Mixture Of UO2 Research Articles

Fabrication of mixed uranium–plutonium carbide fuel pellets with a low oxygen content and an open-pore microstructure

Mixed uranium–plutonium carbides are considered as potential fuels for Generation IV fast reactors. Compared to that of oxide fuels, their fabrication with regard to required specifications is more difficult. Carbides are obtained by a two-step procedure from the mixture of UO2, PuO2 and graphite powders. Due to their extreme reactivity with oxygen and moisture, fabrication and handling of carbide fuels are performed in glove boxes maintained under a dynamic flow of nitrogen providing operational safety. Nevertheless, oxygen pickup during processing of the material seems unavoidable and thus must be limited by suitable procedures. The oxygen content in the carbides can vary over a wide range of values, which can affect the green density and hence the sintered density. The addition of zinc stearate in a suitable amount to the fuel powder leads to an open-pore microstructure. A high open porosity contributes to minimizing the residual oxygen content in the sintered pellets.

UO2–UN composites with enhanced uranium density and thermal conductivity

A mixed ceramic composite composed of UO2–UN was studied to increase both the thermal conductivity and uranium density of the currently used UO2 pellet. UN powder was synthesized by hydriding and then nitriding spherical U metal powders. Disk-shaped mixed UO2–UN composites were fabricated by hot pressing UO2–UN powder mixtures within a temperature range of 1300 °C–1590 °C. A phase analysis was conducted using X-ray diffraction and a SEM analysis indicated that the sintered composite consists mainly of a mixture of UO2 and UN phases. Compared to pure UO2, the UO2-33vol% UN composite showed an increase in uranium density of 13% (10.91 g/cm3) and an increase in thermal conductivity of up to 100% (8.16 W/m k at 800 °C).

Thermodynamics of formation of coffinite, USiO4

Coffinite, USiO4, is an important U(IV) mineral, but its thermodynamic properties are not well-constrained. In this work, two different coffinite samples were synthesized under hydrothermal conditions and purified from a mixture of products. The enthalpy of formation was obtained by high-temperature oxide melt solution calorimetry. Coffinite is energetically metastable with respect to a mixture of UO2 (uraninite) and SiO2 (quartz) by 25.6 ± 3.9 kJ/mol. Its standard enthalpy of formation from the elements at 25 °C is -1,970.0 ± 4.2 kJ/mol. Decomposition of the two samples was characterized by X-ray diffraction and by thermogravimetry and differential scanning calorimetry coupled with mass spectrometric analysis of evolved gases. Coffinite slowly decomposes to U3O8 and SiO2 starting around 450 °C in air and thus has poor thermal stability in the ambient environment. The energetic metastability explains why coffinite cannot be synthesized directly from uraninite and quartz but can be made by low-temperature precipitation in aqueous and hydrothermal environments. These thermochemical constraints are in accord with observations of the occurrence of coffinite in nature and are relevant to spent nuclear fuel corrosion.

ChemInform Abstract: Trivalent Cation‐Controlled Phase Space of New U(IV) Fluorides, Na3MU6F30 (M: Al3+, Ga3+, Ti3+, V3+, Cr3+, Fe3+): Mild Hydrothermal Synthesis Including an in situ Reduction Step, Structures, Optical, and Magnetic Properties.

AbstractSingle crystals of the title compounds are hydrothermally synthesized in nearly quantitative yields from aqueous mixtures of UO2(OAc)2, NaF, HF, and Ti2O3, or V2O5, GaF3, or FeF2 for Na3MU6F30 (M: Ti, V, Ga, Fe), and from aqueous mixtures of NaUF5, HF, and Na3AlF6, or CrF3 for Na3MU6F30 (M: Al, Cr) (autoclave, 200 °C, 1 d).

Influence of Carbon Nanotube Dispersion in UO2–Carbon Nanotube Ceramic Matrix Composites Utilizing Spark Plasma Sintering

A novel uranium dioxide (UO2)–carbon nanotube (CNT) ceramic matrix composite fuel concept has been proposed for a nuclear fuel with increased thermal conductivity. Investigations were performed to analyze the dispersion of CNTs in a UO2 matrix utilizing homogenization and sonication techniques. Ethanol and ortho-dichlorobenzene (ODCB) were utilized as solvents during the mixing process. Distributions of both multiwalled carbon nanotubes and single-walled carbon nanotubes (SWNTs) were analyzed. It has been determined that CNTs can be homogeneously distributed into a UO2 matrix using mechanical processes, sonication, and homogenization in the organic solvent ODCB. The powder mixture of UO2 and CNTs was sintered at 1300°C with a hold time of 5 min and 40-MPa pressure in a spark plasma sintering furnace, and the resulting grain size distribution was analyzed. It was observed that where the distribution of CNTs was not well controlled, significant grain growth of UO2 occurred. However, where the CNT distribution is well controlled, the grain growth is limited by the pinning effect of the CNTs along the grain boundaries. The resulting pellet thermal conductivity was improved by 29.7% with the addition of 5 vol % SWNT, relative to pure UO2 values. Raman spectroscopy in conjunction with scanning electron microscopy shows that most CNTs survive both the mixing and sintering processes.

