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.     

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.     

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.     

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.     

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.     

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.     

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.     

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.%.     

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.     

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.     

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.     

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.     

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.     

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.     

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.     

Radionuclide release from high burnup spent fuel during corrosion in salt brine in the presence of hydrogen overpressure

In the case of a contact between groundwater and Fe-based spent fuel disposal containers in a repository large amounts of hydrogen will be produced by the corrosion of iron, which may result in significant hydrogen pressures. To quantify to what extent the hydrogen overpressure may counteract radiolysis enhanced matrix dissolution, related experimental work has been performed. High burnup spent fuel was corroded in 5.6 mol (kg H₂O)⁻¹ NaCl solution applying H₂ overpressures (experimental set 1) <0.17 bar by radiolysis, (experimental set 2) 2.8 bar by Fe corrosion, (experimental set 3) 3.2 bar by external H₂ gas. In the absence of Fe (experimental set 3) the UO₂ matrix dissolution rate decreased by a factor of about 10. In this test the concentrations of U, Np, Tc in solution were found to be decreasing by at least two orders of magnitude, and ranging within the same level as in the presence of Fe powder (experimental set 2). However, Pu and Am concentrations (experimental set 3) were less affected, due to the high sorption capacity for these radioelements onto Fe corrosion products.     

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™.     

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.     

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.