Phase diagram and thermodynamic properties of the system MnCl2-UCl4, prediction and investigation results

The MnCl₂-UCl₄ system was studied employing differential thermal analysis, common thermal analysis and high-temperature cryometry. From the results obtained, both the phase diagram and the thermodynamic characteristics of liquid phases of the system have been determined. The results were discussed in terms of regularities found in the phase equilibria of binary salt systems with common anion and compared with those for other systems of the type MCl_n-UCl₄. It appeared that the results confirmed earlier predictions on the title system.     

Thermodynamics of the O-U system. I - Oxygen chemical potential critical assessment in the UO2-U3O8 composition range

A critical assessment of oxygen chemical potential of UO_2+x, U₄O₉ and U₃O₈ oxide non-stoichiometric phases as well as of diphasic related domains has been performed in order to build up primary input data files used in a further optimization procedure of thermodynamic and phase diagram data for the uranium-oxygen system in the UO₂-UO₃ composition range. Owing to the fact that original data are very numerous, more than 500 publications, a twofold process is used for the assessment - (i) first a critical selection of data is performed for each method of measurement together with a careful estimate of their uncertainties, (ii) second a reduction of the total number of data on the basis of a chart with fixed intervals of temperature and composition that allows a comparison to be made of the results from the various experiments. Results are presented for chemical potentials of oxygen with their associated uncertainties.     

Phase relations in the quaternary Fe-Pu-U-Zr system

The phase diagram in the quaternary Fe-Pu-U-Zr system was established at 923 K in the uranium-rich region to understand better the compatibility between the metal fuel and stainless steel cladding in a fast reactor. The experimental phase relation data obtained in this study was applied to the thermodynamic methodology for construction of the phase diagram. The calculated phase diagram was consistent and well within the experimental data. The applicability of the thermodynamic model to other temperatures was confirmed by comparing the present results of differential thermal analysis with the calculated phase diagrams. These consistencies mean that both the thermodynamic model and the assessed parameters in the binary and ternary subsystems developed so far are reasonable. The calculated phase diagram established in this study was also in good agreement with the analysis of the diffusion zone in tests on Pu-U-Zr/Fe couples. This suggests that the diffusion zone formed at the fuel-cladding interface in the reactor system can be assessed using the phase diagram in the quaternary Fe-Pu-U-Zr system.     

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.     

Chemical thermodynamic representation of AmO2−x

The AmO_2−x solid solution data set for the dependence of the oxygen potential on the composition, x, and temperature was retrieved from the literature and represented by a thermodynamic model. The data set was analysed by least-squares using equations derived from the classical thermodynamic theory for the solid solution of a solute in a solvent. Two representations of the AmO_2−x data were used, namely the Am_5/4O₂–AmO₂ and AmO_3/2–AmO₂ solid solution. No significant difference was found between the two, and the Am_5/4O₂–AmO₂ solution was preferred on the basis of the phase diagram. From the results the Gibbs energy of formation of Am_5/4O₂ has been derived.     

Enthalpies of formation of U-, Th-, Ce-brannerite: implications for plutonium immobilization

Brannerite, ideally MTi₂O₆, (M=actinides, lanthanides and Ca) occurs in titanate-based ceramics proposed for the immobilization of plutonium. Standard enthalpies of formation, ΔH⁰_f at 298 K, for three brannerite compositions (kJ/mol): CeTi₂O₆ (−2948.8 ± 4.3), U_0.97Ti_2.03O₆ (−2977.9 ± 3.5) and ThTi₂O₆ (−3096.5 ± 4.3) were determined by high temperature oxide melt drop solution calorimetry at 975 K using 3Na₂O · 4MoO₃ solvent. The enthalpies of formation were also calculated from an oxide phase assemblage (ΔH⁰_f-ox at 298 K): MO₂ + 2TiO₂=MTi₂O₆. Only UTi₂O₆ is energetically stable with respect to an oxide assemblage: U_0.97Ti_2.03O₆ (ΔH⁰_f-ox=−7.7±2.8 kJ/mol). Both CeTi₂O₆ and ThTi₂O₆ are higher in enthalpy with respect to their oxide assemblages with (ΔH⁰_f-ox=+29.4±3.6 kJ/mol) and (ΔH⁰_f-ox=+19.4±1.6 kJ/mol) respectively. Thus, Ce- and Th-brannerite are entropy stabilized and are thermodynamically stable only at high temperature.     

