Determination of oxygen in ternary uranium oxides by a gravimetric alkaline earth addition method

The applicability of a gravimetric method based on alkaline earth metal addition for the determination of oxygen in ternary uranium oxides of the type M—U—O (M=La, Ce and Th) is described. The oxide sample is mixed with MgO or Ba_2.8UO_5.8 and heated in air under suitable conditions. Because uranium is completely oxidized to the hexavalent state during the reaction, oxygen can be determined from the weight change. Oxygen in La_yU_1-y O_2+x is determined up to y = 0.8 with a standard deviation for x of ±0.006 with MgO. For Th_yU_1-y O_2+x, the value of x is determined with Ba_2.8UO_5.8 with a standard deviation of ±0.01 at y = 0.8. For Ce_yU_1-y O_2+x, the method can be applied only for low cerium concentrations where y = 0–0.2; the value for x with Ba_2.8UO_5.8 at y = 0.2 showed a standard deviation of ±0.002.     

Oxidation state and coordination environment of uranium in sodium iron aluminophosphate glasses

An analysis of the X-ray absorption near edge structure (XANES) and the extended X-ray absorption fine structure (EXAFS) of uranium determined the oxidation state and coordination environment of uranium atoms in glasses containing 40 mol % Na₂O, 10 mol % Al₂O₃, 10 mol % Fe₂O₃, and 40 mol % P₂O₅ to which uranium oxides were added to a concentration of 50 wt % (above 100%). If the added amount of UO₂ was small, uranium occurred as U(IV) in a near-octahedral oxygen environment with an average U–O distance in the first coordination sphere of 2.25 Å. At higher concentrations of uranium oxides introduced both as UO₂ and as UO₃, uranium occurred as U(V) and U(VI); the first coordination sphere is split; shorter (~1.7–1.8 Å) and longer (2.2–2.3 Å) distances were observed, which corresponded to the axial and equatorial U–O bonds in uranyl ions, respectively; and the redox equilibrium shifted toward U(VI). The glass with the maximal (~33 wt %) UO₃ concentration contained mainly U(VI). The existence of low-valence uranium species can be related to the presence of Fe(II) in glasses. The second coordination sphere of uranium manifests itself only at high concentrations of uranium oxides.     

Characterization of new catalysts based on uranium oxides

Catalysts based on uranium oxides were systematically studied for the first time. Catalysts containing various amounts of uranium oxides (5 and 15%) supported on alumina and mixed Ni-U/Al₂O₃ catalysts were synthesized. The uranium oxide catalysts were characterized using the thermal desorption of argon, the low-temperature adsorption of nitrogen, X-ray diffraction analysis, and temperature-programmed reduction with hydrogen and CO. The effects of composition, preparation conditions, and thermal treatment on physicochemical properties and catalytic activity in the reactions of methane and butane oxidation, the steam and carbon dioxide reforming of methane, and the partial oxidation of methane were studied. It was found that a catalyst containing 5% U on alumina calcined at 1000°C was most active in the reaction of high-temperature methane oxidation. For the Ni-U/Al₂O₃ catalysts containing various uranium amounts (from 0 to 30%), the introduction of uranium as a catalyst constituent considerably increased the catalytic activity in methane steam reforming and partial oxidation.     

Consecutive recovery of uranium oxides from uranyl-containing molten alkali metal sulfate electrolytes

The oxygen coefficient (atomic ratio O/U) of a cathode product and the cathode current yield in consecutive recovery of uranium oxides from molten salt mixtures (nLi₂SO₄ + (1 ? n)Cs₂SO₄) + UO₂SO₄ and (Li,Na,K)₂SO₄ + UO₂SO₄ in air were analyzed as dependent on the electrolyte and the mean solvent-salt cation radius. The increasing mean solvent-salt cation radius and decreasing UO₂SO₄ concentration in the electrolyte during long-term electrolysis suppress the solvolysis of uranyl ions and decrease the oxygen coefficient in the cathode product. The cathode current yield decreases systematically during electrolysis and drops to zero long before the uranium is completely recovered from the melt. The ultimate uranium recovery increases as the mean solvent-salt cation radius increases.     

Nombre de transport et conductivite cationique dans les oxydes de calcium et de strontium

The cationic transference numbers have been determined for alkaline earth oxides using sample linear expansion measurements coupled with coulometric titration.Single crystal CaO has been studied in the temperature range 1200–1450°C, under oxygen partial pressures varying from 1 to 10^?5 atm. Under air, t_Ca at 1300°C is close to 0.02.Polycrystalline samples of SrO, under air, were tested from 1100°C to 1350°C, and showed a t_Sr value of 0.012 at 1300°C.The total conductivity of CaO was measured to obtain a value for the cationic conductivity. This value is compared with the data obtained from the self diffusion coefficient of Ca in CaO.     

