vol: 456, Pages: 74-84, ISSN: 0022-3115
Ahmad NE, Fearn S, Lee WE, et al., 2014, Preliminary surface study of short term dissolution of UK high level waste glass, 2nd International Summer School on Nuclear Glass Wasteform: Structure, Properties and Long-Term Behavior, SumGLASS 2013, Pages: 230-236
Zapata-Solvas E, Jayaseelan DD, Brown PM, et al., 2014, Effect of La2O3 addition on long-term oxidation kinetics of ZrB2-SiC and HfB2-SiC ultra-high temperature ceramics, JOURNAL OF THE EUROPEAN CERAMIC SOCIETY, Vol: 34, Pages: 3535-3548, ISSN: 0955-2219
Gonzalez-Julian J, Cedillos-Barraza O, Doring S, et al., 2014, Enhanced oxidation resistance of ZrB2/SiC composite through in situ reaction of gadolinium oxide in patterned surface cavities, JOURNAL OF THE EUROPEAN CERAMIC SOCIETY, Vol: 34, Pages: 4157-4166, ISSN: 0955-2219
Lee WE, Giorgi E, Harrison R, et al., 2014, Nuclear Applications for Ultra-High Temperature Ceramics and MAX Phases, Ultra-High Temperature Ceramics: Materials for Extreme Environment Applications, Pages: 391-415, ISBN: 9781118700785
© 2014 The American Ceramic Society. All rights reserved. Future nuclear reactor systems and the severe conditions under which they will operate are reviewed. Current nuclear applications of ceramics are predominantly as oxide fuels as well as ceramic/glassy waste forms, although non-oxides do find niche uses such as graphite moderators and B4C control rods. UHTCs properties of interest to the nuclear industry include that they may be fissile, and that they have high thermal conductivity, refractoriness, and phase stability. Using such properties, future nuclear ceramics will potentially include UHTCs, for example, as non-oxide fuels (U/Pu carbides and nitrides) and fuel cladding (TaC, ZrC, HfC). MAX phases may also find application as fuel cladding. Oxide and non-oxide composite (e.g., SiC/SiC) and inert matrix fuel systems are under development for future fission reactors while uses of ceramics in fusion reactor systems will be both functional (such as the ceramic superconductors in the magnet systems for controlling the plasma) and structural in various locations outside of the first wall in magnetic confinement fusion. Finally, the importance of thermodynamics in severe conditions and the need for accurate thermodynamics databases are highlighted.
Fahrenholtz WG, Wuchina EJ, Lee WE, et al., 2014, Ultra-High Temperature Ceramics: Materials for Extreme Environment Applications, ISBN: 9781118700785
© 2014 The American Ceramic Society. All rights reserved. The first comprehensive book to focus on ultra-high temperature ceramic materials in more than 20 years Ultra-High Temperature Ceramics are a family of compounds that display an unusual combination of properties, including extremely high melting temperatures (>3000°C), high hardness, and good chemical stability and strength at high temperatures. Typical UHTC materials are the carbides, nitrides, and borides of transition metals, but the Group IV compounds (Ti, Zr, Hf) plus TaC are generally considered to be the main focus of research due to the superior melting temperatures and stable high-melting temperature oxide that forms in situ. Rather than focusing on the latest scientific results, Ultra-High Temperature Ceramics: Materials for Extreme Environment Applications broadly and critically combines the historical aspects and the state-of-the-art on the processing, densification, properties, and performance of boride and carbide ceramics. In reviewing the historic studies and recent progress in the field, Ultra-High Temperature Ceramics: Materials for Extreme Environment Applications provides: • Original reviews of research conducted in the 1960s and 70s • Content on electronic structure, synthesis, powder processing, densification, property measurement, and characterization of boride and carbide ceramics. • Emphasis on materials for hypersonic aerospace applications such as wing leading edges and propulsion components for vehicles traveling faster than Mach 5 • Information on materials used in the extreme environments associated with high speed cutting tools and nuclear power generation Contributions are based on presentations by leading research groups at the conference "Ultra-High Temperature Ceramics: Materials for Extreme Environment Applications II" held May 13-19, 2012 in Hernstein, Austria. Bringing together disparate researchers from academia, government, and
Patra N, Jayaseelan DD, Lee WE, 2014, Synthesis of Biopolymer-Derived Zirconium Carbide Powder by Facile One-Pot Reaction, Journal of the American Ceramic Society, Vol: 98, Pages: 71-77, ISSN: 1551-2916
Zirconium carbide (ZrC) was synthesized by polycondensation and carbothermal reduction reactions from an organic–inorganic hybrid complex. A natural biopolymer Gum Karaya (GK) and zirconyl oxychloride octahydrate (ZOO) were used as the sources of carbon and zirconium, respectively. FTIR of as-synthesized dried complexes revealed formation of Zr–O. Pyrolysis of the complexes at 1200°C/1 h under argon resulted in tetragonal and monoclinic zirconia which after heat treatment at 1400°C–1550°C transformed to zirconium carbide. Thermal analysis shows that the GK–ZOO complexes lost less mass than the pristine GK to 600°C. The intensity of exothermic decomposition decreases and shifted to higher temperature for the hybrid complexes indicating that zirconia induced thermal stability. A maximum ZrC yield of ~60 wt% is obtained for the intermediate GK–ZOO ratio of 1:2. Particles pyrolyzed for 1 h at 1550°C were coarser (5–10 μm) with flakes for lower GK–ZOO weight ratio, but were spheroidal with narrow size distribution (~1 μm) with increasing GK–ZOO weight ratio.
