Browsing by Author "Moricca, SA"

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    Alternative Synroc formulations
    (Cambridge University Press/Springer Nature, 2011-02-25) Vance, ER; Smith, KL; Thorogood, GJ; Begg, BD; Moricca, SA; Angel, PJ; Stewart, MWA; Ball, CJ
    Perovskite is the least durable of the resistate minerals comprising Synroc-C and it is desirable to reduce its abundance in Synroc. Kinetic limitations and competition with Csapparently affect the incorporation of Sr into hollandite during hot-pressing at 1200°C/20 MPa so that ~ 10% of perovskite (a value below the percolation limit) is probably an optimum target. Zirconolite-rich Synroc formulations have been prepared for actinide-rich wastes. Background XRD and TEM studies have also been performed to study the crystal-chemical behaviour of Nd (a simulant of trivalent actinides) in zirconolite. Either rare-earth compensated perovskite or freudenbergite in Synroc can evidently be used to immobilise Na-bearing HLW. © 1992 Materials Research Society
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    Colloidal processing of zirconium diboride ultra-high temperature ceramics
    (John Wiley and Sons, 2013-05-21) Tallon, C; Chavara, DT; Gillen, AL; Riley, D; Edwards, L; Moricca, SA; Franks, GV
    Colloidal processing of the Ultra-High Temperature Ceramic (UHTC) zirconium diboride (ZrB2) to develop near−net-shaping techniques has been investigated. The use of the colloidal processing technique produces higher particle packing that ultimately enables achieving greater densification at lower temperatures and pressures, even pressureless sintering. ZrB2 suspension formulations have been optimized in terms of rheological behavior. Suspensions were shaped into green bodies (63% relative density) using slip casting. The densification was carried out at 1900°C, 2000°C, and 2100°C, using both hot pressing at 40 MPa and pressureless sintering. The colloidally processed materials were compared with materials prepared by a conventional dry processing route (cold pressed at 50 MPa) and subjected to the same densification procedures. Sintered densities for samples produced by the colloidal route are higher than produced by the dry route (up to 99.5% relative density by hot pressing), even when pressureless sintering is performed (more than 90% relative density). The promising results are considered as a starting point for the fabrication of complex-shaped components that can be densified at lower sintering temperatures without pressure. © 2013, The American Ceramic Society.
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    Crystal chemistry and phase manipulation in Synroc
    (Trans Tech Publications Ltd, 1991) Vance, ER; Moricca, SA; Thorogood, GJ; Lumpkin, GR
    Synroc is a multi-phase ceramic designed for geological immobilisation of radioactive waste produced by reprocessing nuclear fuel from power reactors [1]. The main crystalline phases are hollandite, perovskite, zirconolite, and reduced titanium oxide. The compositions of these phases and the nuclides they can incorporate in solid solution are shown in Table 1. Table 1. Principal Phases comprising Synroc Phase Nominal Composition Waste nuclides incorporated Estimated wt%* [2] Hollandite Ba1.14(Al, Tr3+)2.28Ti6O16 Cs, Sr, Rb 25 Perovskite CaTio3 Sr, RE, An 20 Zirconolite CaZrTi2O7 RE, An 20 Titanium Oxide TinO2n-1 - 35 *No HLW present RE = rare earths, An = actinides. The main (Synroc-C) formulation is designed for Purex reprocessing waste and the standard composition is wt%: Al2O3(4.3); BaO(4.4); CaO(8.8); ZrO2(5.6); TiO2(57.9); waste oxides (20). The loading of high-level waste (HLW) oxides can be varied if desired, but probably cannot exceed a value of 30-35% [2]. Several variants of this composition have been formulated at the laboratory scale, with Synroc-D, E and F being directed towards Savannah River (U.S.A.) military waste, encapsulation of high-level nuclear reprocessing waste and unreprocessed spent fuel respectively. © 1991 Trans Tech Publications Ltd.
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    Flexible process options for the immobilisation of residues and wastes containing plutonium
    (American Society of Mechanical Engineers (ASME), 2007-09-02) Stewart, MWA; Moricca, SA; Begg, BD; Day, RA; Scales, CR; Maddrell, ER; Eilbeck, AB
    Residues and waste streams containing plutonium present unique technical, safety, regulatory, security, and sociopolitical challenges. In the UK these streams range from lightly plutonium contaminated materials (PCM) through to residues resulting directly from Pu processing operations. In addition there are potentially stocks of Pu oxide powders whose future designation may be either a waste or an asset, due to their levels of contamination making their reuse uneconomic, or to changes in nuclear policy. While waste management routes exist for PCM, an immobilisation process is required for streams containing higher levels of Pu. Such a process is being developed by Nexia Solutions and ANSTO to treat and immobilise Pu waste and residues currently stored on the Sellafield site. The characteristics of these Pu waste streams are highly variable. The physical form of the Pu waste ranges from liquids, sludges, powders/granules, to solid components (e.g., test fuels), with the Pu present as an ion in solution, as a salt, metal, oxide or other compound. The chemistry of the Pu waste streams also varies considerably with a variety of impurities present in many waste streams. Furthermore, with fissile isotopes present, criticality is an issue during operations and in the store or repository. Safeguards and security concerns must be assessed and controlled. The process under development, by using a combination of tailored waste form chemistry combined with flexible process technology aims to develop a process line to handle a broad range of Pu waste streams. It aims to be capable of dealing with not only current arisings but those anticipated to arise as a result of future operations or policy changes.
