Release of fission products (Xe, Kr, I, Cs) implanted in polycrystalline UO2

International audienceUnder irradiations in nuclear reactors, the microstructure of oxide nuclear fuel changes. To improve the modeling of the UO2 fuel behavior under irradiation, it is necessary to understand the elementary mechanisms of fission products diffusion. Among them, rare gas Xenon and Krypton represent 30% of created elements and fission products such as Iodine and Caesium are corrosive for the clad. Our experimental work consists in the measurement of the fission products release kinetics by Knudsen cell mass spectrometry. For that, 8mm diameter-1 mm height fresh polycrystalline UO2 pellets are implanted with different concentrations in 129Xe, 83Kr, 127I, 133Cs to understand the effect of the fission products density on the diffusion. The release kinetics is studied either during the heating at a given rate from room temperature to about 2050DC (Figure) or during isothermal annealing every 100 DC from 100 to 1600 DC

Paper 58148 HIGH TEMPERATURE INTERACTION BETWEEN UO 2 AND CARBON: APPLICATION TO TRISO PARTICLES FOR VERY HIGH TEMPERATURE REACTORS CEA Marcoule -DEN/MAR/DTEC/SGCS/LMAC 30207 Bagnols-sur-Cèze, France

ABSTRACT For Very High Temperature Reactors (V-HTR), the study of the Uranium-Carbon-Oxygen system is of major importance to predict the high temperature behaviour of the TRISO fuel particle. Firstly, the high level operating temperature of the fuel materials in normal and accidental conditions requires studying the possible chemical interaction between the UO 2 fuel kernel and the surrounding structural materials (C, SiC) that could damage the particle. The formation of the gaseous carbon oxides at the fuel (UO 2 )-buffer (C) interface that leads to the build up of the internal pressure in the particle has to be predicted. Secondly, the U-C-O ternary system is also involved in the fabrication process of "UCO" kernels made of a mixture of UO 2 and UC 2 . For the fabrication of such mixture of uranium oxide and carbide, the phase diagram and thermodynamic properties of the U-C-O system are necessary to investigate in order to perform adequate heat treatments. For both reasons, a new study of the U-C-O ternary system has been undertaken. Firstly, some thermodynamic calculations (equilibrium CO (g) and CO 2(g) pressures, phase diagrams) were performed using the thermodynamic FUELBASE database dedicated to generation IV fuels In a second step, the partial pressures of CO (g) and CO 2(g) resulting from the UO 2 /C interaction have been measured by high temperature mass spectrometry. Two types of samples were used (i) pellets made of a mixture of UO 2 and C powders or (ii) UO 2 kernels disseminated in a carbon bed. The kinetic measurements of the release of CO (g) and CO 2(g) lead to measured pressures that are lower than the equilibrium pressures predicted from thermodynamic calculations. This discrepancy can be explained by limitations due to distinct kinetic mechanisms. Rates of CO (g) formation have been established taking into account the oxygen stoichiometry of uranium oxide and temperature. The major gaseous product is always CO (g) which release significantly starts at 1473 K. The influence of the different geometries is shown. The limitative kinetic step can be an interface or a diffusion process as a function of the type of sample. These results underline the up most importance of kinetic factors for studying the UO 2 / C interaction to determine realistic CO (g) pressure levels inside a TRISO particle or to improve the fabrication process of the "UCO" kernels

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

International audienceFor 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 (Gueneau, Chatain, Gosse, 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

Etude expérimentale des transitions de phases sur le côté riche en O du système Zr-O

