A XAS study of the local environments of cations in (U, Ce)O2

Mixed oxide (MOX) fuel is usually considered as a solid solution formed by uranium and plutonium dioxides. Nevertheless, some physico-chemical properties of (U_1−y, Pu_y)O₂ samples manufactured under industrial conditions showed anomalies in the domain of plutonium contents ranging between 3 and 15 at.%. Cerium is commonly used as an inactive analogue of plutonium in preliminary studies on MOX fuels. Extended X-ray Absorption Fine Structure (EXAFS) measurements performed at the European Synchrotron Radiation Facility (ESRF) at the cerium and uranium edges on (U_1−y, Ce_y)O₂ samples are presented and discussed. They confirmed on an atomic scale the formation of an ideal solid solution for cerium concentrations ranging between 0 and 50 at.%.     

Fuel-to-cladding gap evolution and its impact on thermal performance of high burnup fast reactor type uranium-plutonium oxide fuel pins

Drastic evolution of fuel-to-cladding gap is observed in high burnup JOYO Mk-II driver and MONJU type uranium-plutonium oxide fuel pins. The effect of the evolution is examined from viewpoints of fuel restructuring, gaseous FP release and retention and cesium migration behaviors. Its thermal impact on fuel pin performance is also studied by one-dimensional steady state thermal analysis. Threshold condition of the evolution depends on fuel pellet characteristics, burnup and probably temperature. The evolution directly relates to as-fabricated microstructures and to gaseous FP release and retention behavior. A comparison of fuel restructuring with predicted temperature profiles indicates that, even where large residual gaps are observed, non-gaseous filler always improves the heat transfer across the gaps.     

Change of fuel-to-cladding gap width with the burn-up in FBR MOX fuel irradiated to high burn-up

In order to study the dependence of the gap width change on the burn-up, the fuel-to-cladding gap widths were investigated by ceramography in a large number of FBR MOX fuel pins irradiated to high burn-up. The dependence of gap widths on the burn-up was closely connected with the formations of JOG (joint oxyde-gaine) and rim structure. The gap widths decreased gradually due to the fuel swelling until ∼30 GWd/t, but beyond this burn-up the dependence showed two different tendencies. With the increase of burn-up, the gap widths decreased due to the increase of fuel swelling in the low fuel temperature region where the rim structure was observed, but they increased in the high fuel temperature region where the JOG rich in Cs and Mo formed in the gap.     

Modeling the activity of I and Cs in the primary coolant and CVCS resin of an operating PWR

Mathematical models have been developed to describe the activities of ¹²⁹I and ¹³⁷Cs in the primary coolant and resin of the chemical and volume control system (CVCS) during constant power operation in a pressurized water reactor (PWR). The models, which account for the source releases from defective fuel rod(s) and tramp uranium, rely on the contribution of CVCS resin and boron recovery system as a removal process, and differences in behavior for each nuclide. The current models were validated through measured coolant activities of ¹³⁷Cs. The resultant scaling factors agree reasonably well with the results of the test resin of the coolant and the actual resins from the PWRs of other countries.     

A model for the behavior of thorium uranium mixed oxide kernels in the pelletizing process

A behavior model of nuclear fuel kernels in the pelletizing process was developed to predict the microstructure of (Th,5%U)O₂ sintered pellets. Methods, equipments and components were developed in order to measure the density, the specific surface area and the crushing strength of the kernels and produce fuel pellets. It enables a correlation between the kernels properties and the microstructure, density and open porosity that were obtained in the fuel pellet produced with these kernels. It was possible to obtain a mathematical expression that allows one to calculate, from the kernel density and specific surface, the density that will be obtained in the fuel pellet for each compactation pressure value. The investigation showed which kernels properties are desired to obtain fuel pellets that satisfy the quality requirements for a stable performance in a power reactor. This model has been validated by experimental results and fuel pellets were obtained with an optimized microstructure that satisfies the fuel specification for an in-pile stable behavior.     