Thermodynamic modelling of LiF-LnF3 and LiF-AnF3 phase diagrams

The phase diagrams of the LiF-LnF₃ series, where Ln = La-Sm, and of LiF-AnF₃, where An = U, Pu, have been optimized using Redlich-Kister functions. The phase diagrams of LiF-AmF₃ and LiF-PuF₃-AmF₃ have been calculated. The necessary Gibbs energy functions for americium trifluoride were defined by use of a semi-empirical method. The excess Gibbs energy terms, which are expressed as Redlich-Kister polynomials and describe the effect of interaction between the two fluoride components in the liquid phase, were obtained by translating the trends observed in the lanthanide trifluoride series into the actinide series. A single eutectic has been found in the LiF-AmF₃ system with the eutectic point at ?33 mole% AmF₃ and at ?951 K.     

Thermal variation of the optical absorption of UO2: determination of the small polaron self-energy

The temperature variation of UV-VIS-NIR optical spectra of UO₂ have been investigated from room temperature up to 1173 K with careful in situ oxygen partial pressure control. The deduced optical absorption edge exhibits a strong temperature dependence. Its value decreases from ∼2 eV at room temperature to ∼0.8 eV at 1173 K. Such thermal behaviour is interpreted as the consequence of the existence of a strong electron-phonon coupling (small polaron). In the temperature range 300-1173 K, the model yields a hopping radius of ∼2 Å and a polaron self-energy of E_p=−0.38 eV.     

On the solubility of chromium sesquioxide in uranium dioxide fuel

Chromium sesquioxide doped uranium dioxide systems with a variety of initially added Cr₂O₃ amounts were prepared under different sintering conditions. The solubility limit of Cr for each sintering condition is derived from the measured content of the dopant dissolved in the UO₂ matrix using electron probe microanalysis (EPMA). It is found that the solubility of chromium in a uranium dioxide system for these series sintered at 1600, 1660 and 1760 °C is limited to respectively 0.065±0.002, 0.086±0.003 and 0.102±0.004 wt% Cr. The lattice parameters of the different Cr₂O₃ doped fuels have been examined with X-ray diffraction (XRD). The XRD peaks of the samples sintered at 1760 °C as well as a UO₂ reference without Cr, prepared under the same conditions, were measured and a value for the lattice parameter a for each sample was obtained using the unit cell refinement method. A slight contraction of the lattice parameter is observed with increasing dopant content.     

Effects of natural aging on the tensile properties of water-quenched U-6%Nb alloy

Uranium-6 wt% niobium (U-6%Nb) alloy has been in use for many years in the water-quenched (WQ) condition. The purpose of this work was to determine the effect of natural aging on tensile properties of the WQ U-6%Nb alloy. The materials studied were hemispherical shells after 15 and 20 years in storage. The alloy was successfully tested in the original curved configuration, using the specially designed tensile test apparatus. Finite element analysis confirmed the validity of the test method. The results of the tensile tests clearly indicated that in the WQ condition, the material is changing and after 15 and 20 years, the yield strength exceeds the original maximum allowable specification. The fracture mode transitions from highly ductile, microvoid coalescence in new material to a mixed mode of shallow dimples and inclusion-induced voids in the naturally aged material.     

