Diffusion Coefficient Of Uranium Research Articles

Investigation of diffusion and binding properties of uranium in the diffusive gradients in thin-films technique

Environmental context Monitoring uranium concentrations and speciation in aquatic systems is important for pollution control and for environmental studies. Although an in situ speciation technique based on diffusion and uptake of uranium has been developed, known as DGT, there were uncertainties over some parameters affecting the accuracy of the measurements. This study resolved those uncertainties by investigating diffusion and binding properties of uranium in DGT and provided confidence in monitoring uranium in the environment. Rationale Diffusive gradients in thin-films (DGT) ha been used for uranium speciation measurements since 2006, but interpretational difficulties have arisen due to uncertainties in the pH dependence of U diffusion coefficients (D) within the diffusive gel. This work tested the hypothesis that differences in measured D values between laboratories are due to methodological artifacts. Methodology The properties of uranium binding and diffusion of U within DGT were systematically investigated at different conditions between pH 3–8 and ionic strength from 1 to 500 mmol L−1. Previous uncertainties were attributed to loss of U by adsorption within the diffusion cell, identified by mass balance calculations, and a binding efficiency dependent on the type of binding gel. Results The diffusion coefficient in an agarose cross-linked polyacrylamide gel measured using Metsorb DGT of 5.26 ± 0.17 × 10−6 cm2 s−1 agreed well with both values obtained using a pre-conditioned diffusion cell (< 7% difference) and by modelling. The binding efficiencies for Chelex DGT (BChelex = 0.86–0.89) and Fe-oxide DGT (BFe-oxide = 0.64 were smaller than for Metsorb DGT (BMetsorb = 1). No significant differences were found for BChelex at different pH. Discussion This work demonstrated that the diffusion coefficient of uranium is independent of pH and the binding gel used, provided the binding efficiency (B) is taken into account. These findings indicate that DGT equipped with either Metsorb or Chelex binding gels should be capable of measuring U for a wide range of environmental conditions provided the appropriate diffusion coefficients and binding efficiencies are used.

Determination of diffusion coefficients of uranium in liquid gallium by GITT

In the electrowinning or chemical reductive extraction process, those mass transfer kinetic parameters, for instance the diffusion coefficient in the liquid cathode or medium, have a strong impact upon the practical extraction rate and separation efficiency. This work aims to calculate the diffusion coefficient of uranium in liquid gallium with galvanostatic intermittent titration technique (GITT) and to deduce its corresponding activation energy. By analyzing the data of GITT, before U saturation in liquid gallium at the temperature range of 673-873 K, the temperature dependence of diffusion coefficients of uranium in liquid gallium was measured as ln D U Ga = − 6.90 ± 1.03 − 1.2 ± 0.8 × 10 3 T and the corresponding activation energy was determined as E a = −11.0 ± 6.7 kJ/mol. The possible source of errors to calculate the diffusion coefficients, such as reference electrode, small solubility of uranium in gallium, side reactions and geometrical approximation of cathode in the theory formula are also presented and discussed.

Kinetic Parameters and Diffusivity of Uranium in FLiNaK and ClLiK

Anodic stripping voltammetry (ASV) and cyclic voltammetry (CV) measurements at 773, 823, and 873 K were made of uranium trifluoride (UF3) in lithium fluoride-sodium fluoride-potassium fluoride eutectic (FLiNaK) and uranium trichloride (UCl3) in lithium chloride-potassium chloride eutectic (ClLiK). ASV data were used to estimate the charge transfer coefficients, exchange current densities, and activation energies of the uranium reactions. Charge transfer coefficients of both salt systems were within the range of 0.17 to 0.38. Exchange current densities in the fluoride and chloride salts were estimated within the range of 0.060 to 0.12 A cm−2. Activation energy of uranium exchange current was 16.0 kJ mol−1 in the fluoride salt and 28.9 kJ mol−1 in the chloride salt. Kinetics of charge transfer were found to be faster in FLiNaK. Analyses of the CV data suggest the electrochemical system was diffusion controlled and irreversible in the chloride and fluoride salt mixtures. Diffusion coefficients of uranium in the range of temperatures were on the order of 10−5 cm2 s−1 in both systems. Greater values of diffusivity in ClLiK are attributed to its lower density compared with FLiNaK. Activation energy of uranium diffusion in the fluoride and chloride salt mixtures were 53.4 and 89.4 kJ mol−1, respectively.