Harrison R, Ridd O, Jayaseelan DD, et al., 2014, Thermophysical characterisation of ZrCxNy ceramics fabricated via carbothermic reduction-nitridation, JOURNAL OF NUCLEAR MATERIALS, Vol: 454, Pages: 46-53, ISSN: 0022-3115
Kubal SK, Pleydell-Pearce C, Powson JR, et al., 2014, Postmortem analysis of BOF tuyeres, Pages: 471-476
© 2014 The American Ceramic Society. In a Basic Oxygen Furnace with bottom agitation, wear around the bottom-blowing elements is one of the main factors limiting lining service life. Therefore, worn tuyeres from a Basic Oxygen Steelmaking converter have been retrieved and characterized after a completed campaign. Chemical and microscopic analysis revealed modifications in MgO monolithic material. These changes involved sintering of the working surface to a depth of ≤ 55 mm, and both intergranular and grain boundary cracking, perpendicular to the ramming direction. The tuyere wear mechanism was defined to occur due to cracking and spalling resulting from a combination of thermo-mechanical stresses enhanced by molten metal and slag penetration, and possibly manufacturing method.
Alex J, Vandeperre L, Touzo B, et al., 2014, EFFECT OF SODIUM IMPURITIES ON PHASE AND MICROSTRUCTURE EVOLUTION IN CALCIUM ALUMINATE CEMENT BONDED CASTABLES AT HIGH TEMPERATURES, 13th Unified International Technical Conference on Refractories (UNITECR), Publisher: AMER CERAMIC SOC, Pages: 911-+
Lee WE, Ojovan MI, Thomas GA, 2014, The uk's radioactive waste and waste management programme, Pages: 157-175, ISSN: 1042-1122
Sources of the UKs radioactive wastes including from power production, military programmes, medical uses and research reactors are described along with options for managing controlled wastes from pretreatment, treatment, conditioning and storage stages through to transportation to final disposal. Immobilisation (wasteform), temporary storage and permanent disposal options including near surface, deep and very deep geological disposal are covered.
Ahmad NE, Jones JR, Lee WE, 2014, Durability studies of simulated UK high level waste glass, Pages: 291-296, ISSN: 0272-9172
© 2014 Materials Research Society. A simulated Magnox glass which is Mg- and Al- rich was subjected to aqueous corrosion in static mode with deionised water at 90 °C for 7-28 days and assessed using X-Ray Diffraction (XRD), Scanning Electron Microscopy (SEM) with Energy X-Ray Dispersive Spectroscopy (EDS) and Inductively Coupled Plasma - Optical Emission Spectroscopy (ICP-OES). XRD revealed both amorphous phase and crystals in the glass structure. The crystals were Ni and Cr rich spinels and ruthenium oxide. After two weeks of incubation in deionised water, the glass surface was covered by a ∼ 11 μm thick Si-rich layer whilst mobile elements and transition metals like Na, B, and Fe were strongly depleted. The likely corrosion mechanism and in particular the role of Mg and Al in the glass structure are discussed. Keywords: high level waste glass, durability, corrosion mechanism.