Thermal studies on the interaction behavior of ternary nitrate mixtures of UO2(NO3)2⋅6H2O–NaNO3–Sr(NO3)2 at elevated temperatures

Thermal and X-ray diffraction measurements have been used to study the formation of various nitrato-complexes as intermediates and mixed oxides as end products using different mol ratios of the ternary mixtures of UO2(NO3)2⋅6H2O, NaNO3 and Sr(NO3)2 at high temperatures. These results indicate the interaction of UO2(NO3)2⋅6H2O and NaNO3 in equimolar ratio to form sodium uranyl nitrate monohydrate (Na[UO2(NO3)3]⋅H2O), dimerized sodium uranyl oxynitrate (Na2[(UO2NO3)2(O)2]) and sodium diuranate (Na2U2O7). The decomposition temperature of Sr(NO3)2 decreased by 60°C in the presence of in-situ formed Na2U2O7. The uranium of Na2U2O7 intermediate reacts with Sr(NO3)2 to form various strontium uranium mixed oxides depending on Sr to U ratio in the initial mixtures. During the interaction of Sr(NO3)2 with Na2U2O7, sodium separates out as amorphous Na2O in the mixtures, containing Sr to U ratio ⩾1. Na2O phase is not formed in the uranium rich ternary mixture.

ChemInform Abstract: Topologically Identical, but Geometrically Isomeric Layers in Hydrous α‐, β‐Rb[UO2(AsO3OH)(AsO2 (OH)2)]·H2O and Anhydrous Rb[UO2(AsO3OH)(AsO2(OH)2)

AbstractYellow‐green block‐shaped crystals of RbUO2(AsO3OH)(AsO2(OH)2) (I) and green block‐shaped crystals of β‐Rb[UO2(AsO3OH)(AsO2 (OH)2)]·H2O (II) are simultaneously hydrothermally synthesized from an aqueous mixture of UO2(NO3)2, RbNO3, and As2O3 in the molar ratio of 1:9:10 (autoclave, 150 °C, 48 h).

ChemInform Abstract: Application of a Mild Hydrothermal Approach Containing an in situ Reduction Step to the Growth of Single Crystals of the Quaternary U(IV)‐Containing Fluorides Na4MU6F30 (M: Mn2+, Co2+, Ni2+, Cu2+, and Zn2+) Crystal Growth, Structures, and Magnetic Properties.

AbstractSingle crystals of the title compounds are hydrothermally prepared from aqueous mixtures of UO2(OAc)2, NaF, HF, and MCl2 (M: Mn, Co, Ni, Zn) or Cu(OAc)2 (autoclave, 200 °C, 1 d, 20, 70, 60, 80, and about 100% yield for Mn‐, Co‐, Ni‐, Zn‐, and Cu‐containing materials, respectively).

ChemInform Abstract: Elucidation of Tetraboric Acid with a New Borate Fundamental Building Block in a Chiral Uranyl Fluoroborate.

AbstractRb2 [(UO2)2B8O12F6] ·H2B4O7 is synthesized from mixtures of UO2(NO3)2, H3BO3, RbBF4, and H2O (autoclave, 220 °C, 5 d).

ChemInform Abstract: Hydrothermal Syntheses and Structures of the Uranyl Tellurates AgUO2(HTeO5) and Pb2UO2(TeO6).

AbstractThe title compounds are hydrothermally synthesized in low yields from mixtures of UO2(NO3)2, H6TeO6, and AgNO3 or Pb(NO3)2 (H2O, autoclave, 220 °C, 9 d).

Conversion of Ceramic UO2Pellets for HTR-10 Fuel Elements into Nitrate with N2O4

A 10 MW high-temperature gas-cooled reactor (HTR-10) has been constructed in China. It is important to study reprocessing of spent nuclear fuel from high-temperature gas-cooled reactor (HTGR) for recovery of nuclear resource and reduction of nuclear waste. Ceramic UO2 pellets of 0.5 mm in diameter for HTR-10 fuel elements were converted into nitrate with N2O4. When the molar ratio of water to uranium was 6, the conversion efficiency was above 99% after 180 min at 80 °C, and the conversion product was a mixture of UO2(NO3)2·3H2O and UO2(NO3)2·6H2O. When ceramic UO2 pellets were transformed to U3O8 powder with a particle size below 100 μm under an O2 stream and high temperature, the U3O8 powder was quickly converted into nitrate with N2O4 within 45 min. Water showed a significant effect on the reaction of U3O8 powder with N2O4, whereas other operation conditions had little effect on the reaction.