Influence of vibrational entropy on the concentrations of oxygen interstitial clusters and uranium vacancies in nonstoichiometric UO2

We combine density functional theory (DFT) formation energies and empirical potential calculations of vibrational free energies to calculate the free energies of formation of point defects and clusters of oxygen interstitials, and we use a dilute defect model to calculate the concentrations of defects as a function of temperature and composition. We find that at high temperature oxygen interstitials are dominant, either in isolated form or in clusters depending on the deviation from stoichiometry. At temperatures lower than 1300 K we predict uranium vacancies to be dominant in the stoichiometric material. The disorder in $\mathrm{U}{\mathrm{O}}_{2}$ therefore changes from Schottky to Frenkel type with increasing temperature. Uranium vacancies remain dominant up to deviations form stoichiometry as large as 0.045 at 800 K. Moreover, the concentration of uranium vacancies proves to be non-monotonous as a function of composition. These results are consistent with some experimental data on the evolution with stoichiometry of lattice constant, diffusion coefficients of uranium, positron lifetime and dilatometry measurements.

Evaluation and application of Diffusive Gradients in Thin Films (DGT) technique using Chelex®-100, Metsorb™ and Diphonix® binding phases in uranium mining environments

A new resin- Diphonix® in Diffusive Gradients in Thin Films (DGT) technique for the determination of uranium was investigated and compared with previously used binding phases for uranium, Chelex®-100 and Metsorb™. The DGT gel preparation and the elution procedure were optimized for the new resin. The U uptake on Diphonix® resin gel was 97.4 ± 1.5% (batch method; [U] = 20 μg L−1; 0.01 M NaNO3; pH = 7.0 ± 0.2). The optimal eluent was found to be 1 M 1-hydroxyethane-1, 1-diphosphonic acid (HEDPA) with an elution efficiency of 80 ± 4.2%. Laboratory DGT study on U accumulation using a DGT samplers with Diphonix® resin showed a very good performance across a wide range of pH (3–9) and ionic strength (0.001–0.7 M NaNO3). Diffusion coefficients of uranium at different pH were determined using both, a diffusion cell and the DGT time-series, demonstrating the necessity of the implementation of the effective diffusion coefficients into U-DGT calculations. Diphonix® resin gel exhibits greater U capacity than Chelex®-100 and Metsorb™ binding phase gels (a Diphonix® gel disc is not saturated, even with loading of 10.5 μmol U). Possible interferences with Ca2+ (up to 1.33 × 10−2 M), PO43− (up to 1.72 × 10−4 M), SO42− (up to 4.44 × 10−3 M) and HCO3− (up to 8.20 × 10−3 M) on U-DGT uptake ([U] = 20 μg L−1) were investigated. No effect or minor effect of Ca2+, PO43−, SO42−, and HCO3− on the quantitative measurement of U by Diphonix®-DGT was observed. The results of this study demonstrated the DGT technique with Diphonix® resin is a reliable and robust method for the measurement of labile uranium species under laboratory conditions.

Calculation of the Diffusion Coefficient of Uranium in Compacted Clay: Effect of the Temperature

Diffusion of Uranium in compacted clay was studied through the non-steady state diffusion method for the safety assessment of the radioactive waste in the storage and disposal practices. Since the permeability of the compacted clay is very low, the main mechanism for radionuclide transport is governed by the diffusion phenomenon. The diffusion process of uranium in compacted clay as a porous medium has been modeled by taking into account the effect of sorption. An increase in temperature, caused by the radiation from the canisters, takes place in the repository. Then, the effect of temperature on textural and structural properties of clay and coefficient diffusion has been examined. First, we examined the structure changes of the clay pellets under the influence of the heat in the temperature range 25–100°C. These changes are studied in the region of mesopores and macropores using the method of the mercury porosimetry. Second, the effect of temperature on the diffusion behavior of uranium was simulated. In these simulations, the experimental values of apparent diffusion coefficients of uranium were used to determine the concentration profiles versus time, depth and temperature. As conclusion, it was shown that the temperature plays an important role in the uranium transport through the compacted clay.

Kinetics of dissolution of stoichiometric uraninite

The kinetics of dissolution of stoichiometric uraninite (UO2) synthesized by electroreduction (electrolysis) of uranyl chloride in a melt of an eutectic mixture of alkali metal salts was studied. It follows from the results obtained that, in the initial step of uraninite dissolution under both dynamic and static conditions, schoepite is not an intermediate phase. It is formed, apparently, in the further steps by oxidation of uranium dioxide. The diffusion coefficients of uranium in solutions over uraninite and schoepite were determined. It was found that, at least in the initial step of the dissolution, the forming uranyl hydroxide occurs in the solution either in the dissociated state or in the form of mononuclear hydroxo complexes.

