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.     

Processing and fabrication of mixed uranium/refractory metal carbide fuels with liquid-phase sintering

Optimization of powder processing techniques were sought for the fabrication of single-phase, solid-solution mixed uranium/refractory metal carbide nuclear fuels - namely (U, Zr, Nb)C. These advanced, ultra-high temperature nuclear fuels have great potential for improved performance over graphite matrix, dispersed fuels tested in the Rover/NERVA program of the 1960s and early 1970s. Hypostoichiometric fuel samples with carbon-to-metal ratios of 0.98, uranium metal mole fractions of 5% and 10%, and porosities less than 5% were fabricated. These qualities should provide for the longest life and highest performance capability for these fuels. Study and optimization of processing methods were necessary to provide the quality assurance of samples for meaningful testing and assessment of performance for nuclear thermal propulsion applications. The processing parameters and benefits of enhanced sintering by uranium carbide liquid-phase sintering were established for the rapid and effective consolidation and formation of a solid-solution mixed carbide nuclear fuel.     

Thermal and physical properties of molten fluorides for nuclear applications

LiF-BeF₂-ThF₄ is a key system in molten salt reactor fuel studies. In this paper we give an overview of some important features of this ternary system. We discuss the phase behavior, vapor pressure, density and viscosity, based on what is known in the literature and on our own data from previous work on the thermodynamic assessment of LiF-BeF₂-ThF₄.     

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™.     

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.     

A critical examination of the thermodynamics of water adsorption on actinide oxide surfaces

The reversible adsorption of water from actinide oxide surfaces is examined from several viewpoints in this article. A reinterpretation and critical look at the previously published thermodynamic values for desorption of water from PuO₂ [J. Phys. Chem. 77 (1973) 581] are reexamined in light of more recent mathematical treatments of thermal desorption data from high surface area materials. In addition, the time and temperature dependent process of water adsorption/desorption in closed system experiments is examined using chemical kinetics modeling. A simple experimental method and mathematical treatment of determining adsorption enthalpies based upon a closed system is also described. The desorption enthalpy for reversibly adsorbed water from PuO₂ is determined to be a function of adsorbate coverage with values ranging from 51 to 44 kJ mol⁻¹ for coverages of one to several monolayers (MLs). Consistent desorption enthalpy values are obtained using either approach thus highlighting the importance of proper interpretation of adsorption parameters determined from high surface area powders. Reversible adsorption/desorption equilibrium of water with actinide oxide materials is discussed from the practical standpoint of storage and subsequent pressurization of containers. These results obtained from PuO₂ surfaces are consistent with desorption enthalpies of water from a low surface area UO₂ that has been measured using ultra-high vacuum thermal desorption mass spectroscopy to be 42.2 kJ mol⁻¹.     

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.     

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.     

Mechanical behavior of δ-phase Pu-Ga alloys. Part II: Model verification and application

We validated the mechanical threshold strength (MTS) model, developed in Part I, with approximately 50 different experimental results from the literature for both yield strength and ultimate tensile strength on Pu-Ga alloys. One standard deviation of the differences between the model’s yield-strength predictions and the experiments was 7.5% of the measured yield strength. The model also worked well predicting the ultimate tensile strength (UTS) of the alloys with gallium concentrations of 1 wt% or greater, although the accuracy of the UTS predictions was not as good as for yield strength. After validating the model, we studied the effects of gallium concentration, grain size, iron and nickel content, and carbon concentration on the yield strength of Pu-Ga alloys. The gallium concentration affected the yield strength more than any other microstructural variable. The yield strength increased 50% between 1 at.% Ga and 5.4 at.% Ga alloying addition. The grain size also produced a measurable strengthening effect, typical of other face-centered cubic metals. The yield strength increased 15% with a reduction in grain size from 50 μm to 10 μm. Finally, we found that there were no observable yield-strength effects resulting from different amounts of iron, nickel, or carbon impurities.     

Production of uranium-molybdenum particles by spark-erosion

With the spark-erosion method we have produced spheroidal particles of an uranium-molybdenum alloy using pure water as dielectric. The particles were characterized by optical metallography, scanning electron microscopy, energy dispersive spectrometry and X-ray diffraction. Mostly spherical particles of UO₂ with a distinctive size distribution with peaks centered at 70 and 10 μm were obtained. The particles have central inclusions of U and Mo compounds.     