Modelling the activity of I in the primary coolant of a CANDU reactor

A mathematical treatment has been developed to describe the activity levels of ¹²⁹I as a function of time in the primary heat transport system during constant power operation and for a reactor shutdown situation. The model accounts for a release of fission-product iodine from defective fuel rods and tramp uranium contamination on in-core surfaces. The physical transport constants of the model are derived from a coolant activity analysis of the short-lived radioiodine species. An estimate of 3×10⁻⁹ has been determined for the coolant activity ratio of ¹²⁹I/¹³¹I in a CANDU Nuclear Generating Station (NGS), which is in reasonable agreement with that observed in the primary coolant and for plant test resin columns from pressurized and boiling water reactor plants. The model has been further applied to a CANDU NGS, by fitting it to the observed short-lived iodine and long-lived cesium data, to yield a coolant activity ratio of ∼2×10⁻⁸ for ¹²⁹I/¹³⁷Cs. This ratio can be used to estimate the levels of ¹²⁹I in reactor waste based on a measurement of the activity of ¹³⁷Cs.     

Synthesis and characterization of low-temperature precursors of thorium-uranium (IV) phosphate-diphosphate solid solutions

Several compositions of new precursor of thorium-uranium (IV) phosphate-diphosphate solid solutions (Th_4−xU_x(PO₄)₄P₂O₇, called β-TUPD) were synthesized in closed PTFE containers either in autoclave (160 °C) or on sand bath (90-160 °C). All the samples appeared to be single phase. From XRD data and TEM observations, the diffraction lines matched well with that of pure thorium phosphate-hydrogenphosphate hydrate (TPHPH), Th₂(PO₄)₂(HPO₄) · H₂O, which confirmed the preparation of a complete solid solution between pure thorium and uranium (IV) compounds. TGA/DTA experiments showed that samples of thorium-uranium (IV) phosphate-hydrogenphosphate hydrate (TUPHPH) prepared at 150-160 °C were monohydrated leading to the proposed formula Th_2−x/2U_x/2(PO₄)₂(HPO₄) · H₂O. The variation of the XRD diagrams versus the heating temperature showed that TUPHPH remained crystallized and single phase from room temperature to 200 °C. After heating between 200 °C and 800 °C, the presence of diphosphate groups in the solid was evidenced. In this range of temperature, the solid was transformed into the low-temperature monoclinic form of thorium-uranium (IV) phosphate-diphosphate (α-TUPD). This latter compound finally turned into well-crystallized, homogeneous and single-phase β-TUPD (orthorhombic form) above 930-950 °C for x values lower than 2.80. For higher x values, a mixture of β-TUPD, α-Th_1−zU_zP₂O₇ and U_2−wTh_wO(PO₄)₂ was obtained. By this new chemical route of preparation of β-TUPD solid solutions, the homogeneity of the samples is significantly improved, especially considering the distribution of thorium and uranium.     

Kinetic and thermodynamic studies of the dissolution of thoria-urania solid solutions

The dissolution of Th_1−xU_xO₂ was investigated through leaching experiments combined with X-ray photoelectron spectroscopy (XPS) and X-ray absorption spectroscopy (XAS) analyses. These experiments were performed in acidic and in oxidizing conditions (nitric solutions), for several compositions of solid solutions ranging from x = 0.24 to 0.81. Static sequential experiments in acidic media performed at room temperature confirmed that higher concentration of uranium in the solid solution leads to higher release of uranium in the leachate whatever the pH. The normalized dissolution rate in oxidizing media is increasing all the more the content of uranium is increases in the mixed oxide. While for Th enriched solids, kinetic parameters remain similar to that of ThO₂, in the case of uranium enriched solids, a drastic change is observed, and kinetic parameters are similar to that of UO₂ ones. For x > 0.50, the saturation is reached in the leachate after 100 days. XPS and EXAFS analysis on leached samples pointed out an oxidation of U(IV) at the surface for x < 0.5, and in the bulk for x > 0.5. Enrichment in Th is also observed at the surface of the solid, indicating the formation of a protective layer of hydrated thorium oxide, or hydroxide. Finally, the solubility product of secondary phase was determined. The values obtained are in good agreement with that of ThO₂, Th(OH)₄ and ThO₂, xH₂O reported in the literature.     

Examination of U valence states in the brannerite structure by near-infrared diffuse reflectance and X-ray photoelectron spectroscopies

The valence state of uranium doped into a f⁰ thorium analog of brannerite (i.e., thorutite) has been examined using near-infrared (NIR) diffuse reflectance (DRS) and X-ray photoelectron (XPS) spectroscopies. NIR transitions of U⁴⁺, which are not observed in spectra of brannerite, have been detected in the samples of U_xTh_1−xTi₂O₆, and we propose that strong specular reflectance is responsible for the lack of U⁴⁺ features in UTi₂O₆. Characteristic U⁵⁺ bands have been identified in samples in which sufficient Ca²⁺ has been added to nominally effect complete oxidation to U⁵⁺. XPS results support the assignments of U⁴⁺ and U⁵⁺ by DRS. The presence of residual U⁴⁺ bands in the spectra of the Ca-doped samples is consistent with segregation of Ca²⁺ to the grain boundaries during high temperature sintering.     