Implication of volume changes in uranium oxides: A density functional study

In severe nuclear accident scenarios (in air environments and high temperatures) UO₂ fuel pellets oxidise to produce uranium oxides with higher oxygen content, e.g., U₄O₉ or U₃O₈. As a first step in investigating the microstructural changes following UO₂ oxidation to hexagonal high temperature phase of U₃O₈, density functional quantum mechanical calculations of the structure, elastic properties and electronic structure of U₃O₈ have been performed. The calculated properties of hexagonal phase of U₃O₈ are compared to those of the orthorhombic pseudo-hexagonal phase which is stable at room temperature. The total energy technique based on the local density approximation plus Hubbard U as implemented in the CASTEP code is used to investigate changes in the lattice constants. The first-principles calculations predict a 35–42% increase in volume per uranium atom as a result of the transformation from UO₂ to U₃O₈, in agreement with experimental data. The implications of this prediction on the linear expansion and fragmentation of fuel are discussed.     

Near infrared reflectance spectroscopy as a process signature in uranium oxides

Near-infrared (NIR) reflectance spectroscopy was examined as a potential tool for the determination of forensic signatures indicative of the chemical process history of uranium oxides. The ability to determine the process history of nuclear materials is a desired, but underdeveloped, area of technical nuclear forensics. Application of the NIR technique potentially offers a quick and non-destructive tool to serve this need; however, few data have been published on the compounds of interest. The viability of NIR was investigated through the analysis of a combination of laboratory-derived and real-world uranium precipitates and oxides. A set of reference uranium materials was synthesized in the laboratory using the commonly encountered aqueous precipitation reactions for uranium ore concentration and chemical separation processes (ammonia, hydrogen peroxide, sodium hydroxide, ammonium carbonate, and magnesia). NIR spectra were taken on a range of samples heat treated in air between 85 and 750 °C. X-ray diffraction patterns were also obtained to complement the NIR analysis with crystal phase information. Similar analyses were performed using a set of real-world samples, with process information obtained from the literature, to provide a comparison between materials synthesized in the laboratory and samples representative of industrial processes.     

Reaction of uranium monocarbide powder in oxidizing atmospheres

Thermogravimetric records have been obtained of uranium monocarbide oxidized in oxygen and in carbon dioxide under similar conditions. It is shown that the reactions are very similar. Initially, free carbon is formed during oxidation as well as UO₃ in O₂ or UO₂ in CO₂. In a second reaction step non-stoichiometric carbonates are formed, depending on experimental conditions. The carbonates decompose to stoichiometric oxides in the next step.Well crystallized α-UO₃ can be obtained by oxidation of uranium monocarbide in oxygen. The solid products containing UO₂ which slowly formed in carbon dioxide, are unstable in air.Infrared and X-ray analysis have been used to-compare related, solid structures. Activation energies have been determined of non-isothermal reactions, recorded below 750°C.     

Defect chemistry of uranium oxides

Uranium oxides are known as nonstoichiometric compounds whose composition changes according to external conditions such as temperature and oxygen partial pressure. The change of composition caused by the formation of defect structure results in a change of their properties. In this paper, the compositional changes of UO₂ and doped UO₂ [(U, M)O₂; M=La, Ti, Pu, Th, Nb, Cr, etc.] and also those of other uranium oxides (U₄O₉, U₃O₈) are shown against oxygen partial pressure. From the results of doped UO₂, it is concluded that the valence control rule holds to a first approximation. The defect structures are estimated both from log x vs. log Po₂ (x: deviation from the stoichiometric composition and Po₂: oxygen partial pressure) and log vs. log Po₂ (: electrical conductivity) relations. The defect structures of UO₂ and doped UO₂ are derived based on the Willis model for UO_2+x. The detect structure of U₄O₉ phase is similar to that of UO_2+x, but the defect structures of U₃O₈ phase are complicated due to the existence of many higher-order phase transitions. The thermodynamic data such as the partial molar enthalpy and entropy and the heat capacity are important to characterize the defect structure. The high temperature heat capacities of UO₂ doped with Gd show pronounced increases at high temperatures the onset temperature decreases as the dopant content increases. The increase of heat capacity is interpreted to be due to the formation of lattice defects. The heat capacity measurements on U₄O₉ and U₃O₈ clucidate the presence of the phase transition. The mechanisms of these phase transitions are discussed.     

Effects of oxidant and particle size on uranium leaching from coal ash

In this paper, uranium leaching efficiency from the coal ash was 45.82% on the condition of 0.3 M H₂SO₄, 1:20 g/L, 60 °C and 12 h. According to the result of Sequential chemical extraction, leaching trials and X-ray diffraction, oxidants promoted the uranium leaching with the generation of baratovite. There were differences between uranium content of leachable states and leaching trials. Although the uranium content bounding with exchangeable, carbonates and Fe–Mn oxides all peaked at particle size fraction 75–150 μm, the highest uranium content in leaching trials peaked at size fraction 48–75 μm.     

