UO2 dissolution in Boom Clay conditions

The solubility of uranium dioxide (UO₂) was measured in real and synthetic Boom Clay waters with varying concentrations of humic acids and carbonate under reducing conditions at 20 °C. Uranium concentrations in function of time suggest the reduction of U(VI) to U(IV) by the humic acids which is occurring faster in real clay water than in synthetic clay waters. Humic acids induce also a competition to complex U(VI) in carbonate-containing solution, but they are not able to control the uranium concentration at high bicarbonate concentration (0.02 mol dm⁻³). Nevertheless they may play a role at low carbonate concentration. In our experimental conditions, the geochemical calculations indicate that two uranium secondary phases (U₄O₉ and UO_2(c)) are susceptible to control the uranium concentration in solution. These calculations are in good agreement with results of the X-ray photoelectron spectroscopy. At the end of tests, uranium concentrations reach steady-state values between 3 × 10⁻⁸ and 5 × 10⁻⁸ mol dm⁻³ in the bicarbonate-rich solutions. Although these concentrations are considered as conservative, they are 10-100 times higher than in natural Boom Clay. The consequence is that spent fuel could slowly dissolve in the interstitial clay water undersaturated with respect to UO₂/UO_2+x of the fuel.     

The effects of alpha-radiolysis on UO2 dissolution determined from batch experiments with Pu-doped UO2

The effects of alpha dose-rate on UO₂ dissolution were investigated by performing dissolution experiments with ²³⁸Pu-doped UO₂ materials containing nominal alpha-activity levels of ∼1-100 Ci/kg UO₂ (actual levels 0.4-80 Ci/kg UO₂), in 0.1 M NaClO₄ and in 0.1 M NaClO₄ + 0.1 M carbonate. Dissolution rates increased less than 10-fold for an almost 100-fold increase in doping level and fall within the range of predictions of the Mixed Potential Model (a detailed mechanistic model for used fuel dissolution). Dissolution rates were lower in carbonate-free solutions and enrichment of ²³⁸Pu on the UO₂ surface was suggested in carbonate solutions. Effective G values, defined as the ratio of the total amount of U dissolved divided by the maximum possible amount of U dissolved by radiolytically produced H₂O₂, increased with decreasing doping levels. This suggests that the dissolution reaction at high dose rates is limited by the reaction rate between UO₂ and H₂O₂, but becomes increasingly limited by the rate of production of H₂O₂ at lower dose rates.     

Leaching behaviour of unirradiated high temperature reactor (HTR) UO2-ThO2 mixed oxides fuel particles

The dissolution of different mixed oxide (U, Th)O₂ 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 UO₂ 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 < 10⁻⁶ g m⁻² d⁻¹). Under reducing conditions, the matrix of HTR fuel particles presents significant intrinsic radionuclide confinement properties.     

The influence of calcium ions on the development of acidity in corrosion product deposits on SIMFUEL, UO2

In order to clarify the influence of groundwater constituents on the formation of corrosion products and secondary phase deposits on corroding/dissolving nuclear fuel surfaces under waste disposal conditions we have investigated the influence of Ca²⁺, present as CaCl₂. The influence of calcium ions on the anodic dissolution of SIMFUEL (doped uranium dioxide) has been characterized over the potential range 0–500 mV (vs. SCE). Through the use of X-ray photoelectron spectroscopy (XPS) the surface composition over this potential range has been determined. Ca²⁺ was found not to influence the conversion of U^IVO₂ to , but to suppress the subsequent formation of a U^VI surface species which lead to the formation of a hydrated deposit, UO₃ · yH₂O. The adsorption of Ca²⁺ on the UO₂ surface is believed to inhibit fuel dissolution either via inhibiting the stabilization of the cation precursor (UO₂(OH)₂)_ads or by blocking the O²⁻ anion transfer reaction from the fuel surface.     

Thermophysical studies on the binary system UO2(NO3)2·6H2O - Sr(NO3)2

Thermoanalytical (TG-DTA-EGA) and X-ray diffraction measurements have been used to study the reaction between uranyl nitrate hexahydrate and strontium nitrate. The results confirmed the absence of a direct interaction between the two compounds. The presence of strontium nitrate, however, ensured that the extent of hydrolysis and polymerisation of uranyl nitrate hexahydrate during its dehydration and decomposition to UO₃ is significantly reduced. DTA curves recorded in both heating and cooling modes gave evidence to the occurrence of a reaction between molten strontium nitrate and uranium trioxide to form nitrato-complexes of uranium and strontium. X-ray diffraction data on reaction residues obtained at different temperatures and cooled to room temperature also showed evidence for the formation of such complexes. The results obtained indicated an increase in thermal stability of these nitrato-complexes with increase in Sr/U ratio. The complex with an Sr/U ratio of 2.0 is stable up to 660 °C and the complex with Sr/U ratio of 4.0 is stable up to 680 °C. These complexes decompose at higher temperatures to give strontium uranates.     

