Melting behavior of MgO-based inert matrix fuels containing (Pu,Am)O2−x

The melting behavior of MgO-based inert matrix fuels containing (Pu,Am)O_2−x ((Pu,Am)O_2−x-MgO fuels) was experimentally investigated. Heat-treatment tests were carried out at 2173 K, 2373 K and 2573 K each. The fuel melted at about 2573 K in the eutectic reaction of the Pu-Am-Mg-O system. The (Pu,Am)O_2−x grains, MgO grains and pores grew with increasing temperature. In addition, Am-rich oxide phases were formed in the (Pu,Am)O_2−x phase by heat-treatment at high temperatures. The melting behavior was compared with behaviors of PuO_2−x-MgO and AmO_2−x-MgO fuels.     

Thermal conductivities of hypostoichiometric (U, Pu, Am)O2−x oxide

The thermal conductivities of (U_0.68Pu_0.30Am_0.02)O_2.00−x solid solutions (x = 0.00-0.08) were studied at temperatures from 900 to 1773 K. The thermal conductivities were obtained from the thermal diffusivities measured by the laser flash method. The thermal conductivities obtained experimentally up to about 1400 K could be expressed by a classical phonon transport model, λ = (A + BT)⁻¹, A(x) = 3.31 × x + 9.92 × 10⁻³ (mK/W) and B(x) = (−6.68 × x + 2.46) × 10⁻⁴ (m/W). The experimental A values showed a good agreement with theoretical predictions, but the experimental B values showed not so good agreement with the theoretical ones in the low O/M ratio region. From the comparison of A and B values obtained in this study with the ones of (U,Pu)O_2−x obtained by Duriez et al. [C. Duriez, J.P. Alessandri, T. Gervais, Y. Philipponneau, J. Nucl. Mater. 277 (2000) 143], the addition of Am into (U, Pu)O_2−x gave no significant effect on the O/M dependency of A and B values.     

Effect of oxygen potential on the sintering behavior of MgO-based heterogeneous fuels containing (Pu, Am)O2−x

The effect of oxygen potential on the sintering behavior of MgO-based heterogeneous fuels containing (Pu, Am)O_2−x was experimentally investigated. Sintering tests in various atmospheres, i.e. air, moisturized 4%H₂-Ar, and 4%H₂-Ar atmosphere, were carried out. The sintering behavior was found to be significantly affected by the oxygen potential in the sintering atmosphere. The sintered density decreased with decreasing oxygen potential. The (Pu, Am)O_2−x phase sintered in a reductive atmosphere had hypostoichiometry. The aggregates of the (Pu, Am)O_2−x phase sintered in the reductive atmosphere grew, in comparison with those in the oxidizing one. The sintering mechanism was discussed in terms of the difference in sintering behavior of (Pu, Am)O_2−x and MgO.     

Thermodynamic properties of ThxU1−xO2 (0 < x < 1) based on quantum-mechanical calculations and Monte-Carlo simulations