Thermal and kinetic analysis of uranium salts

Thermal decomposition of U(C₂O₄)₂·6H₂O was studied using TG method in nitrogen, air, and oxygen atmospheres. The decomposition proceeded in five stages. The first three stages were dehydration reactions and corresponded to removal of four, one, and one mole water, respectively. Anhydrous salt decomposed to oxide products in two stages. The decomposition products in nitrogen atmosphere were different from those in air and oxygen atmospheres. In nitrogen atmosphere UO_1.5(CO₃)_0.5 was the first product and U₂O₅ was the second product, while these in air and oxygen atmospheres were UO(CO₃) and UO₃, respectively. The second decomposition products were not stable and converted to stable oxides (nitrogen: UO₂, air–oxygen: U₃O₈). The kinetics of each reaction was investigated with using Kissinger–Akahira–Sunose and Flynn–Wall–Ozawa methods. These methods were combined with modeling equations for thermodynamic functions, the effective models were investigated and thermodynamic values were calculated.     

The melting behaviour of uranium/neptunium mixed oxides

The melting behaviour in the pseudo-binary system (UO₂ + NpO₂) has been studied experimentally for the first time in this work with the help of laser heating under controlled atmosphere. It has been observed that the solidus and liquidus lines of this system follow an ideal solution behaviour (negligible mixing enthalpy) between the well-established solid/liquid transition temperatures of pure UO₂ (3130 K) and that recently assessed for NpO₂ (T = 3070 K). Pre- and post-melting material characterizations performed with the help of X-ray diffraction and Raman spectroscopy are also consistent with ideal mixing of the two end members. Such behaviour follows the similar structure and bonding properties of tetravalent uranium and neptunium and the similar melting points of the two oxides. The interest of this investigation is twofold. From a technological viewpoint, it indicates that the incorporation of NpO₂ in UO₂ fuel or transmutation targets is a viable option to recycle neptunium without inducing any relevant change in the chemical or thermal stability of the uranium dioxide matrix, even up to the melting point. From a more fundamental perspective, it confirms that actinide dioxides, and particularly UO₂, tend to mix in a way closer to ideal, the closer are the atomic numbers, 5-f electron shell filling, atomic radii and oxygen potentials of the metals forming the pure dioxides.     

Thermal and kinetic analysis of uranium salts

The thermal decomposition kinetics of UO₂C₂O₄·3H₂O were studied by TG method in a flowing nitrogen, air, and oxygen atmospheres. It is found that UO₂C₂O₄·3H₂O decomposes to uranium oxides in four stages in all atmosphere. The first two stages are the same in the whole atmosphere that correspond to dehydration reactions. The last two stages correspond to decomposition reactions. Final decomposition products are determined with X-Ray powder diffraction method. Decomposition mechanisms are different in nitrogen atmosphere from air and oxygen atmosphere. The activation energies of all reactions were calculated by model-free (KAS and FWO) methods. For investigation of reaction models, 13 kinetic model equations were tested and correct models, giving the highest linear regression, lowest standard deviation, and agreement of activation energy value to those obtained from KAS and FWO equations were found. The optimized value of activation energy and Arrhenius factor were calculated with the best model equation. Using these values, thermodynamic functions (??H ^*, ??S ^*, and ??G ^*) were calculated.     

Continuous recovery of uranium oxides by electrolysis of M2MoO4-M2Mo2O7-UO2MoO4 melts (M = Li, Na, K, Cs)

Effect of the electrolyte composition and of the solvent-salt cation on the oxygen coefficient of the cathodic product (O/U atomic ratio) and basic characteristics of the potentiostatic electrodeposition of uranium dioxide in prolonged recovery of uranium oxides from electrolytes of the system M₂MoO₄-M₂Mo₂O₇-UO₂MoO₄ Melts (M = Li, Na, K, Cs) in air was analyzed. A decrease in the UO₂MoO₄ concentration and accumulation of M₂Mo₂O₇ in the electrolyte in the course of a prolonged electrolysis suppress the solvolysis of uranyl ions and make lower the oxygen coefficient of the cathodic product. Li₂MoO₄-based melts possessing pronounced oxygenacceptor properties exhibit an anomalous behavior in these experiments. The current efficiency, initial current density, and deposition rate of the product decrease as electrolytes are depleted of uranium. In discussions of numerical data, it is necessary to take into account the formation of lower valence forms of uranium due to the chemical corrosion of the cathodic product, and in the case of melts of the lithium system, the additional cathodic process in which the solvent is reduced.     