International audienc

Thermodynamic study of the uranium–vanadium system

International audienceTemperatures of solid/liquid transitions and vanadium thermodynamic activity data are measured in the U–V system to improve the thermodynamic description of the U–V and C–U–V systems. Binary alloys are synthesized from the pure metals in a high vacuum furnace. With that apparatus, both liquidus temperatures and vanadium activities are measured for each sample. During the experiments, the temperature of the samples is monitored with an optical pyrometer. In parallel, the activity of vanadium referred to pure vanadium is measured for xV = 0.18, 0.40 and 0.62 in the (1850 to 2090) K temperature range using high temperature mass spectrometry coupled to a multiple Knudsen cell system. The quenched microstructure of the alloys is analysed by electron microscopy. These new data together with the few ones from the literature are finally used to obtain a consistent set of thermodynamic parameters for the U-V system using the Calphad method

ATTILHA : un nouvel outil pour l'acquisition de données thermodynamiques à haute température

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Experimental and thermodynamic calculations results on pwr and srf corium subsystems

International audienceDuring a severe accident in a pressurized water reactor (PWR), the nuclear oxide fuel (UO2 or (U,Pu)O2) could react with Zircaloy cladding, Inconel spacer grids, the steel controls rods cladding and the neutronic absorbers (Ag-Cd-In-B4C), leading to the relocation in the lower-head of the reactor vessel of a mixture of liquid and solid phases called in-vessel corium. If the reactor vessel is lost, the molten core can pour onto the containment concrete leading to the formation of the ex-vessel corium. As a first approximation and considering the major components, the system U-Pu-Zr-Fe-Al-Ca-Si-O can be considered as representative of the ex-vessel corium. At the Laboratory of Modelling, Thermodynamics and Thermochemistry, CEA Saclay, a series of experimental results have been achieved using ATTILHA, the novel experimental setup developed at the laboratory. These data are fundamental for the development of a reliable thermodynamic database for corium. Thermodynamic calculations have been performed to better interpret our experimental results. These calculations allow also to study the evolution of corium behavior varying different parameters, as for example the nature of the atmosphere (reducing/oxidizing), to reproduce different scenarios. Using the thermodynamic database of the representative corium developed at the laboratory, it is also possible to estimate the partition of elements in each phase as a function of temperature and composition. It can be shown that at 3000 K, in presence of a miscibility gap in the liquid phase, Pu preferentially segregate in one of the two immiscible liquids, namely in the oxide liquid

Development of ATTILHA - Experimental setup for high temperature investigations of U-containing systems (…soon)

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ATTILHA A novel experimental setup for thermodynamic and thermophysical properties measurements on nuclear materials

International audienceIn order to obtain experimental data on the complex corium pool involved during a severe accident in a nuclear reactor as a first approximation U-Zr-Fe-O experiments at very high temperatures have to be performed (T>2300 K). However, when samples are in direct contact with the instrumentation (e.g., thermocouples) or with a crucible, inevitable chemical interactions affect the experimental measurements. Furthermore, handling radioactive materials impose radioprotection restrictions.In this framework, a novel experimental setup called ATTILHA has been conceived and developed CEA Saclay. The setup has the final objective to study the high temperature thermodynamics and thermophysical properties of liquid nuclear materials. This apparatus is based on a laser heating technique coupled with contactless temperature monitoring and an aerodynamic levitation system. Spherical samples levitate within a controlled gas flux out of an Al levitation nozzle. Experimental gaseous conditions reducing or oxidizing are suited choosing the carrier levitation gas. The setup may be placed inside an and#945;-shield glove-box to comply with radioprotection limitations.In a first step, the setup has been used to investigate the liquid miscibility gap in the Fe-Zr-O ternary system a tie-line has been obtained between a metallic liquid enriched in Fe and an oxide liquid of composition close to ZrO2. The solubility of Fe in ZrO2 liquid has been therefore quantified. UO2 will be used to validate the setup with radioactive materials.To validate the setup for thermal-physical properties acquisition, experiments have been conducted first on non-radioactive materials, namely liquid Al2O3 and ZrO2. Forced oscillations were transmitted through the levitation gas by an acoustic system. Surface tension and density were obtained using an ultra-high speed camera coupled with a Python code, which allows to detect the resonant oscillations of the levitating liquid sample on the camera images