Hiezl Z, Hambley D, Lee WE, 2014, Preparation and characterization of UO<inf>2</inf>-based AGR SIMFuel, Journal of Fluid Mechanics, Vol: 1665, ISSN: 0022-1120
Copyright © 2014 Materials Research Society . Preparation and characterization of a Simulated Spent Nuclear Fuel (SIMFuel), which replicates the chemical state and microstructure of Spent Nuclear Fuel (SNF) discharged from UK Advanced Gas-cooled Reactor (AGR) after a cooling time of 100 years is described. Thirteen stable elements were added to depleted UO2 and sintered to simulate the composition of fuel pellets after burn-ups of 25 and 43 GWd/tU and, as a reference, pure UO2 pellets were also investigated. The fission product distribution was calculated using the Fispin code provided by NNL. SIMFuel pellets exhibit a microstructure up to 92% TD. During the sintering process in H2 atmosphere Mo-Ru-Rh-Pd metallic precipitates and grey-phase ((Ba, Sr)(Zr, RE)O3 oxide precipitates) formed within the UO2 matrix. These secondary phases are present in real PWR and AGR SNF, although they are smaller in size than those examined in this study. The grain size of the produced SIMFuel is in good agreement with literature references.
Hiezl Z, Hambley D, Lee WE, 2014, Preparation and characterization of UO<inf>2</inf>-based AGR SIMFuel, Pages: 245-251, ISSN: 0272-9172
© 2014 Materials Research Society. Preparation and characterization of a Simulated Spent Nuclear Fuel (SIMFuel), which replicates the chemical state and microstructure of Spent Nuclear Fuel (SNF) discharged from UK Advanced Gas-cooled Reactor (AGR) after a cooling time of 100 years is described. Thirteen stable elements were added to depleted UO2 and sintered to simulate the composition of fuel pellets after burn-ups of 25 and 43 GWd/tU and, as a reference, pure UO2 pellets were also investigated. The fission product distribution was calculated using the Fispin code provided by NNL. SIMFuel pellets exhibit a microstructure up to 92% TD. During the sintering process in H2 atmosphere Mo-Ru-Rh-Pd metallic precipitates and grey-phase ((Ba, Sr)(Zr, RE)O3 oxide precipitates) formed within the UO2 matrix. These secondary phases are present in real PWR and AGR SNF, although they are smaller in size than those examined in this study. The grain size of the produced SIMFuel is in good agreement with literature references.
Quadling A, Vandeperre L, Lee WE, et al., 2014, HIGH TEMPERATURE CHARACTERISTICS OF REFRACTORY ZIRCONIA CRUCIBLES USED FOR VACUUM INDUCTION MELTING, 13th Unified International Technical Conference on Refractories (UNITECR), Publisher: AMER CERAMIC SOC, Pages: 107-+
Ye J, Zhang S, Lee WE, 2013, Molten salt synthesis and characterization of SiC coated carbon black particles for refractory castable applications, JOURNAL OF THE EUROPEAN CERAMIC SOCIETY, Vol: 33, Pages: 2023-2029, ISSN: 0955-2219
Ye J, Thackray RP, Lee WE, et al., 2013, Microstructure and rheological properties of titanium carbide-coated carbon black particles synthesised from molten salt, JOURNAL OF MATERIALS SCIENCE, Vol: 48, Pages: 6269-6275, ISSN: 0022-2461
Manara D, Jackson HF, Perinetti-Casoni C, et al., 2013, The ZrC-C eutectic structure and melting behaviour: A high-temperature radiance spectroscopy study, JOURNAL OF THE EUROPEAN CERAMIC SOCIETY, Vol: 33, Pages: 1349-1361, ISSN: 0955-2219
Zapata-Solvas E, Jayaseelan DD, Lin HT, et al., 2013, Mechanical properties of ZrB2- and HfB2-based ultra-high temperature ceramics fabricated by spark plasma sintering, JOURNAL OF THE EUROPEAN CERAMIC SOCIETY, Vol: 33, Pages: 1373-1386, ISSN: 0955-2219
Lee WE, Gilbert M, Murphy ST, et al., 2013, Opportunities for Advanced Ceramics and Composites in the Nuclear Sector, JOURNAL OF THE AMERICAN CERAMIC SOCIETY, Vol: 96, Pages: 2005-2030, ISSN: 0002-7820
Zapata-Solvasa E, Jayaseelan DD, Brown PM, et al., 2013, Thermal properties of La2O3-doped ZrB2- and HfB2-based ultra-high temperature ceramics
Thermal properties of La2O3-doped ZrB2- and HfB2-based ultra high temperature ceramics (UHTCs) have been measured at temperatures from room temperature to 2000 °C and compared with SiC-doped ZrB2- and HfB2-based UHTCs and monolithic ZrB2 and HfB2. Thermal conductivities of La2O3-doped UHTCs remain constant around 55–60 W/mK from 1500 °C to 1900 °C while SiC-doped UHTCs showed a trend to decreasing values over this range.
