Experimental investigation of the conditions of the reduction and precipitation of uranium by minerals

The article gives the results of an experimental investigation of the processes of reduction and precipitation of uranium by certain minerals widely distributed in hydrothermal uranium deposits. It is assumed that iron, sulphur, and arsenic, which are included in the composition of these minerals, reduced U(VI) present in hydrother-mal solutions, as a result of which primary uranium minerals were deposited. The experimental investigation of these processes, which has been almost uninvestigated hitherto, is of great practical importance because it extends present representations of the physicochemical conditions favorable to the localization of uranium mineralization.The experimental data obtained indicates that as a result of the reduction of U (VI) to U(IV) in acid solutions at temperature of 100–350° C, crystalline uraninite, collomorphic uranium pitch, and sooty uranium blacks are formed. The character of the accessory minerals depends on the composition of the mineral-precipitant, the reaction temperature and in some cases, the initial uranium concentration in the solution.The results of the experiments confirm the working possibility of the formation of primary uranium minerals under natural conditions as a result of the reaction of solutions containing U (VI) with ore and vein minerals.     

Oxygen mass transport effects during hydrogen reduction of UO2 + x

Reduction of UO_{2 + x} was studied in a molecular beam system, in which a collisionless flux of atomic hydrogen strikes a portion of the surface of the solid and the reaction product water is detected by an in-situ mass spectrometer. The experiment revealed kinetic limitations due to both surface chemical reaction and to transport processes, the latter including surface diffusion of oxygen as well as bulk diffusion. A quantitative model of the reduction process showed that the surface reaction rate was linear in the H atom beam intensity and in the excess O/U ratio above a minimum value for detectable reaction. The bulk diffusion coefficient of oxygen determined from the data at 585°C agreed well with the measurement of Lay. The surface diffusion coefficient at the same temperature was found to be ~1 cm²/s, suggesting ideal two-dimensional-gas behavior of oxygen on the urania surface.     

注凝成型制备UO2陶瓷燃料核芯

高温气冷堆采用UO2微球作为燃料核芯,目前的主要制备方法采用溶胶凝胶工艺.为简化工艺流程、减少废液量,本工作研究采用注凝成型工艺制备UO2陶瓷微球.研究表明该工艺具有工艺简单、废液量少等优点.分析了溶胶凝胶和注凝工艺过程中的化学变化,研究了影响陶瓷微球直径的因素.采用该工艺制备出的UO2微球平均直径为710 μm,n(O)/n(U)≤2.01,密度为10.70 g/cm3.     

Control of UO2 microstructure by oxidative sintering

Sintering of UO₂ under oxidative conditions enables more easily than conventional sintering the control of the density, of the open porosity and of the grain structure. Coarsening and spheroidization as well as closure of pores are obtained by “solarization”. Large grains and desired grain size distributions are realized by “controlled path sintering” related to the U-O phases in the hyperstoichiometric range of the U-O phase diagram.     

New model of equiaxial growth of uranium dioxide grains under irradiation conditions

Conclusions The experimental data give evidence of a retardation in the growth of equiaxial grains when oxide fuel is irradiated. This retardation in grain growth can be explained as being due to the formation of short-lived defect structures on the grain boundaries under the action of irradiation. An analysis of a broad spectrum of experimental data indicates that a model based on this hypothesis correctly predicts the growth of equiaxial grains in a wide range of temperatures, irradiation intensities, and fuel depletions. The work was carried out with the partial support of the Russia Fund for Fundamental Research under grant number 96-02-18686. State Scientific Center of the Russian Federation, Institute of Innovative and Thermonuclear Research, Troitsk. Translated from Atomnaya énergiya, Vol. 84, No. 4, pp. 329–334, April, 1998.     