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    M(n+1)AXn phases are they tolerant/resistant to damage
    (Australian Institute of Physics, 2011-02-03) Whittle, KR; Riley, DP; Blackford, MG; Aughterson, RD; Moricca, SA; Lumpkin, GR; Zaluzec, NJ
    Ternary carbide materials have been proposed as having applications within the future nuclear technologies, both fusion (ITER/DEMO) and fission (Gen IV). These new designs require a material to have the ability to tolerate radiation damage to high levels, with a high level of predictability. As part of such a process two systems, specifically Ti3AlC2 and Ti3SiC2 have been studied to determine their radiation tolerance, using in-situ ion beam irradiation with 1 MeV Xe ions, coupled with transmission electron microscopy. Irradiations have shown that both systems show little amorphisation at 300K up to doses of at least 6.25 x 1015 ions cm-2 (~28-30 dpa). However, there is a subtle difference between Ti3AlC2 and Ti3SiC2, with Ti3SiC2 showing more evidence for damage. Further irradiations using 500 KeV Xe to fluences equivalent to 100 dpa have also been undertaken, with crystalline material visible and evidence of recrystallisation. Explanations and possible mechanisms for recovery from damage are presented, along with implications for future potential uses.
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    Microstructural evolution and final properties of a cold-swaged multifunctional Ti-Nb-Ta-Zr-O alloy produced by a powder metallurgy route
    (Elsevier, 2013-07-15) Guo, W; Quadir, MZ; Moricca, SA; Eddows, T; Ferry, M
    Body centred cubic (BCC) β-phase multifunctional titanium alloys have been developed with a very unique combination of thermal and mechanical properties. In this investigation, a very low porosity Ti–36.8–Nb–2.7Zr–2.0Ta–0.44O (wt%) alloy was produced by powder sintering, hot forging, solution treatment and cold swaging. X-ray diffraction and transmission electron microscopy (TEM) of the solution treated alloy revealed the presence of a small amount of ω-phase in a predominantly BCC β-phase matrix. Electron backscatter diffraction (EBSD) of the swaged alloy revealed a highly elongated and fragmented microstructure, and a strong 〈110〉 fibre texture. TEM also revealed the existence of stress-induced twin lamella, dislocations and ω-phase. Consistent with previous studies on these types of alloys, the swaged alloy exhibited non-linear elasticity during tensile straining, low elastic modulus (45.4 GPa), high elastic limit (2.3%), high elongation to failure (8.1%), and a high yield strength (880 MPa) and tensile strength (940 MPa). The coefficient of thermal expansion was also low (∼5×10−6 K−1 between 50 and 300 °C) in this alloy. © 2013, Elsevier B.V.
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    Radiation tolerance of M(n+1)AX(n) phases, Ti3AlC2 and Ti3SiC2
    (Elsevier, 2010-08-01) Whittle, KR; Blackford, MG; Aughterson, RD; Moricca, SA; Lumpkin, GR; Riley, DP; Zaluzec, NJ
    During investigations of novel material types with uses in future nuclear technologies (ITER/DEMO and GenIV fission reactors), ternary carbides with compositions Ti3AlC2 and Ti3SiC2 have been irradiated with high Xe fluences, 6.25 × 1015 ions cm−2 (25–30 dpa), using the IVEM-TANDEM facility at Argonne National Laboratory. Both compositions show high tolerance to damage, and give indications that they are likely to remain crystalline to much higher fluences. There is a visible difference in tolerance between Ti3AlC2 and Ti3SiC2 that can be related to the changes in bonding within each material. These initial findings provide evidence for a novel class of materials (+200 compounds) with high radiation resistance, while, significantly, both of these materials are composed of low-Z elements and hence exhibit no long-term activation. © 2010, Elsevier Ltd.
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    Safe immobilization of high-level radioactive waste in waste forms for geological repositories
    (Elsevier, 2011-10) Vance, ER; Stewart, MWA; Moricca, SA; Lumpkin, GR; Begg, BD
    The idea that spent fuel and other hazardous radioactive high-level wastes (HLW) would need to be dealt with arose soon after the first experimental demonstration of nuclear reactors in 1942. HLW is spent nuclear power plant fuel or waste deriving directly from reprocessing or recycling of spent fuel. This latter waste consists of mainly fission products (FPs), as well as minor actinides such as Np, Pu, Am and Cm. Table 1 shows the main components and relevant half-lives of reprocessing waste from the well-known Purex process. Also, there are abundant wastes from the production of Pu for nuclear weapons, mainly in the US and Russia. These wastes (Table 2), although designated as HLW in the US, have only around 0.1- 1% of the radioactivity per unit volume of the Purex-type HLW, and if they were located in most other countries would be categorised as intermediate-level waste (ILW). Current HLW inventories around the world run into tens of millions of tonnes. The manageability of HLW impacts directly on the sustainability of nuclear power. © 2011 Elsevier Ltd.