The effect of Y 2O 3 on the dynamics of oxidative dissolution of UO 2

In this work, we have studied the impact of Y 2O 3 on the kinetics of oxidative dissolution of UO 2 and the consumption of H 2O 2. The second order kinetics of catalytic consumption of H 2O 2 on Y 2O 3 was investigated in aqueous Y 2O 3 powder suspensions by varying the solid surface area to solution volume ratio. The resulting second order rate constant is 10 −8 m s −1, which is of the same magnitude as for the reaction between H 2O 2 and UO 2. Powder experiments with mixtures of UO 2 and Y 2O 3 show that Y 2O 3 has no effect on the oxidative dissolution of UO 2, whereas the consumption of H 2O 2 seems to be slightly slower in the presence of Y 2O 3 and H 2 respectively. UO 2 pellets with solid inclusions of Y 2O 3 show a decrease in oxidative dissolution by a factor of 3.3 and 5.3 under inert and hydrogen atmosphere, respectively. The rate of H 2O 2 consumption is similar for all cases and is well in line with kinetic data from powder experiments. The effects of H 2 and Y 2O 3 on the oxidative dissolution of UO 2 under gamma irradiation are similar to those found in experiments with H 2O 2. No significant difference in dissolution between inert and reducing atmosphere can be observed for pure UO 2.

Determination of PuO2% in power reactor mixed oxide fuel blends (0.4–44% PuO2) by neutron well co-incidence counting technique

Neutron Well Coincidence Counting has been explored as a non destructive assay technique for determining the percentage of PuO2 in blended mixture of UO2 and PuO2 powders. The method has been applied to MOX blends having PuO2 content varying from 0.4 to 44% for both thermal and fast reactors. The use of Neutron Well Coincidence Counting technique is validated as a process control step for determining PuO2 content in the fabrication of MOX fuel by comparing it with chemical analysis of sintered pellets. It has been used at Advanced Fuel fabrication Facility, Bhabha Atomic Research Centre, Tarapur during the manufacture of MOX fuel of various types for thermal and fast reactors.

High Temperature Interaction Between UO2 and Carbon: Application to TRISO Particles for Very High Temperature Reactors

For very high temperature reactors, the high level operating temperature of the fuel materials in normal and accidental conditions requires studying the possible chemical interaction between the UO2 fuel kernel and the surrounding structural materials (C, SiC) that could damage the tristructural isotropic particle. The partial pressures of the gaseous carbon oxides formed at the fuel (UO2)-buffer (C) interface leading to the build up of the internal pressure in the particle have to be predicted. A good knowledge of the phase diagram and thermodynamic properties of the uranium-carbon-oxygen (UCO) system is also required to optimize the fabrication process of “UCO” kernels made of a mixture of UO2 and UC2. Thermodynamic calculations using the FUELBASE database dedicated to Generation IV fuels (Guéneau, Chatain, Gossé, Rado, Rapaud, Lechelle, Dumas, and Chatillon, 2005, “A Thermodynamic Approach for Advanced Fuels of Gas Cooled Reactors,” J. Nucl. Mater., 344, pp. 191–197) allow predicting the phase equilibria involving carbide and/or oxycarbide phases at high temperature. Very high levels of CO(g) and CO2(g) equilibrium pressures are obtained above the UO2±x fuel in equilibrium with carbon that could lead to the failure of the particle in case of high oxygen stoichiometry of the uranium dioxide. To determine the deviation from thermodynamic equilibrium, measurements of the partial pressures of CO(g) and CO2(g) resulting from the UO2/C interaction have been performed by high temperature mass spectrometry on two types of samples: (i) pellets made of a mixture of UO2 and C powders or (ii) UO2 kernels embedded in carbon powder. Kinetics of the CO(g) and CO2(g) as a function of time and temperature was determined. The measured pressures are significantly lower than the equilibrium ones predicted by thermodynamic calculations. The major gaseous product is always CO(g), which starts to be released at 1473 K. From the analysis of the partial pressure profiles as a function of time and temperature, rates of CO(g) formation have been assessed. The influence of the different geometries of the samples is shown. The factors that limit the gas release can be related to interface or diffusion processes as a function of the type of sample. The present results show the utmost importance of kinetic factors that govern the UO2/C interaction.

