Characterization of the reaction layer in U-7wt%Mo/Al diffusion couples

The reaction layer in chemical diffusion couples U-7wt%Mo/Al was investigated using optical and scanning electron microscopy, electron probe microanalysis and X-ray diffraction (XRD) techniques. When the U-7wt%Mo alloy was previously homogenized and the γ(U, Mo) phase was retained, the formation of (U, Mo)Al₃ and (U, Mo)Al₄ was observed at 580 °C. Also a very thin band was detected close to the Al side, the structure of the ternary compound Al₂₀UMo₂ might be assigned to it. When the decomposition of the γ(U, Mo) took place, a drastic change in the diffusion behavior was observed. In this case, XRD indicated the presence of phases with the structures of (U, Mo)Al₃, Al₄₃U₆Mo₄, γ(U, Mo) and α(U) in the reaction layer.     

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.     

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.     

Measurement of the specific heat of Zr–40 wt%U metallic fuel

The specific heat capacities of un-irradiated and irradiated metallic Zr–40 wt%U fuel have been measured between 50 °C and 1000 °C with a differential scanning calorimetry. The irradiated fuels have three different burnup levels of 0.38, 0.70 and 0.92 g-fission product (FP)/cm³. The measured specific heat for the un-irradiated fuel is representative and consistent with the values estimated from the Neumann–Kopp rule. The irradiated fuels exhibited a complicated behavior of the heat capacities. The unique characteristics of the specific heat capacities can be explained by the recovery of radiation damage, the formation of fission gas bubbles and fission gas release, and a phase transition in the irradiated fuels. An examination of the microstructure revealed that multiple large bubbles were formed in the irradiated fuel during specific heat measurement. The measured specific heat is expected to enable us to estimate the stored energy in the metallic fuel during certain accident scenarios and to determine the thermal conductivity of zirconium–uranium metallic fuel.     

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.     

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.     

Development of anodic stripping voltammetry for determination of gallium in U–Ga alloy

Deposition potential, deposition time, square-wave frequency, rotation speed of the rotating disc electrode and gallium concentration have been studied in detail, for trace concentration level determination of gallium metal in U–Ga alloy by square-wave voltammetry anodic stripping analysis, in 1 M NaClO₄ + 0.5 M NaSCN at mercury film electrode (MFE). Optimum conditions have been found for Ga(III) determination by obtaining calibration graphs for the range 1–10 × 10⁻⁷ M gallium. Error and standard deviation less than 1% were assessed of this method with all gallium standard solutions. The developed methodology was applied successfully as a subsidiary method for the determination of gallium content in synthetic U–Ga samples with very good precision and accuracy (under 1% error and std. dev.).     

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.     

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.     

Constituent redistribution in U-Pu-Zr fuel during irradiation

The thermo-migration fluxes of U, Pu and Zr in U-Pu-Zr metallic alloy fuel during irradiation in the Experimental Breeder Reactor II (EBR-II) were calculated using the constituent redistribution profiles measured in postirradiation examinations. Based on these fluxes, the diffusion coefficients, and the sums of heat of transport and enthalpy of solution for the γ, γ+ζ and δ+ζ phases in U-Pu-Zr were obtained. Using these data, the predicted concentration redistribution profiles are consistent with the measurements. The effect of minor actinide (Am and Np) addition was also examined. Am migration generally followed that of Zr with local precipitation, while Np behaved similarly to Pu.     

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.     

Determining the shear properties of the PyC/SiC interface for a model TRISO fuel

The fracture behavior of TRISO-coated fuel particles is dependent on the shear strength of the interface between the inner pyrolytic carbon (PyC) and silicon carbide coatings. This study evaluates the interfacial shear properties and the crack extension mechanism for TRISO-coated model tubes using a push-out technique. The interfacial debond shear strength was found to increase with increasing sample thickness and finally approached a constant value. The intrinsic interfacial debond shear strength of ∼280 MPa was estimated. After the layer is debonded, the applied load is primarily transferred by interfacial friction. A non-linear shear-lag model predicts that the residual clamping stress at the interface is ∼350 MPa, and the coefficient of friction is ∼0.23, yielding a frictional stress of ∼80 MPa. These relatively high values are attributed to the interfacial roughness. Of importance in these findings is that this unusually high interfacial strength could allow significant loads to be transferred between the inner PyC and SiC in application, potentially leading to failure of the SiC layer.     