Laser ablation microprobe inductively coupled plasma mass spectrometry study on diffusion of uranium into cement materials

Abstract The diffusion of uranium (U(VI)) in solid cement monoliths of ordinary portland cement (OPC) and low-heat portland cement containing 30[wt.]% fly ash (FAC) was measured by an in-diffusion technique. Detailed sharp depth profiles of uranium in the solid cement matrices were successively and quantitatively measured using laser ablation microprobe inductively coupled plasma mass spectrometry (LAMP-ICP-MS), and the apparent (D a) and effective (D e) diffusion coefficient of uranium in cement matrix were calculated as: D a=∼4×10−16 m2 s−1 and D e=∼3×10−11 m2 s−1 for OPC, and D a=∼2×10−17 m2 s−1 and D e=∼6×10−13 m2 s−1 for FAC.

Low-temperature sintering mechanism on uranium dioxide

Based on a point defect model, the mechanism of low-temperature sintering of uranium dioxide was studied in this paper. The diffusion coefficient of uranium in UO2 + x, sintering temperature and densification equation in low-temperature sintering were analyzed by both the point defect model and low-temperature sintering experiments. The results showed that the diffusion activation energy of uranium in over-stoichiometric UO2 + x was lowered by 3.0 ev than that in stoichiometric UO2. And the diffusion coefficient of uranium in UO2 + x was proportional to x2. In addition, the theoretical low temperature sintering temperature was calculated to be in the range of 1089–1151 °C, which indicated that it was necessary to maintain proper over-stoichiometric oxygen for low-temperature sintering process. Moreover, the calculation results by the point defect model matched perfectly with the experiment results, PDM might be a good model to describe the relationships between defects concentration and atmosphere composition.

Fission enhanced diffusion of uranium in zirconia

This paper deals with the comparison between thermal and Fission Enhanced Diffusion (FED) of uranium into zirconia, representative of the inner face of cladding tubes. The experiments under irradiation are performed at the Institut Laue Langevin (ILL) in Grenoble using the Lohengrin spectrometer. A thin 235UO2 layer in direct contact with an oxidised zirconium foil is irradiated in the ILL high flux reactor. The fission product flux is about 1011ionscm−2s−1 and the target temperature is measured by an IR pyrometer. A model is proposed to deduce an apparent uranium diffusion coefficient in zirconia from the energy distribution broadening of two selected fission products. It is found to be equal to 10−15cm2s−1 at 480°C and compared to uranium thermal diffusion data in ZrO2 in the same pressure and temperature conditions. The FED results are analysed in comparison with literature data.

Thermal- and radiation-enhanced diffusion of uranium in oxidised zirconium

This paper is devoted to the study of the defect influence on uranium diffusion in zirconia in the context of nuclear waste disposal. The experiments in reactor conditions are performed at the Institut Laue Langevin in Grenoble using the Lohengrin spectrometer. A thin UO 2 layer in direct contact with a zirconium foil is irradiated in the ILL high-flux reactor. The fission product rate is around 3×10 11 ions s −1 and the neutron flux is equal to 5×10 14 n cm −2 s −1. The target temperature is controlled by an IR pyrometer and ranges from 470 to 490 °C. In these conditions, a zirconium oxidation is first observed before uranium diffusion. A model is proposed to deduce an apparent uranium diffusion coefficient in zirconia (ZrO 2) from the energy distribution broadening of a selected fission product ( A=90). It is found to be equal to 10 −15 cm 2 s −1. The study of thermal diffusion is performed by using ion beam techniques (ion implantation and Rutherford Backscattering). ZrO 2 samples are implanted with 800 keV uranium ions at a dose of 10 16 ions cm −2 and annealed at a pressure of 7.5×10 −1 Pa. No uranium diffusion could be observed up to 800 °C. The influence of irradiation defects mainly due to fission products, both on zirconium oxidation and on uranium diffusion, is clearly demonstrated.

