Preparation, sintering and leaching of optimized uranium thorium dioxides (original) (raw)
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The possible usage of ex-ADU uranium dioxide fuel pellets with low-temperature sintering
Journal of Nuclear Materials, 2000
UO 2 fuel pellets are prepared conventionally by high-and low-temperature sintering. Both ex-ammonium uranyl carbonate (AUC) and ex-ammonium diuranate (ADU) UO 2 are used in these techniques. Larger quantities of chemicals (NH 3 and CO 2 ) are required for the AUC process which is a signi®cant complication compared with the ADU process. In this work, it is aimed to search the important parameters such as theoretical density (TD) and grain size of ex-ADU UO 2 pellets which are in an oxidative atmosphere at low temperature. A density of %95% TD is obtained at 1150°C. The grain structure of the pellets sintered in an oxidative atmosphere is monomodal unlike the duplex grain structure of ex-AUC oxide, but it is in accord with ex-ammonium uranate (AU) oxide. The grain structure, distribution and size are compared with the ex-AUC and ex-AU conversion processes. It is shown that the grain structure development depends on both the powder properties and the sintering atmosphere. Ó
Journal of Nuclear Materials, 2014
Diffusion mechanisms occurring during the sintering of oxide ceramics are affected by the oxygen content of the atmosphere, as it imposes the nature and the concentration of structural defects in the material. Thus, the oxygen partial pressure, p(O 2), of the sintering gas has to be precisely controlled, otherwise a large dispersion in various parameters, critical for the manufacturing of ceramics such as nuclear oxides fuels, is likely to occur. In the present work, the densification behaviour and the solid solution formation of a mixed uraniumplutonium oxide (MOX) were investigated. The initial mixture, composed of 70% UO 2 + 30% PuO 2 , was studied at p(O 2) ranging from 10 À15 to 10 À4 atm up to 1873 K both with dilatometry and in situ high temperature X-ray diffraction. This study has shown that the initial oxides UO 2+x and PuO 2Àx first densify during heating and then the solid solution formation starts at about 200 K higher. The densification and the formation of the solid solution both occur at a lower temperature when p(O 2) increases. Based on this result, it is possible to better define the sintering atmosphere, eventually leading to optimized parameters such as density, oxygen stoichiometry and cations homogenization of nuclear ceramics and of a wide range of industrial ceramic materials.
Three different fuels UO2-only, UO2-Gd2O3(5%), and UO2-Gd2O3(10%) were produced by sol-gel technique. Their powder characteristics such as flowability, BET surface area, average pore diameter, and cumulative pore volume were determined. The pore size distributions of powders, green pellets, and sintered fuels were determined by using a mercury porosimeter. The theoretical densities of sintered fuels were found to be 98.01, 95.3, and 95.9%, respectively. Their ruggedness fractal dimensions were 1.111, 1.044, and 1.042, while the fractal dimensions associated with the size distribution of grains were 1.44, 1.58, and 1.60, respectively.
IOP Conference Series: Materials Science and Engineering
Sintering process is the final stage of fuel kernel manufacturing prior to the coating process. This stage is very important part of the whole process, because it will determine the feasibility of UO2 kernel to comply with the specifications of HTR reactor fuel. The objective of this research was to obtain UO2 kernel with the density of ≥ 95% TD. The results showed that the highest density reached 92.56% TD or about 10.1441 g / cm3. This condition of sintering was gained at the temperature of 1400 °C with sintering time of 2 hours. The sintering product diameter gained was around 919 μm, the specific surface area 4.4213 m 2 / g, and a total pore volume 4,751 x10-3 cm 3 / g. The density of UO2 kernel produced from this research is the best compared to previous finding because of its density already approaches the HTR fuel specification requirements
Uranium dioxide-gadolinium oxide fuel was produced by the sol-gel technique. The effects of different parameters such as calcination and reduction temperature, compaction pressure, particle size of powder, type of binder, sintering temperature, sintering atmosphere, and duration of sintering on pore size distribution were investigated. The experiments were carried out on three different fuels, (a) pure urania, (b) urania-gadolinia (10%), and (c) urania-gadolinia (10%)-titania (0.1%) doped fuel. It was observed that compaction pressure as low as 200 MPa is sufficient to obtain high-density pellets, while the use of binder or grinding the powder below 400 mesh does not affect densities. Reduction of powder at 1,000 K always gives lower density fuels than the powder reduced at 873 K. Sintering at high temperature and the use of a wet atmosphere each independently increases the fuel density.
Synthesis and Sintering of UN-UO2 Fuel Composites
Journal of Nuclear Materials, 2015
The design and development of an economical, accident tolerant fuel (ATF) for use in the current light water reactor (LWR) fleet is highly desirable for the future of nuclear power. Uranium mononitride has been identified as an alternative fuel with higher uranium density and thermal conductivity when compared to the benchmark, UO 2 , which could also provide significant economic benefits. However, UN by itself reacts with water at reactor operating temperatures. In order to reduce its reactivity, the addition of UO 2 to UN has been suggested. In order to avoid carbon impurities, UN was synthesized from elemental uranium using a hydride-dehydridenitride thermal synthesis route prior to mixing with up to 10 wt% UO 2 in a planetary ball mill.
