Study of the U-Am-O ternary phase diagram (original) (raw)
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Melting behaviour of uranium-americium mixed oxides under different atmospheres
The Journal of Chemical Thermodynamics, 2019
In the context of a comprehensive campaign for the characterisation of transmutation fuels for next generation nuclear reactors, the melting behaviour of mixed uranium-americium dioxides has been experimentally studied for the first time by laser heating, for Am concentrations up to 70 mol. % under different types of atmospheres. Extensive post-melting material characterisations were then performed by X-ray absorption spectroscopy and electron microscopy. The melting temperatures observed for the various compositions follow a markedly different trend depending on the experimental atmosphere. Uraniumrich samples melt at temperatures significantly lower (around 2700 K) when they are laser-heated in a strongly oxidizing atmosphere compressed air at (0.300 ± 0.005) MPa, compared to the melting points (beyond 3000 K) registered for the same compositions in an inert environment (pressurised Ar). This behaviour has been interpreted on the basis of the strong oxidation of such samples in air, leading to lower-melting temperatures. Thus, the melting temperature trend observed in air is characterized, in the purely pseudo-binary dioxide plane, by an apparent maximum melting temperature around 2850 K for 0.3 < x(AmO 2) < 0.5. The melting points measured under inert atmosphere uniformly decrease with increasing americium content, displaying an approximately ideal solution behaviour if a melting point around 2386 K is assumed for pure AmO 2. In reality, it will be shown that the (U, Am)-oxide system can only be rigorously described in the ternary U-Am-O phase diagram, rather than the UO 2-AmO 2 pseudo-binary, due to the aforementioned over-oxidation effect in air. Indeed, general departures from the oxygen stoichiometry (Oxygen/Metal ratios-2.0) have been highlighted by the X-ray Absorption Spectroscopy (XAS). Finally, to help interpret the experimental results, thermodynamic computations based on the CALPHAD method will be presented.
Inorganic Chemistry, 2017
In the frame of minor actinide transmutation, americium can be diluted in UO 2 and (U, Pu)O 2 fuels burned in fast neutron reactors. The first mandatory step to foresee the influence of Am on the in-reactor behavior of transmutation targets or fuel is to have fundamental knowledge of the Am−O binary system and, in particular, of the AmO 2−x phase. In this study, we coupled HT-XRD (high-temperature X-ray diffraction) experiments with CALPHAD thermodynamic modeling to provide new insights into the structural properties and phase equilibria in the AmO 2−x −AmO 1.61+x − Am 2 O 3 domain. Because of this approach, we were able for the first time to assess the relationships between temperature, lattice parameter, and hypostoichiometry for fcc AmO 2−x. We showed the presence of a hyperstoichiometric existence domain for the bcc AmO 1.61+x phase and the absence of a miscibility gap in the fcc AmO 2−x phase, contrary to previous representations of the phase diagram. Finally, with the new experimental data, a new CALPHAD thermodynamic model of the Am−O system was developed, and an improved version of the phase diagram is presented.
Environmental Research, Engineering and Management, 2017
The reprocessing of actinides with long half-life has been non-existent except for plutonium (Pu). This work looks at reducing the actinides inventory nuclear fuel waste meant for permanent disposal. The uranium oxide fuel (UOX) assembly, as in the open cycle system, was designed to reach a burnup of 46GWd/T and 68GWd/T using the MURE code. The MURE code is based on the coupling of a static Monte Carlo code and the calculation of the evolution of the fuel during irradiation and cooling periods. The MURE code has been used to address two different questions concerning the mono-recycling of americium (Am) in present French pressurised water reactors (PWR). These are reduction of americium in the clear fuel cycle and the safe quantity of americium that can be introduced into mixed oxide (MOX) as fuel. The spent UOX was reprocessed to fabricate MOX assemblies, by the extraction of plutonium and addition of depleted uranium to reach burnups of 46GWd/T and 68GWd/T, taking into account various cooling times of the spent UOX assembly in the repository. The effect of cooling time on burnup and radiotoxicity was then ascertained. After 30 years of cooling in the repository, the spent UOX fuel required a higher concentration of Pu to be reprocessed into MOX fuel due to the decay of Pu-241. Americium, with a mean half-life of 432 years, has a high radiotoxicity level, high mid-term residual heat and is a precursor for other long-lived isotopes. An innovative strategy would be to reprocess not only the plutonium from the UOX spent fuel but also the americium isotopes, which presently dominate the radiotoxicity of waste. The mono-recycling of Am is not a definitive solution because the once-through MOX cycle transmutation of Am in a PWR is not enough to destroy all americium. The main objective is to propose a 'waiting strategy' for both Am and Pu in the spent fuel so that they can be made available for further transmutation strategies. The MOX and americium isotopes (MOXAm) fuel was fabricated to see the effect of americium in MOX fuel on the burnup, neutronic behaviour and radiotoxicity. The MOXAm fuel showed relatively good indicators on both burnup and radiotoxicity. A 68GWd/T MOX assembly produced from a reprocessed fuel spent 46GWd/T UOX assembly showed a decrease in radiotoxicity as compared with the open cycle. All fuel types understudied in the PWR cycle showed a good safety inherent feature with the exception of some MOXAm assemblies that have a positive void coefficient in specific configurations, which would not be consistent with safety features.