Uranium-doped britholites Ca xLa y(SiO 4) 6− u(PO 4) uO t:U synthesis, characterization and preliminary study of uranium diffusion

Natural apatitic phosphosilicates, generally called “britholites”, were found to contain certain amounts of radioactive elements or fission products. Thus, britholites have been the object of a great interest because of their potential application as typical matrices for the conditioning of nuclear wastes. In the present report, uranium-doped oxy-silicophosphates Ca x La y (SiO 4) 6− u (PO 4) u O t /U were synthesized by high-temperature solid-state reaction. The prepared compounds were characterized by X-ray diffraction method, Raman diffusion spectroscopy and X-ray microprobe. The uranium solubility in such compounds was studied. The uranium thermal diffusion in the britholite with the theoretical formula Ca 5La 5(SiO 4) 3(PO 4) 3O 2 was estimated. The bands present in the Raman spectra were identified and assigned to the corresponding vibration modes of SiO 4 and PO 4 anions. The diffusion coefficient of uranium in the studied apatite was found close to that calculated for europium and lanthanum in some other apatites.

Zirconium surface modification under fission product irradiation. Application to nuclear fuel cladding tubes

Nuclear fuel devices of pressurised water reactors are composed of uranium oxide pellets which are enclosed in zircaloy-cylinders. During reactor operation in the contact with the fuel the zircaloy oxidises non-uniformly in depths of the order of some micrometers. Furthermore, energy deposition of fission recoils leads to sputtering of uranium onto the inner surface of the cladding material. Thus, sputtered uranium ions start to migrate outwards. Experiments were performed at Institut Laue Langevin (ILL) on the Lohengrin mass spectrometer in order to simulate the alteration of cladding tubes in contact with the nuclear fuel during irradiation. A thin UO 2 deposition in direct contact with a zirconium foil was put into the ILL high flux reactor (5×10 14 neutrons cm −2 s −1). The variation of energy loss of selected fission products in the foil put in evidence a zirconium oxidation whose kinetics constant is deduced from these experiments. Following oxidation the fission product mean kinetic energy is stabilised, but still a broadening of the energy distribution is observed, which is characteristic of uranium diffusion inside the zirconia target. Using Fick's model, a uranium diffusion coefficient in zirconia under irradiation is deduced.

Diffusion of Uranium in Compacted Bentonites in the Reducing Condition with Corrosion Products of Iron

AbstractIn high-level waste repository, a carbon steel overpack will be corroded after closure. This will create a reducing environment in the vicinity of the repository. Reducing conditions are expected to retard the migration of redox-sensitive radionuclides such as uranium.The apparent diffusion coefficients of uranium were measured in compacted bentonites (Kunigel VI® and Kunipia F®, JAPAN) in contact with carbon steel and its corrosion products under reducing conditions or without carbon steel under oxidizing conditions for comparison. The apparent diffusion coefficients measured were 10-12 to 10-14 m2/s under oxidizing conditions and 10-13 to 10-14 m2/s under reducing conditions. There were significant effects of redox conditions, dry density (0.8 to 2.0 g/cm3) and montmorillonite contents (60% for Kunigel VI or 100% for Kunipia F) on the apparent diffusion coefficients. The sorption of uranium on corrosion products of iron was smaller than that on montmorillonite. Montmorillonite density could be a good index to explain density dependence of the diffusion coefficients under the reducing conditions. Uranium would diffuse in free pore water with diffusion coefficients greater than 10-13 m2/s in low density bentonites under both redox conditions because the diffusion coefficients depended on sorption coefficients. Since diffusion coefficients were independent of sorption coefficients in high density bentonites where free pore water is scarce, surface diffusion might occur with coefficients were of about 10-13 m2/s.

Diffusion of uranium in titanium-doped uranium dioxide

The diffusion coefficients of 233U in (U 1− y Ti y )O 2+ x ( y=0, 0.001, 0.007) were measured in the oxygen partial pressure range from 10 −8 to 10 −2 Pa at 1673 and 1773 K by the α-energy degradation method. The ratios of the diffusion coefficient of uranium( D U) in titanium-doped UO 2+ x to those in undoped UO 2+ x were about 6 for y=0.007 at 1673 K and about 1 for y=0.001 at 1773 K. The higher values of D U in titanium-doped UO 2+ x than in undoped UO 2+ x at 1673 K cannot be explained by the substitutional model based on valency control rule assuming titanium(IV), but can be explained by the interstitial model. The ratio of the D U value in doped UO 2+ x to that in undoped UO 2+ x tended towards unity at higher temperatures, because the concentrations of cation vacancies formed by the thermal energy were larger than those formed by doping. It was found that the defect structure of hyperstoichiometric UO 2+ x can be explained by Lidiard's model.