Oxygen potential and thermal conductivity of (U, Pu) mixed oxides

(U, Pu) mixed oxides, (U_1−yPu_y)O_2−x, with y = 0.21 and 0.28 are being considered as fuels for the Prototype Fast Breeder Reactor (PFBR) in India. The use of urania-plutonia solid solutions in PFBR calls for accurate measurement of physicochemical properties of these materials. Hence, in the present study, oxygen potentials of (U_1−yPu_y)O_2−x, with y = 0.21 and 0.28 were measured over the temperature range 1073-1473 K covering an oxygen potential range of −550 to −300 kJ mol⁻¹ (O/M ratio from 1.96 to 2.000) by employing a H₂/H₂O gas equilibration technique followed by solid electrolyte EMFmeasurement. (U_1−yPu_y)O_2−x, with y = 0.40 is being used in the Fast Breeder Test Reactor (FBTR) in India to test the behaviour of fuels with high plutonium content. However, data on the oxygen potential as well as thermal conductivity of the mixed oxides with high plutonium content are scanty. Hence, the thermal diffusivity of (U_1−yPu_y)O₂, with y = 0.21, 0.28 and 0.40 was measured and the results of the measurements are reported.     

Kinetic and thermodynamic studies of the dissolution of thoria-urania solid solutions

The dissolution of Th_1−xU_xO₂ was investigated through leaching experiments combined with X-ray photoelectron spectroscopy (XPS) and X-ray absorption spectroscopy (XAS) analyses. These experiments were performed in acidic and in oxidizing conditions (nitric solutions), for several compositions of solid solutions ranging from x = 0.24 to 0.81. Static sequential experiments in acidic media performed at room temperature confirmed that higher concentration of uranium in the solid solution leads to higher release of uranium in the leachate whatever the pH. The normalized dissolution rate in oxidizing media is increasing all the more the content of uranium is increases in the mixed oxide. While for Th enriched solids, kinetic parameters remain similar to that of ThO₂, in the case of uranium enriched solids, a drastic change is observed, and kinetic parameters are similar to that of UO₂ ones. For x > 0.50, the saturation is reached in the leachate after 100 days. XPS and EXAFS analysis on leached samples pointed out an oxidation of U(IV) at the surface for x < 0.5, and in the bulk for x > 0.5. Enrichment in Th is also observed at the surface of the solid, indicating the formation of a protective layer of hydrated thorium oxide, or hydroxide. Finally, the solubility product of secondary phase was determined. The values obtained are in good agreement with that of ThO₂, Th(OH)₄ and ThO₂, xH₂O reported in the literature.     

A critical examination of the thermodynamics of water adsorption on actinide oxide surfaces

The reversible adsorption of water from actinide oxide surfaces is examined from several viewpoints in this article. A reinterpretation and critical look at the previously published thermodynamic values for desorption of water from PuO₂ [J. Phys. Chem. 77 (1973) 581] are reexamined in light of more recent mathematical treatments of thermal desorption data from high surface area materials. In addition, the time and temperature dependent process of water adsorption/desorption in closed system experiments is examined using chemical kinetics modeling. A simple experimental method and mathematical treatment of determining adsorption enthalpies based upon a closed system is also described. The desorption enthalpy for reversibly adsorbed water from PuO₂ is determined to be a function of adsorbate coverage with values ranging from 51 to 44 kJ mol⁻¹ for coverages of one to several monolayers (MLs). Consistent desorption enthalpy values are obtained using either approach thus highlighting the importance of proper interpretation of adsorption parameters determined from high surface area powders. Reversible adsorption/desorption equilibrium of water with actinide oxide materials is discussed from the practical standpoint of storage and subsequent pressurization of containers. These results obtained from PuO₂ surfaces are consistent with desorption enthalpies of water from a low surface area UO₂ that has been measured using ultra-high vacuum thermal desorption mass spectroscopy to be 42.2 kJ mol⁻¹.     

Calculation of the threshold displacement energies in UO2 using ionic potentials

A set of ionic potentials matching exactly the crystallographic, elastic and dielectric properties of the uranium dioxide is established. It is further validated upon some basic thermodynamic properties as well as upon the Frenkel pairs formation energies and the activation energies for lattice migration in UO₂. The threshold displacement energies, useful to characterise the radiation resistance of the materials, are calculated for the uranium dioxide along various crystallographic directions applying the optimised force field within the sudden approximation approach.     

