The influences of Pu and Zr on the melting temperatures of the UO2–PuO2–ZrO2pseudo-ternary system (original) (raw)
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Journal of Nuclear Materials, 2005
In the framework of the severe accident R&D studies led by CEA, a better knowledge of ÔcoriumÕ, mainly a UO 2 -ZrO 2 mixture, which would result from the melting of a nuclear reactor core, is fundamental to mastering this kind of hypothetical accident. Among the available corium physical characteristics, the knowledge of the final crystalline compounds and their quantities gives basic information about the physico-chemistry mechanisms and the solidification path (thermodynamic calculation). Meta-stable phases in the UO 2 -ZrO 2 system (tetragonal solid solution), which appear during transient phases of severe accidents, have been observed in a recent real material experiment. A specific XRD (X-ray diffraction) approach has been developed for corium analysis coupling identification of the solid solutions and Rietveld method for quantitative data. Similar transient phases had been characterized on samples coming from TMI-2 nuclear reactor corium samples.
Journal of Nuclear Materials, 2005
In the framework of the severe accident R&D studies led by CEA, a better knowledge of ÔcoriumÕ, mainly a UO 2-ZrO 2 mixture, which would result from the melting of a nuclear reactor core, is fundamental to mastering this kind of hypothetical accident. Among the available corium physical characteristics, the knowledge of the final crystalline compounds and their quantities gives basic information about the physico-chemistry mechanisms and the solidification path (thermodynamic calculation). Meta-stable phases in the UO 2-ZrO 2 system (tetragonal solid solution), which appear during transient phases of severe accidents, have been observed in a recent real material experiment. A specific XRD (X-ray diffraction) approach has been developed for corium analysis coupling identification of the solid solutions and Rietveld method for quantitative data. Similar transient phases had been characterized on samples coming from TMI-2 nuclear reactor corium samples.
Thermochemistry of UO2 – ThO2 and UO2 – ZrO2 fluorite solid solutions
The Journal of Chemical Thermodynamics, 2017
The enthalpies of formation of cubic urania – thoria (c-Th x U 1-x O 2+y) and urania – zirconia (c-Zr x U 1-x O 2 , x < 0.3) solid solutions at 25 °C from end-member binary oxides (c-UO 2 , and c-ThO 2 or m-ZrO 2) have been measured by high temperature oxide melt solution calorimetry. The enthalpies of mixing for both systems are zero within experimental error. The interaction parameters for binary solid solutions MO 2 – M 0 O 2 (M, M 0 = U, Th, Ce, Zr, and Hf), fitted by regular and subregular thermodynamic models using both calorimetric and computational data, increase linearly with the corresponding volume mismatch. Cubic UO 2 – ZrO 2 appears to be an exception to this correlation and shows a zero heat of mixing despite large size mismatch, suggestive of some short-range ordering and/or incipient phase separation to mitigate the strain. The incorporation of ZrO 2 into UO 2 stabilizes the system and makes it a potential candidate for immobilization and disposal of nuclear waste.
Effect of oxygen-to-metal ratio on properties of corium prepared from UO2 and zircaloy-2
Journal of Nuclear Materials, 2013
The UO 2 and zircaloy system has been studied in severe accident research because these materials form most of a molten corium. In addition, during core meltdown, it is expected that water vapor and hydrogen which is produced in the H 2 O-zircaloy reaction fill the reactor core space. In this study, simulated corium specimens were prepared considering the ambient atmosphere in an accident, and adjustment of the oxygen-to-metal (O/M) ratio. Phase observations by XRD and EPMA and measure ment of thermal conductivity were then done on the specimens. It was confirmed that the prepared O/M ratio had hardly any effe ct on melting temperature although UO 2+x has obvious dep endency on the O/M ratio. Thermal conductivity of the molten specimens showed little dependency on the O/M ratio and temperature. It appeared that the thermal conductivity had already been significantly decreased by the solution of U and Zr. Microstr ucture information as lattice constant and phase segregation was obtained by XRD and EPMA observation. These properties will be basic data for the development of condit ioning techniques for an actual corium.
Corium lavas: structure and properties of molten UO2-ZrO2 under meltdown conditions
Scientific Reports, 2018
In the exceedingly rare event of nuclear reactor core meltdown, uranium dioxide fuel reacts with Zircaloy cladding to produce eutectic melts which can subsequently be oxidized by coolant/moderator water. Oxidized corium liquids in the xUO2·(100 − x)ZrO2 system were produced via laser melting of UO2-ZrO2 mixtures to temperatures in excess of 3000 K. Contamination was avoided by floating the droplets on a gas stream within an aerodynamic levitator and in-situ high-energy x-ray diffraction experiments allowed structural details to be elucidated. Molecular dynamics simulations well reproduced diffraction and density data, and show less compositional variation in thermal expansion and viscosity than suggested by existing measurements. As such, corium liquids maintain their highly penetrating nature irrespective of the amount of oxidized cladding dissolved in the molten fuel. Metal-oxygen coordination numbers vary with both composition and temperature. The former is due to mismatch in nat...
