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.     

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.     

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 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.     

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.%.     

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.     

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 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.     

X-ray and thermal studies of the Na-U-Mo-O system

In the Na-U-Mo-O system, five compounds with composition Na₂UMo₂O₁₀, Na₂U₂Mo₂O₁₃, Na₂U₂Mo₃O₁₆, Na₂UMo₄O₁₆ and Na₂U₂Mo₄O₁₉ were prepared by solid state reaction of Na₂MoO₄, UO₃ and MoO₃ in the required stoichiometric ratio. The compounds were characterized by X-ray powder diffraction, infrared and thermal analysis techniques. The XRD data of all the above-mentioned compounds were indexed on the orthorhombic system. All the compounds showed thermal stability up to 600 °C in air and decomposed at 950 °C to form Na₂U₂O₇. Infrared spectra of all the compounds show strong spectral bands in the range 700-950 cm⁻¹ due to tetrahedra and the group. A pseudo-ternary phase diagram of Na₂O-UO₃-MoO₃ was drawn using the quaternary compounds and information available on Na-U-O, Mo-U-O and Na-Mo-O ternary systems. The various phase fields prepared during this work were established by XRD analysis.     

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.     

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.     

Post-irradiation examination of uranium-7 wt% molybdenum atomized dispersion fuel

Two low-enriched uranium fuel plates consisting of U-7wt%Mo atomized powder dispersed in an aluminum matrix, have been irradiated in the FUTURE irradiation rig of the BR2 reactor at SCK•CEN. The plates were submitted to a heat flux of maximum 353 W/cm² while the surface cladding temperature is kept below 130 °C. After 40 full power days, visual examination and profilometry of the fuel plates revealed an increase of the plate thickness. In view of this observation, the irradiation campaign was prematurely stopped and the fuel plates were retrieved from the reactor, having at their end-of-life a maximum burn-up of 32.8% ²³⁵U (6.5% FIMA). The microstructure of one of the fuel plates has been characterized in an extensive post-irradiation campaign. The U(Mo) fuel particles have been found to interact with the Al matrix, resulting in an interaction layer which can be identified as (U,Mo)Al₃ and (U,Mo)Al₄. Based on the composition of the interaction layer it is shown that the observed physical parameters like thickness of the interaction layer between the Al matrix and the U(Mo) fuel particles compare well to the values calculated by the MAIA code, an U(Mo) behavior modeling code developed by the Commissariat à l’énergie atomique (CEA).     

Correlation of yield stress and microhardness in 08Cr16Ni11Mo3 stainless steel irradiated to high dose in the BN-350 fast reactor

The relationship between the microhardness and the engineering yield stress in 08Cr16Ni11Mo3 steel after irradiation in the BN-350 reactor has been experimentally derived and agrees with a previously published correlation developed by Toloczko for unirradiated 316 in a variety of cold-work conditions. Even more importantly, when the correlation is derived in the K_Δ format where the correlation involves changes in the two properties, excellent agreement is found with a universal K_Δ correlation developed by Busby and coworkers. Additionally, this report points out that microhardness measurements must take into account that sodium exposure at high temperature and neutron fluence alters the metal surface to produce ferrite, and therefore the altered layers should be removed prior to testing.     

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.     

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.     

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.).     

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™.     

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.     

Change in flow stress and ductility of δ-phase Pu-Ga alloys due to self-irradiation damage

An internal state variable model for the mechanical behavior of aged Pu-Ga alloys is developed and used to predict the change of the material with accumulated self-irradiation damage or age. The material model incorporates microstructural data such as the primary irradiation-induced defect density from cascades, the density and average size of helium bubbles, the initial dislocation density, and the initial average segment length of the dislocation density as input parameters, and then evaluates the stress-strain response of a representative volume element of the material. Given this response at a single material point, the deformation behavior of tensile specimens is predicted, and it forecasts increased strength, decreased strain hardening, and more strain localization with aging. Although the material point behavior showed some slight strain softening, this strain softening is expected to be masked by statistical variations of different volume elements and by the strain rate sensitivity of the material. Hence, it is not expected to appear in the stress-strain response of macroscopic tensile specimens, and only the increase in flow stress will be measured.     

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.