Calculations on fission gas behaviour in the high burnup structure

The behaviour of fission gas in high burnup fuel during steady-state and transient conditions is of special interest for safety reasons. Despite this, mechanistic models that reflect the fission gas transport processes and reliably predict the evolution of the remaining fission gas in the high burnup structure (HBS) are largely missing today. We start to address this problem by developing a one-dimensional, mass balance model and apply it to LWR UO₂ fuel at the moderate temperatures found in the rim region. We examine the quantity of gas remaining in the HBS fuel matrix at steady state and compare it with experimental values. We find that the current model reproduces the 0.2 wt% observed xenon concentration under certain conditions, viz. fast grain boundary diffusion and an effective volume diffusion coefficient. A sensitivity analysis is also conducted for the model parameters, the relative importance for which is not well established a priori.     

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.     

An interdiffusional model for prediction of the interaction layer growth in the system uranium–molybdenum/aluminum

The codes PLACA and DPLACA, elaborated in this working group, simulate the behavior of a plate-type fuel containing in its core a foil of monolithic or dispersed fissile material, respectively, under normal operation conditions of a research reactor. Dispersion fuels usually consist of ceramic particles of a uranium compound in a high thermal conductivity matrix. The use of particles of a U–Mo alloy in a matrix of Al requires especially devoted subroutines able to simulate the growth of the interaction layer that develops between the particles and the matrix. A model is presented in this work that gives account of these particular phenomena. It is based on the assumption that diffusion of U and Al through the layer is the rate-determining step. Two moving interfaces separate the growing reaction layer from the original phases. The kinetics of these boundaries are solved as Stefan problems. In order to test the model and the associated code, some previous, simpler problems corresponding to similar systems for which analytical solutions or experimental data are known were simulated. Experiments performed with planar U–Mo/Al diffusion couples are reported in the literature, which purpose is to obtain information on the system parameters. These experiments were simulated with PLACA. Results of experiments performed with U–Mo particles disperse in Al either without or with irradiation, published in the open literature were simulated with DPLACA. A satisfactory prediction of the whole reaction layer thickness and of the individual fractions corresponding to alloy and matrix consumption was obtained.     

Crystal chemistry of uranium (V) and plutonium (IV) in a titanate ceramic for disposition of surplus fissile material

We report X-ray absorption near-edge structure (XANES) and extended X-ray absorption fine-structure (EXAFS) spectra for the plutonium LIII and uranium LIII edges in titanate pyrochlore ceramic. The titanate ceramics studied are of the type proposed to serve as a matrix for the immobilization of surplus fissile materials. The samples studied contain approximately 10 wt% fissile plutonium and 20 wt% natural uranium, and are representative of material within the planned production envelope. Based upon natural analogue models, it had been previously assumed that both uranium and plutonium would occupy the calcium site in the pyrochlore crystal structure. While the XANES and EXAFS signals from the plutonium LIII are consistent with this substitution into the calcium site within pyrochlore, the uranium XANES is characteristic of pentavalent uranium. Furthermore, the EXAFS signal from the uranium has a distinct oxygen coordination shell at 2.07 Å and a total oxygen coordination of about 6, which is inconsistent with the calcium site. These combined EXAFS and XANES results provide the first evidence of substantial pentavalent uranium in an octahedral site in pyrochlore. This may also explain the copious nucleation of rutile (TiO₂) precipitates commonly observed in these materials as uranium displaces titanium from the octahedral sites.     

Characterization of zirconia-thoria-urania ceramics by X-ray and electron interactions

X-ray and electron interactions with matter were used as probes to characterize the structure and chemistry of zirconia-thoria-urania ceramics. The ceramics were prepared by coprecipitation of Zr, Th and U salts. In this study, transmission electron microscopy (TEM) techniques such as energy dispersive X-ray (EDX) analysis and electron energy loss spectroscopy (EELS) complement X-ray diffraction, extended X-ray absorption fine structure (EXAFS) and X-ray absorption near edge spectroscopy (XANES), techniques to reveal the phase structure and chemistry. The results from XRD and EDX show that these ceramics separate into a Zr-based phase and an actinide-based phase with low mutual affinity of Th and Zr, as well as partial solubility of U in Zr. The comparison of EELS spectra collected for the ceramics with spectra collected for UO₂ and U₃O₈ reference materials also allow us to assess U oxidation state independently in the two separate phases.     