Effect of external gamma irradiation on dissolution of the spent UO2 fuel matrix

Leaching experiments were performed on UO₂ pellets doped with alpha-emitters (^238/239Pu) and on spent fuel, in the presence of an external gamma irradiation source (A⁶⁰Co = 260 Ci,  Gy h⁻¹). The effects of α, β, γ radiation, the fuel chemistry and the nature of the cover gas (aerated or Ar + 4%H₂) on water radiolysis and on oxidizing dissolution of the UO₂ matrix are quantified and discussed. For the doped UO₂ pellets, the nature of the cover gas clearly has a major role in the effect of gamma radiolysis. The uranium dissolution rate in an aerated medium is 83 mg m⁻² d⁻¹ compared with only 6 mg m⁻² d⁻¹ in Ar + 4%H₂. The rate drop is accompanied by a reduction of about four orders of magnitude in the hydrogen peroxide concentrations in the homogeneous solution. The uranium dissolution rates also underestimate the matrix alteration rate because of major precipitation phenomena at the UO₂ pellet surface. The presence of studtite in particular was demonstrated in aerated media; this is consistent with the measured H₂O₂ concentrations (1.2 × 10⁻⁴ mol L⁻¹). For spent fuel, the presence of fission products (Cs and Sr), matrix alteration tracers, allowed us to determine the alteration rates under external gamma irradiation. The fission product release rates were higher by a factor of 5-10 than those of the actinides (80-90% of the actinides precipitated on the surface of the fragments) and also depended to a large extent on the nature of the cover gas. No significant effect of the fuel chemistry compared with UO₂ was observed on uranium dissolution and H₂O₂ production in the presence of the ⁶⁰Co source in aerated conditions. Conversely, in Ar + 4%H₂ the fuel self-irradiation field cannot be disregarded since the H₂O₂ concentrations drop by only three orders of magnitude compared with UO₂.     

Radiolytic modelling of spent fuel oxidative dissolution mechanism. Calibration against UO2 dynamic leaching experiments

Calibration and testing are inherent aspects of any modelling exercise and consequently they are key issues in developing a model for the oxidative dissolution of spent fuel. In the present work we present the outcome of the calibration process for the kinetic constants of a UO₂ oxidative dissolution mechanism developed for using in a radiolytic model. Experimental data obtained in dynamic leaching experiments of unirradiated UO₂ has been used for this purpose. The iterative calibration process has provided some insight into the detailed mechanism taking place in the alteration of UO₂, particularly the role of OH radicals and their interaction with the carbonate system. The results show that, although more simulations are needed for testing in different experimental systems, the calibrated oxidative dissolution mechanism could be included in radiolytic models to gain confidence in the prediction of the long-term alteration rate of the spent fuel under repository conditions.     

Recycling of chemicals from alkaline waste generated during preparation of UO3 microspheres by sol–gel process

Internal gelation process, one of the sol–gel processes for nuclear fuel fabrication, offers many advantages over conventional powder pellet route. However, one of the limitation of the process is generation of large volume of alkaline liquid waste containing hexamethylenetetramine, urea, ammonium nitrate, ammonium hydroxide etc. Presence of ammonium nitrate with hexamethylenetetramine and urea presents a fire hazard which prevents direct disposal of the waste as well as its recycle by evaporation. The paper describes the studies carried out to suitably process the waste. Nitrate was removed from the waste by passing through Dowex 1 × 4 anion exchange resin in OH⁻ form. 1.0 M NaOH was used to regenerate the resin. The nitrate-free waste was further treated to recover and recycle hexamethylenetetramine, urea and ammonium hydroxide for preparation of UO₃ microspheres. The quality of the microspheres obtained was satisfactory. An optimized flow sheet for processing of the waste solution has been suggested.     

Phase diagram of the UO2-FeO1+x system

Phase-relation studies of the UO₂-FeO_1+x system in an inert atmosphere are presented. The eutectic point has been determined, which corresponds to a temperature of (1335 ± 5) °C and a UO₂ concentration of (4.0 ± 0.1) mol.%. The maximum solubility of FeO in UO₂ at the eutectic temperature has been estimated as (17.0 ± 1.0) mol.%. Liquidus temperatures for a wide concentration range have been determined and a phase diagram of the system has been constructed.     