Th_xU_1−xO_2+y binary compositions occur in nature, uranothorianite, and as a mixed oxide nuclear fuel. As a nuclear fuel, important properties, such as the melting point, thermal conductivity, and the thermal expansion coefficient change as a function of composition. Additionally, for direct disposal of Th_xU_1−xO₂, the chemical durability changes as a function of composition, with the dissolution rate decreasing with increasing thoria content. UO₂ and ThO₂ have the same isometric structure, and the ionic radii of 8-fold coordinated U⁴⁺ and Th⁴⁺ are similar (1.14 nm and 1.19 nm, respectively). Thus, this binary is expected to form a complete solid solution. However, atomic-scale measurements or simulations of cation ordering and the associated thermodynamic properties of the Th_xU_1−xO₂ system have yet to be determined. A combination of density-functional theory, Monte-Carlo methods, and thermodynamic integration are used to calculate thermodynamic properties of the Th_xU_1−xO₂ binary (ΔH_mix, ΔG_mix, ΔS_mix, phase diagram). The Gibbs free energy of mixing (ΔG_mix) shows a miscibility gap at equilibration temperatures below 1000 K (e.g., E_exsoln = 0.13 kJ/(mol cations) at 750 K). Such a miscibility gap may indicate possible exsolution (i.e., phase separation upon cooling). A unique approach to evaluate the likelihood and kinetics of forming interfaces between U-rich and Th-rich has been chosen that compares the energy gain of forming separate phases with estimated energy losses of forming necessary interfaces. The result of such an approach is that the thermodynamic gain of phase separation does not overcome the increase in interface energy between exsolution lamellae for thin exsolution lamellae (10 Å). Lamella formation becomes energetically favorable with a reduction of the interface area and, thus, an increase in lamella thickness to >45 Å. However, this increase in lamellae thickness may be diffusion limited. Monte-Carlo simulations converge to an exsolved structure [lamellae || ] only for very low equilibration temperatures (below room temperature). In addition to the weak tendency to exsolve, there is an ordered arrangement of Th and U in the solid solution [alternating U and Th layers || {1 0 0}] that is energetically favored for the homogeneously mixed 50% Th configurations. Still, this tendency to order is so weak that ordering is seldom reached due to kinetic hindrances. The configurational entropy of mixing (ΔS_mix) is approximately equal to the point entropy at all temperatures, indicating that the system is not ordered.     

Oxygen potential of hypo-stoichiometric Lu-doped UO2

Oxygen potentials of hypo-stoichiometric Lu-doped UO₂, (U_0.80Lu_0.20)O_2−x, were experimentally investigated by thermogravimetric analysis using H₂O/H₂ gas equilibria at 1173, 1273 and 1473 K. The oxygen potentials of (U,Lu)O_2−x were higher than those of other forms of rare earth-doped UO₂, specifically (U,Nd)O_2−x, (U,Gd)O_2−x, and (U,Er)O_2−x. Slope analyses for plots of oxygen potential versus deviation from stoichiometry indicated that (U_0.80Lu_0.20)O_2−x had a similar defect structure to that of the other forms of rare earth-doped UO₂. A relationship between the effective ionic radii and oxygen potentials was found for the hypo-stoichiometric rare earth-doped UO₂.     

Oxygen potential measurements of Am0.5Pu0.5O2−x by EMF method

The dependence of the oxygen potentials on oxygen non-stoichiometry and temperature of Am_0.5Pu_0.5O_2−x has been obtained by the electromotive force (EMF) method with the cell: (Pt) air |Zr(Ca)O_2−x| Am_0.5Pu_0.5O_2−x (Pt). The x value of Am_0.5Pu_0.5O_2−x was changed at 1333 K over 0.02 < x ? 0.25 by the coulomb titration method. The temperature dependence of the oxygen potential was also measured over the range of 1173-1333 K. It was found that the oxygen potential decreased from −80 to −360 kJ mol⁻¹ with increasing x from 0.021 to 0.22 at 1333 K and that it remained almost constant at −360 kJmol⁻¹ around x = 0.23. It was concluded that Am_0.5Pu_0.5O_2−x should be composed of the single fluorite-type phase over 0.02 < x ? 0.22 and the mixed phases of fluorite-type and (Am, Pu)₉O₁₆ at around x = 0.23.     

Phasengleichgewichte in den systemen UO2-UO2,67-ThO2 und UO2 + x-NPO2

Solid-state chemical investigations have established that in the compositional range UO₂-UO_2.67-ThO₃ of the U-Th-O ternary system, the following single-phase domains exist: U₃O₈, which does not dissolve any ThO₂ in the solid state; an ordered M₄O₉ phase on the section between U₄O₉ and U₂Th₂O₉, below ≈ 1150 °C; and a phase with fluorite structure which occupies a large part of the system and which at 1250 °C is bounded by the compositions UO₂-UO_2.25 (U_0.43, ThO_0.57)O_0.25-ThO₃. The maximum O/M ratio of the “fluorite” phase is O:(U + Th) = 2.25. The highest oxidation valency of uranium is 5.30; this value falls as more thorium oxide is incorporated in the (U.Th)O_{2 + x} “fluorite” phase.     