Material property correlations for uranium mononitride: III. Transport properties

The experimental data for the tracer diffusion coefficients of nitrogen ( D N in UN ) and uranium ( D N in UN ) in uranium mononitride (UN), electrical resistivity (ρ), and thermal conductivity ( k) of UN have been collected and reviewed. Empirical correlations for these properties as functions of temperature ( T, in K), nitrogen partial pressure ( P N , in atm), grain size ( g, in μm), and porosity ( P, as a volume fraction) were developed and are given as: D N in UN(rmcm 2/s) = 2.252 × 10 −5P N 0.4134g −0737+2.179×10 −4T exp{ −19214.7 T } , D N in UN(cm 2/s) = 2.252 × 10 −11P N 0.6414 exp{ −7989.3 T } , ρ(μΩ cm) = 71.49 e 2.14P (T) 0.125 , k( W/ mK) = 1.864 e −2.14 P ( T) 0.361. The nitrogen diffusion coefficient correlation was constructed from experimental data in the range of 0.001 ≤ P N ≤ 2.0 atm, 25 ≤ g ≤ 5000 μm, and 1503 ≤ T ≤ 2273 K; similarly, the uranium diffusion coefficient correlation incorporated an experimental database in the range of 2.6×10 −5 ≤ P N ≤ 0.66 atm and 1873 ≤ T ≤ 2133 K. The electrical resistivity and thermal conductivity correlations are applicable for temperatures of 298 ≤ T ≤ 1600 K and 298 ≤ T ≤ 1923 K, respectively, and both correlations were developed from data representing porosities of 0–20%.

Anion exchange kinetics of uranium in sulphate media

Experiments have shown that the sorption of uranium from acidic sulphate solutions onto strong base anion exchange resins is particle diffusion controlled in the uranium concentration range 0.0001–0.004M. A simplified diffusion model, based on Fick's Law, fits the kinetic data at each concentration. The rate of sorption falls significantly at lowered solution concentration. This corresponds with a lowered equilibrium loading of uranium and can be predicted with the Nernst-Planck equations using the measured self diffusion coefficient of uranium (1.65 × 10 −8 cm 2/s) and sulphate ions. The importance of this lowering of uranium sorption on the design of ion exchange equipment is stressed.

High temperature diffusion of uranium in refractory metals determined by thermal neutron fission of 235U

The diffusion of uranium in refractory metals, such as Ta and W, has been determined by counting the number of thermal neutron fission tracks of the diffused uranium recorded on a mica detector. The diffusing medium was made of a stack of ribbon filaments mounted on a fixture, and one side of the sample was doped with uranium atoms. A constant dc current was run through the filaments to obtain the required temperature by resistance heating for the diffusion. All processes were carried out in a vacuum bell-jar. The diffusion coefficients of uranium in tantalum at temperatures in the range 1013–2257°C follow the relation D = 1.03 × 10 −6exp(−28 000/ RT), and in tungsten at temperatures in the range 2134–2335°C follow the relation D = 3.34 × 10 −4exp(−61 900/ RT).

The sintering of grain boundary cavities in uranium dioxide

The final stage sintering of uranium dioxide has been investigated in the temperature range 1623–1973 K where the residual porosity is situated mainly on grain boundaries. The shape of the sintering curve is quantitatively predicted from a knowledge of the cavity distribution and from the sintering rate of an individual cavity described by the Hull and Rimmer (1959) equation. The effective grain-boundary diffusion coefficient is estimated from these curves and is shown to be in reasonable agreement with the grain-boundary diffusion coefficient of uranium measured by radio-tracer methods. The effect of entrapped gas appears to be of negligible importance since cavities were observed to sinter to closure; neither did grain growth strongly affect sintering rates, since cavities remained on grain boundaries.

Contribution a l'etude des mecanismes du debut de frittage du bioxyde d'uranium

Sintering kinetics of UO 2 compacts (using three different types of powder) were studied by recording their isothermal shrinkage under hydrogen between 700 and 1400 °C in a high temperature dilatometer. A constant activation energy of 95 kcal/mole was found all over this temperature range, showing no influence of size and morphology of particles. This value is similar to the activation energy of uranium diffusion in UO 2, which suggests that the sintering process in UO 2 is not controlled by oxygen diffusion. When diffusion coefficients of uranium calculated from sintering data are compared to those deduced from radiotracer experiments, it is rather difficult to decide between a volume and/or a grain boundary diffusion mechanism. However, isothermal shrinkage curves can be more satisfactorily interpreted when a sintering theory based on grain boundary diffusion is considered.