Incorporation of cerium and neodymium in uranyl phases

The potential for incorporating rare earth elements (REE) into/onto crystalline compounds has been evaluated by precipitating uranyl phases from aqueous solutions containing either cerium or neodymium. These REEs serve both as monitors for evaluating the potential repository behavior of REE radionuclides, and as surrogate elements for actinides (e.g., Ce⁴⁺ and Nd³⁺ for Pu⁴⁺ and Am³⁺, respectively). The present experiments examined the behavior of REE in the presence of ianthinite , becquerelite (Ca(UO₂)₆O₄(OH)₆(H₂O)₈), and other uranyl hydroxide compounds commonly noted as alteration products during the corrosion of UO₂, spent nuclear fuel, and naturally occurring uraninite. The results of these experiments demonstrate that significant quantities of both cerium (K_d = 1020) and neodymium (K_d = 840) are incorporated within the uranium alteration phases and suggest that ionic substitution and/or adsorption to the uranyl phases can play a key role in the limiting the mobility of REE (and by analogy, actinide elements) in a nuclear waste repository.     

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.     

Thermodynamic properties of the O-U system. II - Critical assessment of the stability and composition range of the oxides UO2+x, U4O9−y and U3O8−z

The published data concerned with the determination of the composition ranges of uranium oxides, UO_2+x, U₄O_9−y and U₃O_8−z, which have been determined using thermogravimetric, X-ray diffraction and electrochemical techniques are critically assessed. U₄O₉ and U₃O₈ have quite small domains of composition and the assessment of such data has carefully considered the uncertainties in the experimental determinations. In addition, the thermodynamic properties of U₄O₉ and U₃O₈, enthalpies of formation and transformation, entropies, and thermal capacities are analyzed and selected to build a primary data base for compounds.     

Structure of zirconium alloy oxides formed in pure water studied with synchrotron radiation and optical microscopy: relation to corrosion rate

A detailed study was undertaken of oxides formed in 360 °C water on four Zr-based alloys (Zircaloy-4, ZIRLO™,¹ Zr-2.5%Nb and Zr-2.5%Nb-0.5%Cu) in an effort to relate oxide structure to corrosion performance. Micro-beam X-ray diffraction was used along with transmitted light optical microscopy to obtain information about the structure of these oxides as a function of distance from the oxide-metal interface. Optical microscopy revealed a layered oxide structure in which the average layer thickness was inversely proportional to the post-transition corrosion rate. The detailed diffraction studies showed an oxide that contained both tetragonal and monoclinic ZrO₂, with a higher fraction of tetragonal oxide near the oxide-metal interface, in a region roughly corresponding to one oxide layer. Evidence was seen also of a cyclic variation of the tetragonal and monoclinic oxide across the oxide thickness with a period of the layer thickness. The results also indicate that the final grain size of the tetragonal phase is smaller than that of the monoclinic phase and the monoclinic grain size is smaller in Zircaloy-4 and ZIRLO than in the other two alloys. These results are discussed in terms of a model of oxide growth based on the periodic breakdown and reconstitution of a protective layer.     

Ab initio energetics of some fission products (Kr, I, Cs, Sr and He) in uranium dioxide

A computational study of some fission products (FP) energetics in uranium dioxide is presented. Krypton, iodine, caesium, strontium and helium are considered. Calculations are made within the density functional theory in the local density approximation with the plane wave pseudopotential method. Three insertion sites are considered: the octahedral interstitial position and the oxygen and uranium substitution sites. The importance of atomic relaxations is estimated on the He and Kr cases. They prove quantitatively important but can be neglected to draw qualitative trends. For each fission product incorporation and solution energies are calculated. The obtained values of the solutions energies of the various FP are in good agreement with their experimental behaviour: Kr, Cs and I atoms are insoluble in uranium dioxide, Sr solubility depends on the stoichiometry of urania. He atoms are found to have little interaction with their environment in uranium doxide.     

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.