Formation of columnar and equiaxed grains by the reduction of U3O8 pellets to UO2+x

The reduction of U₃O₈ pellets to UO_2+x has been investigated at 1300 °C in H₂, Ar and CO₂ gas atmospheres by TGA, SEM, and X-ray diffraction. The selected U₃O₈ pellet was prepared by sintering a U₃O₈ powder compact. The TGA results show that the reduction rate is fastest in H₂ gas, and X-ray diffraction indicates that U₃O₈ reduces to UO_2+x without any intermediate phase. The reduced pellet, UO_2+x, has a special grain structure that consists of equiaxed grains at the surface, columnar grains in the middle, and equiaxed grains in the center. The equiaxed grains and columnar grains are much smaller in H₂ gas than in Ar or CO₂ gas. The reducing gases significantly influence the morphology of the grain structure. This difference can be explained in terms of a relation between oxygen potential and critical nucleus size during the reduction.     

Simulations of heat and oxygen diffusion in UO2 nuclear fuel rods

We present finite-element simulations of coupled heat and oxygen atom diffusion for UO₂ fuel pellets. The expressions for thermal conductivity, specific heat and oxygen diffusivity for the fuel element are obtained directly from previously published correlations, or from analysis of previously published data. We examine the temperature and non-stoichiometry distributions for a varying range of conditions. Simulations are performed for steady-state and transient regime in one-dimensional (purely radial) configurations. For steady-state conditions we perform parametric studies that determine the maximum temperature in the fuel rod as a function of non-stoichiometry and heat generation rate intensity. For transient simulations, we examine the time lag in the response of the temperature and non-stoichiometry distributions with respect to sudden changes in heat generation rate intensity and oxygen removal rate. All simulations are performed with the commercial code COMSOL Multiphysics™.     

Fission product release in high-burn-up UO2 oxidized to U3O8

Results of oxidation experiments on high-burn-up UO₂ are presented where fission-product vaporisation and release rates have been measured by on-line mass spectrometry as a function of time/temperature during thermal annealing treatments in a Knudsen cell under controlled oxygen atmosphere. Fractional release curves of fission gas and other less volatile fission products in the temperature range 800-2000 K were obtained from BWR fuel samples of 65 GWd t⁻¹ burn-up and oxidized to U₃O₈ at low temperature. The diffusion enthalpy of gaseous fission products and helium in different structures of U₃O₈ was determined.     

Calculation of the threshold displacement energies in UO2 using ionic potentials

A set of ionic potentials matching exactly the crystallographic, elastic and dielectric properties of the uranium dioxide is established. It is further validated upon some basic thermodynamic properties as well as upon the Frenkel pairs formation energies and the activation energies for lattice migration in UO₂. The threshold displacement energies, useful to characterise the radiation resistance of the materials, are calculated for the uranium dioxide along various crystallographic directions applying the optimised force field within the sudden approximation approach.     

Modelling intragranular fission gas release in irradiation of sintered LWR UO2 fuel

A model for the release of stable fission gases by diffusion from sintered LWR UO₂ fuel grains is presented. The model takes into account intragranular gas bubble behaviour as a function of grain radius. The bubbles are assumed to be immobile and the gas migrates to grain boundaries by diffusion of single gas atoms. The intragranular bubble population in the model at low burn-ups or temperatures consists of numerous small bubbles. The presence of the bubbles attenuates the effective gas atom diffusion coefficient. Rapid coarsening of the bubble population in increased burn-up at elevated temperatures weakens significantly the attenuation of the effective diffusion coefficient. The solution method introduced in earlier papers, locally accurate method, is enhanced to allow accurate calculation of the intragranular gas behaviour in time varying conditions without excessive computing time. Qualitatively the detailed model can predict the gas retention in the grain better than a more simple model.     