Journal of Nuclear Materials, 2005
The dissolution of different mixed oxide (U, Th)O 2 particles under reducing conditions has been studied using a continuous flow-through reactor. The U/Th ratio seems to have no or little influence on the normalised leaching rate of thorium or uranium, The release rate of uranium from the outer surface of a Th rich matrix seems to follow the behaviour of pure UO 2 even though U is a minor component in these phases and the dissolution rate of Th is much lower. After long time U concentrations will become depleted at the solids surface and it will be expected that U release rates will become controlled by the release rates of thorium (rates at neutral pH < 10 À6 g m À2 d À1 ). Under reducing conditions, the matrix of HTR fuel particles presents significant intrinsic radionuclide confinement properties.
FY2015 Ceramic Fuels Development Annual Highlights
2015
Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2015 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY15 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems. Significant accomplishment include: • Collaborative work by LANL and JAEA staff identified the relationship between oxygen potential and stoichiometry for 20 and 30% Ce compositions as a function of temperature. This facilitated subsequent analysis of the thermal conductivity of these compositions as a function of off stoichiometry. These effects of Ce on the thermochemistry and thermophysical properties of oxide fuel represent a common interest to both programs given the utility of Ce as a surrogate for Pu as well as its importance as a fission product. This data will help improve understanding of fuel behavior under irradiation. These experiments also provide insight and guidance for development of advanced fuel systems that can have improved performance under accident conditions. • Synthesis routes were assessed for uranium bearing boride and silicide feedstocks for ATF composite concepts using, as a starting material, either UF 6 or some intermediate in the existing commercial UO 2 manufacturing process. Of particular interest were the compounds UB 2 , UB 4 , U 3 Si 2 , and U 3 Si 5. This is desirable for new accident tolerant fuels to leverage as much existing industry expertise and infrastructure as possible. A survey of the existing methods reported for producing these feedstock materials revealed no direct route from UF 6. Several new synthesis routes for uranium borides and uranium silicides were proposed. Some attractive processes begin with UO 2 F 2 , a compound already prepared during UO 2 manufacture and/or converting UF 6 to UF 4 by reaction with H 2. • Three different carbothermic reduction/nitridization conversion routes were studied and a refined process was identified. The nominal CTR-N route has been utilized to synthesize bulk UN, however, the precise time-temperature-atmosphere profiles, as well as variations in feedstock, have undergone minimal optimization over the years. The preferred conversion route from this study is believed to be more amenable to commercial production of uranium nitride feedstock. • Measurement of the coefficient of thermal expansion, specific heat capacity, and thermal diffusivity was performed as a function of temperature and composition in order to characterize the thermal conductivity of UN/U 3 Si 5 and UN/U 3 Si 2 composites. Determination of these material properties was necessary for final design of the LANL-1 and Westinghouse FOA ATR irradiation tests. • In-situ FLASH sintering experiments were performed on uranium dioxide at Brookhaven National Laboratory using the X-ray Powder Diffraction beamline (XPD) at the new
FY2014 Ceramic Fuels Development Annual Highlights
2014
Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2015 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY15 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems. Significant accomplishment include: • Collaborative work by LANL and JAEA staff identified the relationship between oxygen potential and stoichiometry for 20 and 30% Ce compositions as a function of temperature. This facilitated subsequent analysis of the thermal conductivity of these compositions as a function of off stoichiometry. These effects of Ce on the thermochemistry and thermophysical properties of oxide fuel represent a common interest to both programs given the utility of Ce as a surrogate for Pu as well as its importance as a fission product. This data will help improve understanding of fuel behavior under irradiation. These experiments also provide insight and guidance for development of advanced fuel systems that can have improved performance under accident conditions. • Synthesis routes were assessed for uranium bearing boride and silicide feedstocks for ATF composite concepts using, as a starting material, either UF 6 or some intermediate in the existing commercial UO 2 manufacturing process. Of particular interest were the compounds UB 2 , UB 4 , U 3 Si 2 , and U 3 Si 5. This is desirable for new accident tolerant fuels to leverage as much existing industry expertise and infrastructure as possible. A survey of the existing methods reported for producing these feedstock materials revealed no direct route from UF 6. Several new synthesis routes for uranium borides and uranium silicides were proposed. Some attractive processes begin with UO 2 F 2 , a compound already prepared during UO 2 manufacture and/or converting UF 6 to UF 4 by reaction with H 2. • Three different carbothermic reduction/nitridization conversion routes were studied and a refined process was identified. The nominal CTR-N route has been utilized to synthesize bulk UN, however, the precise time-temperature-atmosphere profiles, as well as variations in feedstock, have undergone minimal optimization over the years. The preferred conversion route from this study is believed to be more amenable to commercial production of uranium nitride feedstock. • Measurement of the coefficient of thermal expansion, specific heat capacity, and thermal diffusivity was performed as a function of temperature and composition in order to characterize the thermal conductivity of UN/U 3 Si 5 and UN/U 3 Si 2 composites. Determination of these material properties was necessary for final design of the LANL-1 and Westinghouse FOA ATR irradiation tests. • In-situ FLASH sintering experiments were performed on uranium dioxide at Brookhaven National Laboratory using the X-ray Powder Diffraction beamline (XPD) at the new