UO2/Zry-4 chemical interaction layers for intact and leak PWR fuel rods

In this study, the UO₂ pellet-Zry-4 cladding interfaces of intact and leak PWR fuel rods were examined with the help of an optical microscope and a scanning electron microscope to investigate typical chemical interaction layers formed at the pellet-cladding interface during the normal reactor operations. The two intact and two leak fuel rods with the burnup of between 35,000 and 53,000 MWD/MTU were selected to evaluate the effects of gap-gas compositions and fuel burnup on the chemical interaction layer formation. Based on the optical and scanning electron micrographs, it is found that the intact fuel rod generates apparently one interaction layer of (U,Zr)O_2−x at the interface, whereas the leak fuel rod generates apparently two interaction layers of ZrO_2−x and (U,Zr)O_2−x. These interaction layers for the intact and leak fuel rods were predicted by several diffusion paths drawn on a U-Zr-O ternary phase diagram. The variations of chemical element compositions around the interface of one intact rod were generated by an electron probe micro-analyzer to confirm the interaction layers at the pellet-cladding interface. The interaction layer growth rates of the ZrO_2−x and (U,Zr)O_2−x phases were estimated, using the layer thicknesses and the reaction times.     

Structural studies of the phase separation in the UO2–PuO2–Pu2O3 ternary system

In the oxygen hypo-stoichiometric range of (U_1?yPu_y)O_2?x mixed oxide MOX fuels, the U–Pu–O phase diagram is known to exhibit a large biphasic domain depending on the Pu content. However, the phase equilibria are still to be fully described as various representations are proposed in the literature.In the present work, we notify new insights into the phase separation occurring in the UO₂–PuO₂–Pu₂O₃ domain at room temperature. Our microstructural and X-ray diffraction results are compared to the different representations reported in the literature. We provide, for the first time in the hypo-stoichiometric domain, an indisputable experimental observation of a triphasic region at high Pu content, composed of two fluorite-type structures and of one α-Pu₂O₃ sesquioxyde type structure. These results are in contradiction with previous experimental representations of the U–Pu–O ternary system.     

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.     

Lattice parameters of (U, Pu, Am, Np)O2−x

The solid solutions of (U_{1−z−y’−y”}Pu_zAm_y’Np_y”)O_2−x (z = 0-1, y’ = 0-0.12, y” = 0-0.07) were investigated by X-ray diffraction measurements, and a database for the lattice parameters was updated. A model to calculate the lattice parameters was derived from the database. The radii of the ions present in the fluorite structure of (U, Pu, Am, Np)O_2−x were estimated from the lattice parameters measured in this work. The model represented the experimental data within a standard deviation of σ = ±0.025%.     

Thermal studies on fluorite type ZryU1−yO2 solid solutions

Solid state reactions of UO₂ and ZrO₂ in mild oxidizing condition followed by reduction at 1673 K showed enhanced solubility up to 35 mol% of zirconium in UO₂ forming cubic fluorite type Zr_yU_1−yO₂ solid solution. The lattice parameters and O/M (M = U + Zr) ratios of the solid solutions, Zr_yU_1−yO_2+x, prepared in different gas streams were investigated. The lattice parameters of these solid solutions were expressed as a linear equation of x and y: a₀ (nm) = 0.54704 − 0.021x - 0.030y. The oxidation of these solid solutions for 0.1 ? y ? 0.2 resulted in cubic phase MO_2+x up to700 K and single orthorhombic zirconium substituted α-U₃O₈ phase at 1000 K. The kinetics of oxidation of Zr_yU_1−yO₂ in air for y = 0-0.35 were also studied using thermogravimetry. The specific heat capacities of Zr_yU_1−yO₂ (y = 0-0.35) were measured using heat flux differential scanning calorimetry in the temperature range of 334-860 K.     