Study of the U-Am-O ternary phase diagram
2017
Americium isotopes are the main contributors to the long-term radiotoxicity of the nuclear wastes, after the plutonium extraction. Among the reprocessing scenarios, the transmutation in fast neutron reactors using uranium-americium mixed oxide (U,Am)O2±x pellets seems promising. In this frame, the knowledge of the thermodynamics of the U-Am-O ternary system is of essential for the prediction of the behavior of (U,Am)O2 pellets and their possible interaction with the cladding, under normal and accidental conditions. This thesis is dedicated to the experimental investigation of U-Am mixed oxides on a wide range of Am contents (7.5 at.% ≤ Am/(Am+U) ≤ 70 at.%), with the aim to collect data for developing a thermodynamic model based on the semi-empirical CALPHAD method. The obtained results can be classified in three categories: structural, phase diagram and thermodynamic data. For the thermodynamic modeling of the ternary system, the assessment of the binary sub-systems is first require...
Calorimetric measurements on plutonium rich (U,Pu)O2 solid solutions
Thermochimica Acta, 2008
Enthalpy increments of U (1−y) Pu y O 2 solid solutions with y = 0.45, 0.55 and 0.65 were measured using a high-temperature differential calorimeter by employing the method of inverse drop calorimetry in the temperature range 956-1803 K. From the fit equations for the enthalpy increments, other thermodynamic functions such as heat capacity, entropy and Gibbs energy function have been computed in the temperature range 298-1800 K. The results are presented and compared with the data available in the literature. The results indicate that the enthalpies of U (1−y) Pu y O 2 solid solutions with y = 0.45, 0.55 and 0.65 obey the Neumann-Kopp's molar additivity rule.
Journal of Nuclear Materials, 2014
Diffusion mechanisms occurring during the sintering of oxide ceramics are affected by the oxygen content of the atmosphere, as it imposes the nature and the concentration of structural defects in the material. Thus, the oxygen partial pressure, p(O 2), of the sintering gas has to be precisely controlled, otherwise a large dispersion in various parameters, critical for the manufacturing of ceramics such as nuclear oxides fuels, is likely to occur. In the present work, the densification behaviour and the solid solution formation of a mixed uraniumplutonium oxide (MOX) were investigated. The initial mixture, composed of 70% UO 2 + 30% PuO 2 , was studied at p(O 2) ranging from 10 À15 to 10 À4 atm up to 1873 K both with dilatometry and in situ high temperature X-ray diffraction. This study has shown that the initial oxides UO 2+x and PuO 2Àx first densify during heating and then the solid solution formation starts at about 200 K higher. The densification and the formation of the solid solution both occur at a lower temperature when p(O 2) increases. Based on this result, it is possible to better define the sintering atmosphere, eventually leading to optimized parameters such as density, oxygen stoichiometry and cations homogenization of nuclear ceramics and of a wide range of industrial ceramic materials.
Melting behavior of (Th,U)O2 and (Th,Pu)O2 mixed oxides
Journal of Nuclear Materials, 2016
The melting behaviors of pure ThO2, UO2 and PuO2 as well as (Th,U)O2 and (Th,Pu)O2 mixed oxides (MOX) have been studied using molecular dynamics (MD) simulations. The MD calculated melting temperatures (MT) of ThO2, UO2 and PuO2 using two-phase simulations, lie between 3650-3675 K, 3050-3075 K and 2800-2825 K, respectively, which match well with experiments. Variation of enthalpy increments and density with temperature, for solid and liquid phases of ThO2, PuO2 as well as the ThO2 rich part of (Th,U)O2 and (Th,Pu)O2 MOX are also reported. The MD calculated MT of (Th,U)O2 and (Th,Pu)O2 MOX show good agreement with the ideal solidus line in the high thoria section of the phase diagram, and evidence for a minima is identified around 5 atom% of ThO2 in the phase diagram of (Th,Pu)O2 MOX.
Ambient melting behavior of stoichiometric uranium oxides
Frontiers in Nuclear Engineering, 2024
As UO 2 is easily oxidized during the nuclear fuel cycle it is important to have a detailed understanding of the structures and properties of the oxidation products. Experimental work over the years has revealed many stable uranium oxides including UO 2 , U 4 O 9 (UO 2.25), U 3 O 7 (UO 2.33), U 2 O 5 (UO 2.5), U 3 O 8 (UO 2.67), and UO 3 , all with a number of different polymorphs. These oxides are broadly split into two categories, fluorite-based structures with stoichiometries in the range of UO 2 to UO 2.5 and less dense layered-type structures with stoichiometries in the range of UO 2.5 to UO 3. While UO 2 is well characterized, both experimentally and computationally, there is a paucity of data concerning higher stoichiometry oxides in the literature. In this work we determine the ambient melting points of all the six stoichiometric uranium oxides listed above and compare them to the available experimental and/or theoretical data. We demonstrate that a family of the six ambient melting points map out a solid-liquid transition boundary consistent with the high-temperature portion of the phase diagram of uranium-oxygen system suggested by Babelot et al.