Dissolution and precipitation behavior of hydrides in Zircaloy-2 and high Fe Zircaloy

In order to elucidate the terminal solid solubility during the dissolution of hydrides at heatup (TSSD) and during the precipitation of hydrides at cooldown (TSSP) for hydrogenated Zircaloy-2 and high Fe Zircaloy, differential scanning calorimetry (DSC) measurements have been carried out in the temperature range of 50-600 °C. The hydrogen concentrations in the two kinds of alloys ranged from 40 to 542 ppm. There was no difference in either TSSD or TSSP solvi between Zircaloy-2 and high Fe Zircaloy, and best-fit equations were derived for the two curves. In the present TSSP data, two different activation energies, separating into high and low temperature ranges at 260 °C, were obtained. Based on the widths of the DSC peak obtained during cooldown, the average precipitation (nucleation plus growth) rates of zirconium hydrides from super-saturated state were assessed. The activation energy of the precipitation rate was approximately equivalent to reported values of hydrogen diffusion coefficients of Zr and Zircaloys.     

Critical temperature of the lead-bismuth eutectic (LBE) alloy

Liquid metals such as Bi and Pb and Pb-Bi eutectic alloy are serious contenders for use as coolant in LMFBRs in lieu of sodium due to a number of attractive characteristics (high density, low moderation, low neutron absorption and activation, high boiling point and poor interaction with water and air, etc.). Analysis of hypothetical accidents is of relevance to predict the catastrophe involving loss of coolant accident (LOCA) in LMFBRs. One key parameter to take into account is the critical temperature data of the liquid metals for reactor safety analysis. This communication reports the application of a theoretical model called internal pressure approach to predict the critical temperature (T_c) of the LBE alloy for the first time.     

Effect of burn-up on the thermal conductivity of uranium dioxide up to 100.000 MWd t

The thermal diffusivity and specific heat of reactor-irradiated UO₂ fuel have been measured. Starting from end-of-life conditions at various burn-ups, measurements under thermal annealing cycles were performed in order to investigate the recovery of the thermal conductivity as a function of temperature. The separate effects of soluble fission products, of fission gas frozen in dynamical solution and of radiation damage were determined. In this context, particular emphasis was given to the behaviour of samples displaying the high burn-up rim structure. Recovery stages could be thoroughly investigated in samples that were irradiated at low burn-ups and/or at high irradiation temperatures. Other samples, in particular those exhibiting the characteristic rim structure, disintegrated at temperatures slightly higher than the irradiation temperature. Finally, from a database of several thousand measurements, an accurate formula for the in-pile thermal conductivity of UO₂ up to 100 GWd t⁻¹ was developed, taking into account all the relevant effects and structural changes induced by reactor burn-up.     

Examination of U valence states in the brannerite structure by near-infrared diffuse reflectance and X-ray photoelectron spectroscopies

The valence state of uranium doped into a f⁰ thorium analog of brannerite (i.e., thorutite) has been examined using near-infrared (NIR) diffuse reflectance (DRS) and X-ray photoelectron (XPS) spectroscopies. NIR transitions of U⁴⁺, which are not observed in spectra of brannerite, have been detected in the samples of U_xTh_1−xTi₂O₆, and we propose that strong specular reflectance is responsible for the lack of U⁴⁺ features in UTi₂O₆. Characteristic U⁵⁺ bands have been identified in samples in which sufficient Ca²⁺ has been added to nominally effect complete oxidation to U⁵⁺. XPS results support the assignments of U⁴⁺ and U⁵⁺ by DRS. The presence of residual U⁴⁺ bands in the spectra of the Ca-doped samples is consistent with segregation of Ca²⁺ to the grain boundaries during high temperature sintering.