Thermodynamic activity measurements in nickel-base industrial alloys and steels by Knudsen cell – Mass spectrometry

International audienc

Etude de la solubilité de l uranium dans la tchernobylite (Zr,U)SiO4

International audienceLors de l accident grave de Tchernobyl, la perte de contrôle du réacteur a conduit à la production d une grande quantité de chaleur qui a causé l interaction à très haute température (>2400-2600°C) entre le combustible UO2, la gaine d alliage de zirconium et les structures en acier pour former un mélange de phases solides et liquides appelé « corium », principalement constitué d oxyde mixte (U,Zr)O2 et d inclusions d acier [13Pol]. Suite à l explosion, ces fragments de corium ont interagi avec les matériaux riches en SiO2 (béton, sable) à une température d environ 1500-1600°C pour former des phases liquides riches en silice, très radioactives, appelées « Chernobyl lava ». Parmi les phases formées, la tchernobylite (Zr,U)SiO4 a été observée. L objectif de notre étude est de déterminer la solubilité de l uranium dans la tchernobylite pour améliorer la modélisation thermodynamique du système ZrO2-UO2-SiO2 dans les bases de données thermodynamiques qui sont utilisées pour interpréter et prédire la formation des phases dans le corium. Des traitements thermiques de deux semaines à 1600°C ont été effectués sur trois échantillons synthétisés à partir de mélanges de poudres. Des essais de chauffage laser ont également été réalisés pour mesurer les températures de transition de phases sur les mêmes échantillons. Les résultats seront présentés ainsi que la caractérisation des microstructures obtenues. [13Pol] P. Pölm et al, J. Nucl. Mater. 439 (2013) 51-5

Direct Measurements of the Chromium Activity in Complex Nickel Base Alloys by High Temperature Mass Spectrometry

International audienceChromium rich, nickel based alloys Haynes 230 and Inconel 617 are candidate materials for the primary circuit and intermediate heat exchangers (IHX) of (Very)-High Temperature Reactors. The corrosion resistance of these alloys is strongly related to the reactivity of chromium in the reactor specific environment (high temperature, impure helium). At intermediate temperature – 900°C for Haynes 230 and 850°C for Inconel 617 – the alloys under investigation are likely to develop a chromium-rich surface oxide scale. This layer protects from the exchanges with the surrounding medium and thus prevents against intensive corrosion processes. However at higher temperatures, it was shown that the surface chromia can be reduced by reaction with the carbon from the alloy [1] and the bare material can quickly corrode. Chromium appears to be a key element in this surface scale reactivity. Then, quantitative assessment of the surface requires an accurate knowledge of the chromium activity in the temperature range close to the operating conditions (T ≈ 1273 K). High temperature mass spectrometry (HTMS) coupled to multiple effusion Knudsen cells was successfully used to measure the chromium activity in Inconel 617 and Haynes 230 in the 1423- 1548 K temperature range. Appropriate adjustments of the experimental parameters and in-situ calibration toward pure chromium allow to reach accuracy better than ± 5%. For both alloys, the chromium activities are determined. Our experimental results on Inconel 617 are in disagreement with the data published by Hilpert [2]. Possible explanations for the significant discrepancy are discussed

Chemical interaction between uranium dioxide, boron carbide and stainless steel at 1900 °C — Application to a severe accident scenario in sodium cooled fast reactors

International audienceFor the understanding of severe accidents in sodium cooled fast reactors (SFR), it is necessary to understand two prototypic accident scenarios such as ULOF (Unprotected Loss of Flow Accident) and UTOP (Unprotected Transient OverPower). As the base knowledge, it is also important to understand high temperature chemical interaction among major core materials such as MOx fuel (MOx: mixed oxide of uranium and plutonium), steel cladding and B4C neutron absorber have to be investigated. This study aims at providing experimental data on phase formation and phase-stability at various temperature and pressure conditions. A first series of samples containing a mixture of B4C and steel were prepared to obtain a homogenous metallic solid. In a second step, these metallic samples were mixed and melted with small UO2 pieces by arc melting. Then these samples underwent a heat treatment at 1900 °C for 1 hour. EDS, EBSD and EPMA analyses were performed to identify the phases formed during the solidification. In addition, thermodynamic calculations were performed for the interpretation of the results, revealing that a carbo-reduction reaction occurs: UO2 + 2 C = 2 CO + U. A significant amount of uranium from the fuel is dissolved in the metallic liquid phase, leading to the formation of mixed borides (UM3B2, UMB4, UM4B, M=Fe,Cr,Ni). In comparison with the UO2/steel interaction, the present results show that the presence of B and C in the melt improves the wetting behaviour of the metallic liquid towards UO2

Fission Products Chemistry in Simulated PWR fuel up to 2100 °C: Experimental Characterisation and TAF-ID Modelling

International audienceThe capabilities of the Thermodynamic of Advanced Fuels - International Database (TAF-ID) for reproducing the chemical behaviour of irradiated nuclear fuel in severe accident conditions are studied in this work, by comparing calculation results to experimental observations. To this end, Simfuel samples that are representative of Pressurized Water Reactors (PWR) fuel irradiated up to 76 GWd•tU-1 were submitted to 1327 °C in oxidizing conditions, and to 1800, 2000, and 2100 °C under reducing ones. Characterization of the UO2 fuel and Fission Products (FP) surrogates (Ba, Ce, La, Mo, Nd, Pd, Rh, Ru, Sr, and Zr) was done by Electron Probe Micro Analysis (EPMA), High-Resolution X-Ray Diffraction (HR-XRD), and X-Ray Absorption Spectroscopy (XAS). Experimental characterization confirmed (i) the solubility of La, Y, Nd, Ce and to some extend of Sr and Zr in the UO2 fluorite matrix; (ii) the presence of a perovskite phase, (Ba,Sr)(Ce,Mo,Zr,U)O3, in all the samples but which composition is highly dependent on experimental conditions; (iii) the presence of metallic U in reducing conditions, in association with Ru, Rh, and Pd; and (iv) metallic Mo precipitates, in association with Ru and Pd. In addition, a (U,Zr)O2 phase with a non-homogeneous distribution of U and Zr was observed in reducing conditions. Thermodynamic calculations not only reproduce very well the observed phases and the evolution of their composition with temperature and atmosphere, but also allow explaining their microstructure as a result of the progressive solidification of liquids. In addition, calculations explain the non-homogenous distribution of elements in the (U,Zr)O2 phase, which is attributed to the separation of a homogeneous fluorite phase at high temperature into fluorite and monoclinic-ZrO2 phases at low temperature. However, some features such as the U-Pd-Rh association could not be reproduced by calculations, due to the lack of a thermodynamic model for this ternary system. Altogether, these results demonstrate the great potential of the TAF-ID for the prediction of irradiated fuel behaviour in normal and off-normal reactor operating conditions