A thermodynamic study of the Pu–Am–O system
Journal of Nuclear Materials, 2011
Vapour pressure measurements were performed on a (Pu 0.756 Am 0.244)O 2Àx sample using Knudsen cell mass spectrometry. The total and partial vapour pressures of the gaseous species have been measured in the temperature range from 2000 to 2300 K. The evolution of the plutonium and americium bearing species was also determined as a function of time, in order to evaluate the congruent vapour composition. At constant temperature, the energy of ionising electrons was stepwise increased and the ionisation efficiency curves were recorded. The results were combined with the assessment of the Pu-Am-O system using the CALPHAD method. To obtain the model of this ternary system, the data on the Pu-Am and Am-O binaries have been evaluated and the optimised phase diagrams are presented. A consistent thermodynamic description of the ternary was obtained which allows the calculation of the ternary phase diagram, the oxygen potential for (Pu, Am)O 2±x and the equilibrium partial vapour pressures.
An evaluation of the Americium insertion in Uo2 fuel
Annals of Nuclear Energy, 2002
The idea is to evaluate the neutronic performance of a core of UO 2 fuel with the Americium insertion in a standard PWR core. The homogeneous mode is used and changes in the moderator to fuel volume ratio (V m /V f) are evaluated. For the simulation the WIMS-D5 code was utilised.
MARIOS: Irradiation of UO 2 containing 15% americium at well defined temperature
Nuclear Engineering and Design
Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors of long-lived nuclides like 241 Am is, therefore, an option for the reduction of radiotoxicity and residual power packages as well as the repository area. The MARIOS irradiation experiment is the latest of a series of experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS). MARIOS experiment is carried out in the framework of the 4-year project FAIRFUELS of the EURATOM 7th Framework Programme (FP7). During the past years of experimental work in the field of transmutation and tests of innovative nuclear fuel containing americium, the release or trapping of helium as well as swelling has shown to be the key issue for the design of such kinds of target. Therefore, the main objective of the MARIOS experiment is to study the in-pile behaviour of UO 2 containing minor actinides (MAs) in order to gain knowledge on the role of the microstructure and of the temperature on the gas release and on fuel swelling.
A neutronic evaluation of the Americium and Neptunium co-insertion in UO2 fuel
Annals of Nuclear Energy, 2003
In this work the neutronic performance of a core of UO 2 fuel with the Americium and the Neptunium co-insertion in a standard PWR (pressurized water reactor) core is evaluated. The homogeneous mode is used and changes in the moderator to fuel volume ratio (V m /V f) are evaluated. For the simulation the WIMS-D5 code was utilised.
Phase characteristics of a number of U–Pu–Am–Np–Zr metallic alloys for use as fast reactor fuels
Journal of Nuclear Materials, 2010
Metallic fuel alloys consisting of uranium, plutonium, and zirconium with minor additions of americium and neptunium are under evaluation for potential use to transmute long-lived transuranic actinide isotopes in fast reactors. A series of test designs for the Advanced Fuel Cycle Initiative (AFCI) have been irradiated in the Advanced Test Reactor (ATR), designated as the AFC-1 and AFC-2 designs. Metal fuel compositions in these designs have included varying amounts of U, Pu, Zr, and minor actinides (Am, Np). Investigations into the phase behavior and relationships based on the alloy constituents have been conducted using X-ray diffraction and differential thermal analysis. Results of these investigations, along with proposed relationships between observed behavior and alloy composition, are provided. In general, observed behaviors can be predicted by a ternary U-Pu-Zr phase diagram, with transition temperatures being most dependent on U content. Furthermore, the enthalpy associated with transitions is strongly dependent on the as-cast microstructural characteristics.
Alpha self-irradiation effect on the local structure of the U0.85Am0.15O2±x solid solution
Journal of Solid State Chemistry, 2012
Uranium-americium mixed oxides are promising fuels for achieving an efficient Am recycling. Previous studies on U 0.85 Am 0.15 O 2 7 x materials showed that the high a activity of 241 Am induces pellet swelling which is a major issue for cladding materials design. In this context, X-ray Diffraction and X-ray Absorption Spectroscopy measurements were used to study self-irradiation effects on U 0.85 Am 0.15 O 2 7 x local structure and to correlate these results with those obtained at the macroscopic scale. For a cumulative a decay dose equal to 0.28 dpa, it was shown that non-defective fluorite solid solutions were achieved and therefore, that the fluorite structure is stable for the studied doses. In addition, both interatomic distance and lattice parameter expansions were observed, which only partially explains the macroscopic swelling. As expected, an increase of the structural disorder with self-irradiation was also observed.