Cui B, Zapata-Solvas E, Reece MJ, et al., 2013, Microstructure and High-temperature Oxidation Behavior of Ti3AlC2/W Composites, JOURNAL OF THE AMERICAN CERAMIC SOCIETY, Vol: 96, Pages: 584-591, ISSN: 0002-7820
Jantzen CM, Lee WE, Ojovan MI, 2013, Radioactive waste (RAW) conditioning, immobilization, and encapsulation processes and technologies: Overview and advances, Pages: 171-272
The main immobilization technologies that have been demonstrated for radioactive waste disposal are cementation, bituminization, and vitrification. Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Nations that have generated HLW are immobilizing in either alkali borosilicate glass or alkali aluminophosphate glass. The compositions of nuclear waste glasses are tailored for easy preparation and melting, avoidance of glass-in-glass phase separation, avoidance of uncontrolled crystallization, and acceptable chemical durability. Future waste generation is driven by interest in sources of clean energy. The development of advanced waste forms is a necessary component of the new nuclear power plant (NPP) flowsheets. A brief summary is given of existing and advanced waste forms and processing technologies. © 2013 Woodhead Publishing Limited All rights reserved.
Lee WE, Ojovan MI, Jantzen CM, 2013, Radioactive waste management and contaminated site clean-up: Processes, technologies and international experience, Radioactive Waste Management and Contaminated Site Clean-Up: Processes, Technologies and International Experience, Pages: 1-879
Radioactive waste management and contaminated site clean-up reviews radioactive waste management processes, technologies, and international experiences. Part one explores the fundamentals of radioactive waste including sources, characterisation, and processing strategies. International safety standards, risk assessment of radioactive wastes and remediation of contaminated sites and irradiated nuclear fuel management are also reviewed. Part two highlights the current international situation across Africa, Asia, Europe, and North America. The experience in Japan, with a specific chapter on Fukushima, is also covered. Finally, part three explores the clean-up of sites contaminated by weapons programmes including the USA and former USSR. © 2013 Woodhead Publishing Limited. All rights reserved.
Lee WE, Ojovan MI, 2013, Fundamentals of radioactive waste (RAW): Science, sources, classification and management strategies, Pages: 3-49
Classification systems for the types of radioactive waste (RAW) are described along with sources of controlled wastes (including from power production, military programmes, medical uses and research reactors) and uncontrolled or accidental releases. Options for managing controlled wastes from pretreatment, treatment, conditioning and storage stages through to transportation to final disposal are considered. Immobilisation (wasteform), temporary storage and permanent disposal options including near surface, deep and very deep geological disposal are covered as well as strategies for uncontrolled releases. © 2013 Woodhead Publishing Limited All rights reserved.
Ojovan MI, Lee WE, 2013, An Introduction to Nuclear Waste Immobilisation: Second Edition, An Introduction to Nuclear Waste Immobilisation: Second Edition, Pages: 1-362
Drawing on the authors' extensive experience in the processing and disposal of waste, An Introduction to Nuclear Waste Immobilisation, Second Edition examines the gamut of nuclear waste issues from the natural level of radionuclides in the environment to geological disposal of waste-forms and their long-term behavior. It covers all-important aspects of processing and immobilization, including nuclear decay, regulations, new technologies and methods. Significant focus is given to the analysis of the various matrices used, especially cement and glass, with further discussion of other matrices such as bitumen. The final chapter concentrates on the performance assessment of immobilizing materials and safety of disposal, providing a full range of the resources needed to understand and correctly immobilize nuclear waste. The fully revised second edition focuses on core technologies and has an integrated approach to immobilization and hazards Each chapter focuses on a different matrix used in nuclear waste immobilization: cement, bitumen, glass and new materials Keeps the most important issues surrounding nuclear waste - such as treatment schemes and technologies and disposal - at the forefront. © 2014 Elsevier Ltd All rights reserved.
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