Ab initio study on the reaction of uranium with oxygen

The primary reaction products and reaction mechanism of uranium with oxygen were discussed from MP2 method with the relativistic core potential of SDD basis set for U and 6-311+G* for O. The molecular geometries, electronic structure and energies of uranium oxides were obtained. The inspection on the three-dimensional potential energy surfaces of the U–O₂ interaction suggested that the abstraction and insertion mechanism were responsible for the studied reactions. The abstraction reaction channel resulting in the formation of UO and O is favored because the energy barrier is remarkably smaller than the one of the insertion channel resulting in the linear OUO product directly.     

Hydrous oxides in the concentration and removal of fission products from alkaline solutions of uranium

A great variety in retention properties occurs as a result of different methods of preparation of the sorbents. Specific surfaces and porosities, which are mostly connected with the sorption activities, may vary widely. The activation of Al₂O₃, SnO₂ and silica gel, to produce highly active sorbents occurs only if the oxide is contacted with acid immediately after thermal treatment. The efficiency of the separation scheme has been tested using uranium and fission products under static conditions from strongly alkaline aqueous solutions.     

Dosage de l'oxygene dans l'uranium et ses composes par la methode au monochlorure de soufre

A sulfur monochloride method is proposed for the determination of oxygen in uranium compounds. Sulfur monochloride reacts with oxygenated compounds at temperatures depending upon their nature; the sulfur dioxide produced is titrated by iodometry, after the excess reagent has been eliminated by a selective adsorption-desorption process using activated charcoal. This method has been successfully applied to uranium oxides (UO₂, U₃O₈), to mixtures of uranium dioxide with uranium, uranium nitride, and uranium carbide, and to substituted carbides (UC_1-xO_x). The results are generally satisfactory for oxygen contents higher than 500 p.p.m. However, in the presence of free or combined carbon, this limit is considerably higher. A loss of oxygen as carbon monoxide is also possible, and a simultaneous determination of carbon monoxide must be carried out. The relative error is of the order of a few per cent.     

Determination of oxygen to uranium ratio in irradiated uranium dioxide by controlled potential coulometry

This work reports the determination of oxygen to uranium (O/U) ratio in irradiated UO_2+x fuel pellet of burnup of ca. 34 GWd/t by controlled potential coulometry. The method is based on the dissolution of the nuclear fuel in strong phosphoric acid (SPA) at 180–190 °C under an inert atmosphere. After dissolution, 8% sulphuric acid is added in order to obtain a 20% SPA in 8% sulphuric acid. A controlled potential coulometric determination of uranium(VI) is carried out at ?0.60V vs. ferri-ferrocyanide. The uranium(IV) contained in an aliquot of the fuel solution is oxidised to uranium(VI) with cerium(IV) sulphate, and the total uranium content is then determined by coulometry. Optimum experimental conditions have been established using simulated irradiated fuel solution containing various fission products which include cerium, tellurium, palladium, ruthenium, molybdenum and zirconium. Interference of the fission products and the possible removal of their interferences by preelectrolysis at +0.5 V vs. saturated calomel electrode (SCE) have been investigated. The accuracy of the coulometric method is confimed by polarographic measurement using several unirradiated UO_2+x fuel of known stoichiometry.     

Analysis of uranium in dissolver solution of fast reactor carbide fuel reprocessing

Uranium in simulated dissolver solution of FBTR mixed carbide fuel containing fission products is analyzed by modified Davies–Gray method. The uranium is analyzed after the sample is evaporated with 1 M H₂SO₄ and followed all the steps carried out by Davies–Gray method. The proposed method assures analysis of uranium in dissolver solution of FBTR fuel can be titrated directly without prior separation of fission products.     

Effect of cationic composition of electrolytes based on the M2MoO4-M2Mo2O7-UO2MoO4 (M = Li, Na, K, Cs) system on the oxygen factor of the uranium electrolytic oxides

The effects of the electrolyte composition, deposition potential, temperature, and the salt-solvent cation on the oxygen factor (the [O]/[U] atomic ratio) of uranium oxides obtained by potentiostatic electrolysis of molybdate melts are studied. It is shown that the oxygen factor of the cathodic product increases with the melt temperature. The shift of the deposition potential toward more electronegative values, an increase in the Mo₂O^2? ₇ ion concentration, and a decrease in the UO₂MoO₄ concentration favor the cathodic formation of uranium oxides with lower oxygen factors. All other conditions being the same, the oxygen factor decreases with increasing the salt-solvent cation radius. The observed anomalous behavior of the Li₂MoO₄-based electrolytes can be caused by a lower activity in the melts of the O^2? anions that are capable of forming stable complexes Li₃O⁺ with lithium cations. The obtained results can be reasonably explained in terms of the model of ionic composition of the uranyl-containing tungstate melts, which is based on the concept of involvement of the uranyl ions in the complexes’ formation and stepwise solvolysis.