On the Fuel and Coolant Interaction Behavior of Partially Oxidized Corium

To simulate a fuel and coolant interaction phenomenon during a postulated severe accident in a nuclear reactor, a series of experiments were performed using a partially oxidized corium, which is a mixture of UO2, ZrO2, Zr, and stainless steel. The composition of the melt was chosen such that a separation of the oxidic liquid from the metallic liquid occurred due to the existence of a miscibility gap. A melting and solidifying experiment and two fuel and coolant interaction experiments to explore the possibility of an energetic steam explosion were performed in the TROI facility.The placement of a metal-rich layer consisting of U, Fe, and ZrO2 beneath the oxidic corium layer due to the existence of a miscibility gap was observed in the melting and solidifying experiment. An energetic steam explosion with a propagation of the dynamic pressure wave was observed in one test out of the two tests. The physical and chemical analyses were performed for the corium particles collected after the experiments. It is shown that U, Zr, and Fe formed a heterogeneous mixture and the morphology was in irregular shape with many pores at nonuniform sizes. In the case of nonenergetic interaction, where the melt temperature was lower than the energetic case, the mean particle size was bigger than that of the energetic case, and the melt-water interaction resulted in a substantial amount of hydrogen gas generation, while the amount of hydrogen gas generation was negligible in the case with an energetic steam explosion.

An electromagnetic and thermal analysis of a cold crucible melting

To investigate the performance of a cold crucible employed for the melting of corium, which is a mixture of UO 2 and ZrO 2, a computational analysis of the coupled electro-magnetic field, heat transfer, and fluid flow is carried out. Governing differential equations, basic numerical methods, computational model, and the results of the simulation of cold crucible melting are discussed. By comparing the analysis results with the experimental data, it is shown that the proposed computational model reasonably predicts the fundamental characteristics of a cold crucible melting.

Insights from the Recent Steam Explosion Experiments in TROI

The paper discusses the results of steam explosion experiments of TROI-13, TROI-14, and TROI-15, which were performed under the research program named “Test for Real cOrium Interaction with water (TROI).” TROI-13 and TROI-14 used corium, which is a mixture of UO2 and ZrO2 at a 70:30wt%, while TROI-15 used ZrO2. These three cases resulted in spontaneous steam explosions. It is an important observation from the aspect of the explosivity of prototypic material, as it was not observed in the previous experiments. Various aspects of the test results including cold crucible melting, hydrogen generation, melt temperature measurement, debris morphology, size distribution of the debris, dynamic pressure, dynamic force, shape of the melt jet, location of the trigger, response of the vessel pressure, and the response of the water pool temperature are discussed. The potential contributors to the explosivity of corium are suggested.

Insights from the Recent Steam Explosion Experiments in TROI

The paper discusses the results of steam explosion experiments of TROI-13, TROI-14, and TROI-15, which were performed under the research program named “Test for Real cOrium Interaction with water (TROI).” TROI-13 and TROI-14 used corium, which is a mixture of UO2 and ZrO2 at a 70:30wt%, while TROI-15 used ZrO2. These three cases resulted in spontaneous steam explosions. It is an important observation from the aspect of the explosivity of prototypic material, as it was not observed in the previous experiments. Various aspects of the test results including cold crucible melting, hydrogen generation, melt temperature measurement, debris morphology, size distribution of the debris, dynamic pressure, dynamic force, shape of the melt jet, location of the trigger, response of the vessel pressure, and the response of the water pool temperature are discussed. The potential contributors to the explosivity of corium are suggested.

Dissolution of uranium, neptunium, plutonium and americium oxides in tri- n-butyl phosphate saturated with nitric acid

Uranium dioxide and its solid solutions with neptunium, plutonium and americium dioxides can be efficiently and quantitatively dissolved using tri-n-butyl phosphate (TBP) saturated with nitric acid. Individual NpO2 and PuO2 do not dissolve under these conditions. On the treatment of a mechanical mixture of UO2 with PuO2 and NpO2 with TBP–HNO3, a complex of uramiun completely dissolved, whereas Np and Pu remained in the residue.

Dissolution of U, Np, Pu, and Am Oxides and Their Mixtures in Tributyl Phosphate Saturated with HNO 3

It was shown for the first time that weighable amounts of uranium dioxide and its mixtures with Np, Pu, and Am in the form of solid solutions can be efficiently and quantitatively dissolved in the presence of tributyl phosphate saturated with HNO3. Individual PuO2 and NpO2 are not dissolved under these conditions. In treatment of a mechanical mixture of UO2 with PuO2 or NpO2, uranium is completely dissolved, while plutonium and neptunium remain in the precipitate.