Oxygen potential of (U0.88Pu0.12)O2±x and (U0.7Pu0.3)O2±x at high temperatures of 1673-1873 K

The oxygen potential of (U_0.88Pu_0.12)O_2±x (−0.0119 < x < 0.0408) and (U_0.7Pu_0.3)O_2±x (−0.0363 < x < 0.0288) was measured at high temperatures of 1673-1873 K using gas equilibrium method with thermo gravimeter. The measured data were analyzed by a defect chemistry model. Expressions were derived to represent the oxygen potential based on defect chemistry as functions of temperature and oxygen-to-metal ratio. The thermodynamic data, and , at stoichiometric composition were obtained. The expressions can be used for in situ determination of the oxygen-to-metal ratio by the gas-equilibration method. The calculation results were consistent with measured data. It was estimated that addition of 1 wt.% Pu content increased oxygen potential of uranium and plutonium mixed oxide by 2-5 kJ/mol.     

Effect of Nd and Pr addition on the thermal and mechanical properties of (U, Ce)O2

The thermal conductivity, Young’s modulus, and hardness of (U_0.65−xCe_0.3Pr_0.05Nd_x)O₂ (x = 0.01, 0.08, 0.12) were evaluated and the effect of Pr and Nd addition on the properties of (U, Ce)O₂ were studied. The polycrystalline high-density pellets were prepared with solid state reactions of UO₂, CeO₂, Pr₂O₃, and Nd₂O₃. We confirmed that all Ce, Pr, and Nd dissolved in UO₂ and formed solid solutions of (U, Ce, Pr, Nd)O₂. We revealed that the thermal conductivity of (U_0.65−xCe_0.3Pr_0.05Nd_x)O₂ (x = 0.12) was up to 25% lower than that of x = 0.01 at room temperature. The Young’s modulus of (U_0.65−xCe_0.3Pr_0.05Nd_x)O₂ decreased with x, whereas the hardness values were constant in the investigated x range.     

A dilatometric study of the solubility of U4O9 in UO2

Phase relationships in the system UO₂-O₉ were studied using a dilatometei in which the O/U ratio of a UU_2+x specimen could be both controlled and measured. Phase boundary temperatures were indicated by changes in expansion or contraction rate during heating or cooling, respectively. The solubility of U₄O₉ in UO₂ agreed with the results of previous workers using other techniques. The dependence of solubility on temperature is complex, and appears to be influenced by a high-temperature phase transition in U₄0₉.     

Light ion irradiation effects on stuffed Lu2(Ti2−xLux)O7−x/2 (x = 0, 0.4 and 0.67) structures

We have recently synthesized “stuffed” (i.e., excess Lu) Lu₂(Ti_2−xLu_x)O_7−x/2 (x = 0, 0.4 and 0.67) compounds using conventional ceramic processing. X-ray diffraction measurements indicate that stuffing more Lu³⁺ cations into the oxide structure leads eventually to an order-to-disorder (O-D) transition, from an ordered pyrochlore to a disordered fluorite crystal structure. At the maximum deviation in stoichiometry (x = 0.67), the Lu³⁺ and Ti⁴⁺ ions become completely randomized on the cation sublattices, and the oxygen “vacancies” are randomized on the anion sublattice. Samples were irradiated with 400 keV Ne²⁺ ions to fluences ranging from 1 × 10¹⁵ to 1 × 10¹⁶ ions/cm² at cryogenic temperatures (∼77 K). Ion irradiation effects in these samples were examined by using grazing incident X-ray diffraction. The results show that the ion irradiation tolerance increases with disordering extent in the non-stoichiometric Lu₂(Ti_2−xLu_x)O_7−x/2.     