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    Spray-dried microspheres as a route to clay/polymer nanocomposites
    (Wiley-Blackwell, 2008-05-05) Yun, SI; Attard, DJ; Lo, V; Davis, J; Li, HJ; Latella, BA; Tsvetkov, F; Noorman, H; Moricca, SA; Knott, RB; Hanley, HJM; Morcom, M; Simon, GP; Gadd, GE
    A new strategy for the preparation of well-dispersed clays in a polymer matrix by a spray-drying method is presented. Scanning electron microscopy and transmission electron microscopy measurements show that the spray-drying process produces clay/polymer microspheres in which the clay is trapped in a well-dispersed state throughout the polymer matrix. The microspheres have been successfully extruded into clay/poly(methyl methacrylate) nanocomposite bulk structures without any perturbation of the well-dispersed clay nanostructure in the original microspheres. Transmission electron microscopy and small-angle X-ray scattering show that the clay particles in the extruded materials range from single platelets to simple tactoids composed of a few stacked clay platelets, indicating an excellent degree of dispersion. The results show that sprayed microspheres are very good precursors for further processing such as extrusion or melt blending with other polymers for bulk nanocomposite fabrication. © 2008, Wiley-Blackwell. The definitive version is available at www3.interscience.wiley.com
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    Synroc formulations for various high-level nuclear wastes
    (American Ceramic Society, 1991-04-28) Vance, ER; Moricca, SA; Thorogood, GJ; Smith, KL
    Of the chemically resistant titanate phases comprising Synroc, perovskite is the least durable and adjustments to the Synroc formulation to reduce the amount of perovskite are under study. Results are described of experiments designed to reduce the perovskite level to < 5 wt%. A simulated waste loading of 20 wt% was used. Other Synroc formulations for high Al, high (Al + U), high (Na + Al + Fe + U) and actinide-rich waste have been calculated and tested experimentally via leaching, X-ray diffraction and electron microscopy.
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    Thermal expansion of solutions of deuteromethane in fullerite C60 at low temperatures. Isotopic effect
    (American Institute of Physics, 2009-03) Dolbin, AV; Vinnikov, NA; Gavrilko, VG; Esel'son, VB; Manzheliĭ, VG; Gadd, GE; Moricca, SA; Cassidy, DJ; Sundqvist, B
    The thermal expansion of CD4 solutions in the orientational glass C60 with molar concentration of deuteromethane 20 and 50% has been investigated in the temperature range 2.5–23 K. The orientational glass CD4–C60 undergoes a first-order phase transition in the temperature interval 4.5–55 K. This transition is manifested as hysteresis of the linear thermal expansion coefficient α as well as maxima in the temperature dependences α(T) and τ1(T), where τ1 is the characteristic thermalization time of the experimental samples. The characteristic re-orientation times of the C60 molecules and the characteristic phase transformations occurring in the experimental solutions are determined. The results of the present study are compared with the results of a similar study of the solution CH4–C60. It is concluded that tunneling rotation of the CH4 and CD4 molecules occupying interstitial positions in the fullerite C60 lattice occurs. © 2009, American Institute of Physics
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    The world's smallest gas cylinders?
    (American Association for the Advancement of Science (AAAS), 1997-08-15) Gadd, GE; Blackford, MG; Moricca, SA; Webb, N; Evans, PJ; Smith, AM; Jacobsen, GE; Leung, S; Day, A; Hua, Q
    Argon gas was trapped at high pressure within hollow carbon tubes grown in vapor that have an outer diameter of between 20 and 150 nanometers. The gas was forced into the tubes by hot isostatically pressing (HIPing) the carbon material for 48 hours at 650°C under an argon pressure of 170 megapascals. Energy dispersive x-ray spectroscopy maps and line scans across the tubes show that the argon is trapped inside the bore and not in the tube walls. The room temperature argon pressure in these tubes was estimated to be about 60 megapascals, which indicates that equilibrium pressure was attained within the tubes at the HIPing temperature. These findings demonstrate the potential for storing gases in such carbon structures. © 1997 American Association for the Advancement of Science
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    Zirconolite-rich titanate ceramics for immobilisation of actinides - waste form/HIP can interactions and chemical durability
    (Elsevier, 2009-12) Zhang, YJ; Stewart, MWA; Li, HJ; Carter, ML; Vance, ER; Moricca, SA
    Zirconolite-based titanate ceramics containing U plus Th or Pu have been prepared. The final consolidation to produce a dense monolithic waste form was carried out using hot isostatic pressing (HIPing) of the calcined materials within a stainless steel can. The ceramics were characterised and tested for their overall feasibility to immobilise impure Pu or separated actinide-rich radioactive wastes. As designed, tetravalent U and Pu are mainly incorporated in a durable zirconolite phase, together with Gd or Hf added as neutron absorbers. The interaction of the waste form with the HIP can was also examined. No changes in the U valences or the U/Pu-bearing phase distributions were observed at the waste form–HIP can interface. © 2009, Elsevier Ltd.