A fabrication technique for a UO2 pellet consisting of UO2 grains and a continuous W channel on the grain boundary

A new fabrication process of UO₂-W composite fuel has been studied in order to improve the thermal conductivity of the UO₂ pellet by the addition of a small amount of W. A fabrication process was designed from the phase equilibria among tungsten, tungsten oxides and UO₂. The conventionally sintered UO₂ pellet which contains W particles is heat-treated in an oxidizing gas and then in a reducing gas. In the oxidizing heat-treatment W particles are oxidized and liquid tungsten oxide penetrates within the UO₂ grain boundary, and in the reducing heat-treatment liquid oxide is transformed to solid tungsten which forms a continuous channel along the UO₂ grain boundary. This developed technique can provide a continuous W channel covering UO₂ grains for a UO₂-W composite fuel even with a small amount of a metal phase - below 6 vol.%. The thermal diffusivity of the UO₂-6 vol.%W cermet composite increases by about 80% when compared with that of a pure UO₂ pellet.     

On stability of spatial distributions of crystal structure defects in irradiated high burnup UO2 fuel

Conditions of Kinoshita instability development of point defects and dislocation spatial distributions in the crystal structure of UO₂ fuel are studied. As a result of the instability development, spatially non-uniform regions with increased dislocation density are formed. Closed-form expressions of instability increment and spatial scale are derived. Parameters of the instability for irradiation conditions of high burnup UO₂ fuel are obtained by means of numerical simulation. Instability development time is shown to be inversely proportional to fission rate and it increases as dislocation density decreases. Calculated values of instability spatial scale and increment are in accordance with the size of fine grains and their formation rate in the peripheral zones of high burnup LWR fuel pellets.     

Chlorination of UO2, PuO2 and rare earth oxides using ZrCl4 in LiCl-KCl eutectic melt

A new chlorination method using ZrCl₄ in a molten salt bath has been investigated for the pyrometallurgical reprocessing of nuclear fuels. ZrCl₄ has a high reactivity with oxygen but is not corrosive to refractory metals such as steel. Rare earth oxides (La₂O₃, CeO₂, Nd₂O₃ and Y₂O₃) and actinide oxides (UO₂ and PuO₂) were allowed to react with ZrCl₄ in a LiCl-KCl eutectic salt at 773 K to give a metal chloride solution and a precipitate of ZrO₂. An addition of zirconium metal as a reductant was effective in chlorinating the dioxides. When the oxides were in powder form, the reaction was observed to progress rapidly. Cyclic voltammetry provided a convenient way of establishing when the reaction was completed. It was demonstrated that the ZrCl₄ chlorination method, free from corrosive gas, was very simple and useful.     

Thermodynamics of the O-U system. I - Oxygen chemical potential critical assessment in the UO2-U3O8 composition range

A critical assessment of oxygen chemical potential of UO_2+x, U₄O₉ and U₃O₈ oxide non-stoichiometric phases as well as of diphasic related domains has been performed in order to build up primary input data files used in a further optimization procedure of thermodynamic and phase diagram data for the uranium-oxygen system in the UO₂-UO₃ composition range. Owing to the fact that original data are very numerous, more than 500 publications, a twofold process is used for the assessment - (i) first a critical selection of data is performed for each method of measurement together with a careful estimate of their uncertainties, (ii) second a reduction of the total number of data on the basis of a chart with fixed intervals of temperature and composition that allows a comparison to be made of the results from the various experiments. Results are presented for chemical potentials of oxygen with their associated uncertainties.     

Thermal variation of the optical absorption of UO2: determination of the small polaron self-energy

The temperature variation of UV-VIS-NIR optical spectra of UO₂ have been investigated from room temperature up to 1173 K with careful in situ oxygen partial pressure control. The deduced optical absorption edge exhibits a strong temperature dependence. Its value decreases from ∼2 eV at room temperature to ∼0.8 eV at 1173 K. Such thermal behaviour is interpreted as the consequence of the existence of a strong electron-phonon coupling (small polaron). In the temperature range 300-1173 K, the model yields a hopping radius of ∼2 Å and a polaron self-energy of E_p=−0.38 eV.     

Characterization and densification studies on ThO2-UO2 pellets derived from ThO2 and U3O8 powders

ThO₂ containing around 2-3% ²³³UO₂ is the proposed fuel for the forthcoming Indian Advanced Heavy Water Reactor (AHWR). This fuel is prepared by powder metallurgy technique using ThO₂ and U₃O₈ powders as the starting material. The densification behaviour of the fuel was evaluated using a high temperature dilatometer in four different atmospheres Ar, Ar-8%H₂, CO₂ and air. Air was found to be the best medium for sintering among them. For Ar and Ar-8%H₂ atmospheres, the former gave a slightly higher densification. Thermogravimetric studies carried out on ThO₂-2%U₃O₈ granules in air showed a continuous decrease in weight up to 1500 °C. The effectiveness of U₃O₈ in enhancing the sintering of ThO₂ has been established.