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.     

Uranium oxide weathering: spectroscopy and kinetics

Particles of UO_2+x (x≅0.16 ± 0.06) exposed to the atmosphere react by oxidation and formation of complexes (hydrates, hydroxides and carbonates). Surface reactions alter and erode the UO₂ particles. This paper outlines results for measurements of oxidation rates on uranium oxide particles using in situ photoluminescence spectroscopy (PL), X-ray photoelectron spectroscopy (XPS) and secondary ion mass spectrometry (SIMS). Phosphorescence spectra observed during oxidation of UO_2+x were attributed to U(VI) in uranyl-type coordination and in octahedral coordination. Uranyl-type spectra formed during wet oxidation of UO_2+x, and U(VI) octahedral spectra formed during dry oxidation of UO_2+x. The uranyl-type species, although more stable, is more kinetically labile for vacuum reduction than is the octahedral U(VI). Oxidation of U(IV) species are diffusion controlled. Vacuum reduction of uranyl U(VI) in UO₃ follows a field-enhanced cationic diffusion rate law, while re-oxidation follows a diffusion rate law. Post-oxidation core and valence band XPS and SIMS measurements provided qualitative and quantitative measures of uranium oxidation states near uranium oxide surfaces.     

Vitrification of high chrome oxide nuclear waste in iron phosphate glasses

A simulated high level waste (HLW) containing 4 mass% chrome oxide, whose overall composition is representative of the high chrome oxide wastes at Hanford WA USA, was easily vitrified in a phosphate glass at temperatures ranging from 1150 °C, for waste loadings of 55 mass%, to 1250 °C for waste loadings of 75 mass%. Even at these high waste loadings, these wasteforms had an excellent chemical durability. The best chemical durability was achieved when the O/(Si + P) atomic ratio was between 3.5 and 3.8. These wasteforms were also resistant to crystallization although trace amounts of crystalline Cr₂O₃ were present in wasteforms containing more than 70 mass% HLW. It is concluded that up to 45 mass% of the total HLW at Hanford, especially that containing as high as 4.5 mass% chrome oxide, could be directly vitrified into an iron phosphate glass, that meets all of the current chemical durability requirements by simply adding 25-35 mass% P₂O₅ to the waste and melting the mixture at 1150-1250 °C for a few (<6) hours.     

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.     

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.     

On the thermal conductivity of UO2 nuclear fuel at a high burn-up of around 100 MWd/kgHM

A study of the thermal conductivity of a commercial PWR fuel with an average pellet burn-up of 102 MWd/kgHM is described. The thermal conductivity data reported were derived from the thermal diffusivity measured by the laser flash method. The factors determining the fuel thermal conductivity at high burn-up were elucidated by investigating the recovery that occurred during thermal annealing. It was found that the thermal conductivity in the outer region of the fuel was much higher than it would have been if the high burn-up structure were not present. The increase in thermal conductivity is a consequence of the removal of fission products and radiation defects from the fuel lattice during recrystallisation of the fuel grains (an integral part of the formation process of the high burn-up structure). The gas porosity in the high burn-up structure lowers the increase in thermal conductivity caused by recrystallisation.     

Behavior of thorium-uranium (IV) phosphate-diphosphate sintered samples during leaching tests. Part II. Saturation processes