Modeling of fission gas swelling in LWR UO2 fuel under irradiation

An understanding of the behavior of fission gas in uranium dioxide (UO₂) fuel is necessary for the prediction of the performance of fuel rods under irradiation. A mechanistic model for matrix swelling by the fission gas in LWR UO₂ fuel is presented. The model takes into account intragranular and intergranular fission gas bubbles behavior as a function of irradiation time, temperature, fission rate and burn-up. The intragranular bubbles are assumed to be nucleated along the track of fission fragments, which play the dual role of creator and destroyer of intragranular bubbles. The intergranular bubble nuclei is produced until such time that a gas atom is more likely to be captured by an existing nucleus than to meet another gas atom and form a new nucleus. The capability of this model was validated by a comparison with the measured data of fission gas behavior such as intragranular bubble size, bubble density and total fuel swelling. It was found that the calculated intragranular bubble size and density are in reasonable agreement with the measured results in a broad range of average fuel burn-ups 6–83 GW d/tU. Especially, the model correctly predicts the fuel swelling up to a burn-up of about 70 GW d/tU.     

Effects of UO2/Zircaloy-4 Mole Ratios on Reaction of UO2 Fuel with Molten Zircaloy-4

Dissolution of UO₂ crucibles by molten Zircaloy-4 (Zry) was investigated in the temperature range of 2,223-2,373 K and for specimens having UO₂/Zry mole ratios between 7 and 18.2. The uranium concentration in the Zry melt rapidly increased during initial reaction time and approached saturated values, depending on reaction temperature and UO₂/Zry mole ratio. Kinetics of uranium concentration increase in the melt was analyzed based on a natural convection mass transfer model that takes into account the change of contact surface area/melt volume ratio with reaction time. The saturated uranium concentration in the Zry melt was inversely proportional to the U0₂/Zry mole ratio. An empirical correlation of saturated uranium concentration in the Zry melt was obtained as a function of UO₂/Zry mole ratios and reaction temperature. This study of the empirical correlation was intended to estimate maximum UO₂ fuel dissolution by molten Zry cladding during severe fuel damage accidents for three different reactor type fuels.     

Kinetics of high-temperature reaction of UO2 with zirconium

Institute of Atomic Energy, Academy of Sciences of the USSR. Translated from Atomnaya Énergiya, Vol. 70, No. 2, pp. 127–128, February, 1991.     

Behavior of unirradiated Zr based uranium metal fuel under reactivity initiated accident conditions

Reactivity initiated accident (RIA) tests with 4 unirradiated Zr based uranium metal fuel rods were performed to establish a criterion which should be observed under RIA conditions. Of the four tests, fuel failures were observed in the two tests that experienced the maximum energy depositions of 188 and 212 cal/g, respectively. However, the fuel failures were not observed at the place of a maximum energy deposition but at the position where the thermocouples were installed; one failed at the position whose local energy deposition was 150 cal/g, and the other one at the place with energy deposition of 170 cal/g. The fuel failures seem to have occurred because excessive pressure, which was caused by the partial melting of the fuel meat, was applied to the cladding with a reduced thickness. However, other parts of the fuel rods including the place of a maximum energy deposition maintained their integrity and a big change in the temperature and pressure in the internal capsule, which would be an indication of the fragmentation and dispersion of the fuel meat into the internal capsule, was not observed. Visual inspection also showed that, except for the thermocouple positions, there was no trace of clad failure such as the formation of brittle cracks in the cladding or melting of the cladding. Therefore, for the Zr based uranium metal fuel rods, it can be concluded that the threshold energy deposition above which fragmentation and dispersion of fuel meat into the primary coolant system is expected to occur could be higher than 212 cal/g.     

Pyrochemical reduction of uranium dioxide and plutonium dioxide by lithium metal

The lithium reduction process has been developed to apply a pyrochemical recycle process for oxide fuels. This process uses lithium metal as a reductant to convert oxides of actinide elements to metal. Lithium oxide generated in the reduction would be dissolved in a molten lithium chloride bath to enhance reduction. In this work, the solubility of Li₂O in LiCl was measured to be 8.8 wt% at 650 °C. Uranium dioxide was reduced by Li with no intermediate products and formed porous metal. Plutonium dioxide including 3% of americium dioxide was also reduced and formed molten metal. Reduction of PuO₂ to metal also occurred even when the concentration of lithium oxide was just under saturation. This result indicates that the reduction proceeds more easily than the prediction based on the Gibbs free energy of formation. Americium dioxide was also reduced at 1.8 wt% lithium oxide, but was hardly reduced at 8.8 wt%.