Behavior of molybdenum in pyrochemical reprocessing: A spectroscopic study of the chlorination of molybdenum and its oxides in chloride melts

The high temperature reactions of molybdenum and its oxides with chlorine and hydrogen chloride in molten alkali metal chlorides were investigated between 400 and 700 °C. The melts studied were LiCl-KCl, NaCl-CsCl and NaCl-KCl and the reactions were followed by in situ electronic absorption spectroscopy measurements. In these melts Mo reacts with Cl₂ and initially produces MoCl₆²⁻ and then a mixture of Mo(III) and Mo(V) chlorocomplexes, the final proportion depending on the reaction conditions. The Mo(V) content can be removed as MoCl₅ from the melt under vacuum or be reduced to Mo(III) by Mo metal. The reaction of Mo when HCl gas is bubbled into alkali chloride melts yields only MoCl₆³⁻. MoO₂ reacts in these melts with chlorine to form soluble MoOCl₅²⁻ and volatile MoO₂Cl₂. MoO₃ is soluble in chloride melts and then decomposes into the oxychloride MoO₂Cl₂, which sublimes or can be sparged from the melt, and molybdate. Pyrochemical reprocessing can thus be employed for molybdenum since, after various intermediates, the end-products are chloride melts containing chloro and oxychloro anions of molybdenum plus molybdate, and volatile chlorides and oxychlorides that can be readily separated off. The reactions were fastest in the NaCl-KCl melt. The X-ray diffraction pattern of MoO₂Cl₂ is reported for the first time.     

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™.     

Power-to-melts of uranium-plutonium oxide fuel pins at a beginning-of-life condition in the experimental fast reactor JOYO

Power-to-melts of uranium-plutonium oxide fuel pins at an initial startup condition were experimentally obtained from the B5D-2 test in the experimental fast reactor JOYO in Oarai Engineering Center. MCNP code calculations were combined with burnup measurements to determine linear heat rating of the test fuel pins. To identify the axial incipient melting positions corresponding to the power-to-melts, solidified grain morphology and molten fuel axial movements were characterized. Extensive observations on longitudinal ceramographs allowed classifying molten fuel settlements near bottom and top extents of axial fuel melting into three types. The power-to-melts depended slightly on fuel-to-cladding gap sizes and clearly on both oxygen-to-metal ratios and densities of fuel pellets. These dependencies resulted from the fuel pellet cracking and relocation behavior, which fairly improves heat transfers across the gaps. Also, the power-to-melt at the bottom position was higher than that at the top position due to an axial gradient of cladding temperatures in each fuel pin.     

Estimation of maximum coated particle fuel compact packing fraction

This work develops an analytic fuel fraction packing model for a high temperature gas cooled reactor fuel compact fabricated from overcoated particles of a single size. The model includes the effects of one dimensional compression and finite matrix grain size. One dimensional compression limits the maximum fuel packing fraction to about 48% for the pressed compact in this single sized particle system. This limit is due to two effects. The first is that the process of die loading limits the pre-compression packing configuration to one that is stable under gravity, which is not the most space efficient one. The second effect is due to the one dimensional compression which reduces only the axial dimension of the particle lattice rather than uniformly compressing the lattice. The die wall can also limit the maximum packing fraction by preventing the nearby particles from moving into a more space efficient configuration.     

Influence on the structural characteristic of defect solid solution GdxZr1−xO2−x/2 with 0.18 ? x ? 0.62 under longer term annealing studied by Gd Mössbauer spectroscopy

Three kinds of defect solid solution Gd_xZr_1−xO_2−x/2 with 0.18 ? x ? 0.62, including the three single crystal samples with x = 0.21, 0.26 and 0.30, were investigated by ¹⁵⁵Gd Mössbauer spectroscopy at 12 K. Difference in the structural characteristic under longer term annealing were confirmed by comparing the ¹⁵⁵Gd Mössbauer parameters of the polycrystalline samples sintered one time and twice at 1773 K for 16 h in air, respectively. The results indicated that the polycrystalline samples sintered twice have relatively equilibrated structure by comparing with the three single crystal samples. After being sintered twice, basically the local structure around the Gd³⁺ ions does not change, but the degree of the displacements of the six 48f oxygen ions from positions of cubic symmetry becomes slightly smaller, and distribution of the Gd³⁺ ions in the system becomes more homogeneous.     

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.     

Fuel oxidation and thermal conductivity model for operating defective fuel rods

A model has been developed to describe the fuel oxidation behaviour, and its influence on the fuel thermal conductivity, in operating defective nuclear fuel rods. The fuel-oxidation model is derived from adsorption theory and considers the influence of the high-pressure environment that results from coolant entry into the fuel-to-clad gap. This model is in agreement with the fuel-oxidation kinetics observed in high-temperature annealing experiments conducted at 1473-1623 K in steam over a range of pressure from 0.001 to 0.1 MPa. Using a Freundlich adsorption isotherm, the current model is also consistent with recent experiments conducted at a higher pressure of 7 MPa. The model also considers radiolytic effects as a consequence of fission fragment bombardment in the fuel-to-clad gap. This treatment suggests that radiolysis-assisted oxidation is insignificant in operating defective rods (as compared to thermal effects), as supported by limited in-reactor data. The effects of diffusion of the interstitial oxygen ions in the solid in the operating rod is further discussed.     

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.