Sintered pellets of thorium-uranium (IV) phosphate-diphosphate solid solutions (β-Th_4−xU_x(PO₄)₄P₂O₇, β-TUPD) were altered in several acidic media. All the results reported in the first part of this paper confirmed the good chemical durability of the samples. The evolution of the normalized weight loss showed that, in several media, thorium quickly precipitates in a neoformed phosphate-based phase while uranium (IV) is released in the leachate due to its oxidation into the uranyl form. The characterization of neoformed phases was carried out through several techniques involving grazing XRD, infrared and μ-Raman spectroscopies, EPMA, SEM and TEM. SEM micrographies showed that the dissolution mainly occurs at the grain boundaries, leading to the break away of the grains: only the first 15 μm are altered for 2 months in 10⁻¹ M HNO₃. From EPMA and BET measurements, neither the chemical composition nor the specific surface area are significantly modified. Near equilibrium, two neoformed phases were observed and identified by grazing XRD and/or μ-Raman spectroscopy at the surface of the leached pellets: one is found to be amorphous and progressively turns into the crystallized thorium phosphate-hydrogenphosphate hydrate (TPHPH). From the results obtained, a chemical scheme of the dissolution of β-TUPD sintered samples is proposed. The behavior of the actinides in the gelatinous phase appears mainly driven by their oxidation state: thorium remains in the tetrapositive state and is quickly and quantitatively precipitated while uranium (IV) is oxidized into uranyl then released in the leachate. The Th-precipitation as TPHPH first appears scattered then covers the entire surface of the pellet, inducing a delay of the actinides release in the leachate. Both phases act as protective layers and should induce the significant delay of the release of actinides (Th, U) to the biosphere.     

Statistical approach and benchmarking for modeling of multi-dimensional behavior in TRISO-coated fuel particles

The fundamental design for a gas-cooled reactor relies on the behavior of the coated particle fuel. The coating layers, termed the TRISO coating, act as a mini-pressure vessel that retains fission products. Results of US irradiation experiments show that many more fuel particles have failed than can be attributed to one-dimensional pressure vessel failures alone. Post-irradiation examinations indicate that multi-dimensional effects, such as the presence of irradiation-induced shrinkage cracks in the inner pyrolytic carbon layer, contribute to these failures. To address these effects, the methods of prior one-dimensional models are expanded to capture the stress intensification associated with multi-dimensional behavior. An approximation of the stress levels enables the treatment of statistical variations in numerous design parameters and Monte Carlo sampling over a large number of particles. The approach is shown to make reasonable predictions when used to calculate failure probabilities for irradiation experiments of the New Production - Modular High Temperature Gas Cooled Reactor Program.     

Behavior of thorium-uranium (IV) phosphate-diphosphate sintered samples during leaching tests. Part I - Kinetic study

The dissolution of β-TUPD sintered samples was examined in various conditions of pH, temperature, concentrations of anions in the leachate and leaching flow rates. All the normalized dissolution rates were in the range 10⁻⁷ to 10⁻⁴ g m⁻² day⁻¹ even in very aggressive media, showing the good resistance of these ceramics to aqueous alteration. The first part of this paper describes several parameters exhibiting a significant influence on the normalized dissolution rate of the pellets prepared. Both the partial order relative to the proton concentration (n = 0.39-0.41) and the apparent activation energy (E_app = 49 kJ mol⁻¹) were found in good agreement with the data reported for powdered samples showing that the sintering process does not degrade the chemical durability of the ceramics. Moreover, due to the high thermodynamical constant of complexation of phosphate species for tetravalent uranium and thorium, the influence of other ligands such as nitrate, chloride or sulphate on the normalized dissolution rates was limited. Near the equilibrium, the increasing of the leaching time, the temperature or the leachate acidity led to the thorium precipitation at the surface of the pellets either in static or in dynamic conditions. Consequently, the dissolution became clearly incongruent and controlled by saturation processes which are described in the second part of this paper.     

Separation of uranium from (U, Th)O2 and (U, Pu)O2 by solid state reactions route

Solid state reactions of UO₂, ThO₂, PuO₂ and their mixed oxides (U, Th)O₂ and (U, Pu)O₂ were carried out with sodium nitrate upto 900 °C, to study the formation of various phases at different temperatures, which are amenable for easy dissolution and separation of the actinide elements in dilute acid. Products formed by reacting unsintered as well as sintered UO₂ with NaNO₃ above 500 °C were readily soluble in 2 M HNO₃, whereas ThO₂ and PuO₂ did not react with NaNO₃ to form any soluble products. Thus reactions of mixed oxides (U, Th)O₂ and (U, Pu)O₂ with NaNO₃ were carried out to study the quantitative separation of U from (U, Th)O₂ and (U, Pu)O₂. X-ray diffraction, X-ray fluorescence, thermal analysis and chemical analysis techniques were used for the characterization of the products formed during the reactions.