Oxygen potentials of mixed oxide fuels for fast reactors

Oxygen potentials of homogenous (Pu_0.2U_0.8)O_2−x and (Am_0.02Pu_0.30Np_0.02U_0.66)O_2−x which have been developed as fuels for fast breeder reactors were measured at temperatures of 1473-1623 K by a gas equilibrium method using an (Ar, H₂, H₂O) gas mixture. The measured oxygen potentials of (Pu_0.2U_0.8)O_2−x were about 25 kJ mol⁻¹ lower than those of (Pu_0.3U_0.7)O_2−x measured previously and were consistent with the values calculated by Besmann and Lindemer’s model. The measured oxygen potentials of (Am_0.02Pu_0.30Np_0.02U_0.66)O_2−x were slightly higher than those of MOX without minor actinides. No fuel-cladding chemical interaction is affected significantly by adding their minor actinides.     

Equilibrium and nonequilibrium molecular dynamics simulations of heat conduction in uranium oxide and mixed uranium-plutonium oxide

The thermal conductivity of nuclear fuels such as UO_2+x and (U,Pu)O_2−x has been calculated by the molecular dynamics (MD) simulation in terms of oxygen stoichiometric parameter x, temperature and Pu content. In the present study, the MD calculations were carried out in both equilibrium (EMD) and nonequilibrium (NEMD) systems. In the EMD simulation, the thermal conductivity was defined as the time-integral of the correlation function of heat fluxes according to the Green-Kubo relationship. Meanwhile, in the homogeneous NEMD, it was given by the ratio of the time-averaged heat flux to the perturbed external force subjected to each particle in the simulated cell. NEMD, as compared with EMD, gave somewhat precise results efficiently. Furthermore, both MD calculations showed that the thermal conductivity of these oxide fuels decreased with increase of temperature and defects, i.e. excess oxygen or vacancy, and was rather insensitive to Pu content for the stoichiometric fuel.     

Progress in the thermodynamic modelling of the O-U binary system

The fuel of civil nuclear plants, UO₂, melts at 3120 K. During an hypothetical severe accident, urania, submitted to high temperatures and various oxygen potentials, presents a wide non-stoichiometry range: the melting temperature of UO_2±x, related to oxygen potential, decreases in all cases. In this scenario, urania could react with other materials, firstly zircaloy, and the melting temperature of (U, Zr)O_2±x still decreases. That is why the critical assessment of the O-U binary system including the non-stoichiometry range of urania, is a major step to a correct thermodynamic modelling of multicomponent systems for nuclear safety. The very numerous experimental information has been compiled and analysed. The associate model was used for the liquid phase, and a sublattice model for UO_2±x; U₄O_9−y, U₃O₈ and UO₃ were treated as stoichiometric. Phase diagram and thermodynamic properties have been calculated from the optimised Gibbs energy parameters. The calculated consistency with the experimental ones is quite satisfactory.     

Solubility of ThO2 in Gd2Zr2O7 pyrochlore: XRD, SEM and Raman spectroscopic studies

Solubility of ThO₂ in gadolinium zirconate pyrochlore, a potential host for radioactive materials, has been investigated. The phase relations in Gd_2−xTh_xZr₂O_7+x/2 (0.0 ? x ? 2.0) systems have been established under the slow-cooled conditions from 1400 °C. XRD studies reveal that the compositions corresponding to x = 0.0-0.075 are single phasic in nature and beyond x ? 0.1 the biphasic region starts. The first biphasic region comprising of pyrochlore and thoria exist from x = 0.1-0.8, and from x = 1.2 another biphasic region consisting of gadolinia stabilized zirconia (GSZ) and thoria appears which persists till x = 1.6. The end member (i.e. x = 2.0) of the series is found to be a mixture of monoclinic ZrO₂ and thoria. Interestingly, gadolinia which has wide solubility in thoria, did not show any miscibility in thoria in the presence of zirconia. Irregular grains of Gd_1.8Th_0.2Zr₂O_7.1 as shown in SEM supports its biphasic nature. Raman spectra of heavily thoria doped (x = 0.1 and 0.2) samples, indicates the presence of Zr-O₇ mode which implies the samples are highly disordered in nature.