Separation of uranium from (U, Th)O2 and (U, Pu)O2 by solid state reactions route

Solid state reactions of UO₂, ThO₂, PuO₂ and their mixed oxides (U, Th)O₂ and (U, Pu)O₂ were carried out with sodium nitrate upto 900 °C, to study the formation of various phases at different temperatures, which are amenable for easy dissolution and separation of the actinide elements in dilute acid. Products formed by reacting unsintered as well as sintered UO₂ with NaNO₃ above 500 °C were readily soluble in 2 M HNO₃, whereas ThO₂ and PuO₂ did not react with NaNO₃ to form any soluble products. Thus reactions of mixed oxides (U, Th)O₂ and (U, Pu)O₂ with NaNO₃ were carried out to study the quantitative separation of U from (U, Th)O₂ and (U, Pu)O₂. X-ray diffraction, X-ray fluorescence, thermal analysis and chemical analysis techniques were used for the characterization of the products formed during the reactions.     

Thermal conductivities of irradiated UO2 and (U, Gd)O2 pellets

Thermal diffusivities of UO₂ and (U, Gd)O₂ pellets irradiated in a commercial reactor (maximum burnups: 60 GWd/t for UO₂ and 50 GWd/t for (U, Gd)O₂) were measured up to about 2000 K by using a laser flash method. The thermal diffusivities of irradiated UO₂ and (U, Gd)O₂ pellets showed hysteresis phenomena: the thermal diffusivities of irradiated pellets began to recover above 750 K and almost completely recovered after annealing above 1400 K. The thermal diffusivities after recovery were close to those of simulated soluble fission products (FPs)-doped UO₂ and (U, Gd)O₂ pellets, which corresponded with the recovery behaviors of irradiation defects for UO₂ and (U, Gd)O₂ pellets. The thermal conductivities for irradiated UO₂ and (U, Gd)O₂ pellets were evaluated from measured thermal diffusivities, specific heat capacities of unirradiated UO₂ pellets and measured sample densities. The difference in relative thermal conductivities between irradiated UO₂ and (U, Gd)O₂ pellets tended to become insignificant with increasing burnups of samples.     

Thermal and mechanical properties of (U,Er)O2

Erbium is considered as a slow burnable poison suitable for use in light water reactors (LWRs). Addition of a small amount of Er₂O₃ to all UO₂ pellets will make it possible to develop super high burnup fuels in Japanese nuclear facilities which are now under the restriction of the upper limit of ²³⁵U enrichment. When utilizing the (U,Er)O₂ fuels, it is very important to understand the thermal and mechanical properties. Here we show the characterization results of (U_1−xEr_x)O₂ (0 ? x ? 0.1). We measured their thermal and mechanical properties and investigated the effect of Er addition on these properties of (U,Er)O₂. All Er completely dissolved in UO₂, and the lattice parameter decreased linearly with the Er content. Both the thermal conductivity and Young’s modulus of (U,Er)O₂ decreased with the Er content. These results would be useful for us in evaluating the performance of the (U,Er)O₂ fuels in LWRs.     

Thermophysical properties of (U,Y)O2

We prepared polycrystalline pellets of (U,Y)O₂, containing YO_1.5 up to 11 mol.%. We performed indentation tests on the pellets, and evaluated the Young’s modulus and hardness. We measured the heat capacity and the thermal diffusivity, and evaluated the thermal conductivity. We succeeded in evaluating the effect of Y content on the thermophysical properties of (U,Y)O₂. We revealed that the Young’s modulus, hardness, and thermal conductivity of (U,Y)O₂ decreased with increasing the Y content.     

Defect volumes of BO2 doped Y2O3 (B = Ti, Zr, Hf and Ce)

Atomistic simulations have been employed to study the effect of BO₂ (fluorite) incorporation into the bixbyite oxide Y₂O₃. The energetically preferred defect mechanism and the associated lattice parameter changes that occur from BO₂ doping have been predicted. The addition of Group IV elements into Y₂O₃ can follow three different mechanisms. The energetically favourable method is through a mediated reaction for ZrO₂ and HfO₂ while for TiO₂ and CeO₂, reducing B⁴⁺ to B³⁺ provides the lowest energy reaction. ZrO₂ and HfO₂ doping results in the lowest volume changes.     

A XAS study of the local environments of cations in (U, Ce)O2

Mixed oxide (MOX) fuel is usually considered as a solid solution formed by uranium and plutonium dioxides. Nevertheless, some physico-chemical properties of (U_1−y, Pu_y)O₂ samples manufactured under industrial conditions showed anomalies in the domain of plutonium contents ranging between 3 and 15 at.%. Cerium is commonly used as an inactive analogue of plutonium in preliminary studies on MOX fuels. Extended X-ray Absorption Fine Structure (EXAFS) measurements performed at the European Synchrotron Radiation Facility (ESRF) at the cerium and uranium edges on (U_1−y, Ce_y)O₂ samples are presented and discussed. They confirmed on an atomic scale the formation of an ideal solid solution for cerium concentrations ranging between 0 and 50 at.%.     

Properties of lenticular pores in UO2, (U,Pu) O2 and PuO2

Lenticular pore migration rates in oxide muclear fuels were experimentally measured in out-of-pile heating experiments. It is deduced that those pores which are in part responsible for the formation of columnar grains, are only produced in the absence of relevant amounts of filling gas. Specimens containing important concentrations of He, produced by Pu alpha decay, show columnar grain restructuring by grain boundary migration. Some consequences are drawn concerning the possible role played by lenticular pores in the mechanisms of fission gas release from nuclear fuels.     

Calorimetric measurements on (U,Th)O2 solid solutions

Enthalpy increments of urania - thoria solid solutions, (U_0.10Th_0.90)O₂, (U_0.50Th_0.50)O₂ and (U_0.90Th_0.10)O₂ were measured by drop calorimetry in the temperature range 479 - 1805 K. Heat capacity, entropy and Gibbs energy function were computed. The heat capacity measurements were carried out also with differential scanning calorimetry in the temperature range 298 - 800 K. The heat capacity values of (U_0.10Th_0.90)O₂, (U_0.50Th_0.50)O₂ and (U_0.90Th_0.10)O₂ at 298 K are 59.62, 61.02, 63.56 J K⁻¹ mol⁻¹, respectively. The results were compared with the data available in the literature. From the study, the heat capacity of (U,Th)O₂ solid solutions was shown to obey the Neumann - Kopp’s rule.     

(U,Gd)O2芯块试验与工业生产

文章介绍在宜宾核燃料元件厂(YFP)生产线上进行(U,Gd)O2芯块工业规模的生产试验及产品合格性鉴定,对试验结果进行讨论和评价.结果表明YFP的(U,Gd)O2芯块生产线完全具备工业生产能力,并实现(U,Gd)O2芯块制造的国产化.     

Modelling of deposition processes on the TiO2 rutile (1 1 0) surface

Deposition of Ti_xO_y clusters onto the rutile TiO₂ (1 1 0) surface has been modelled using empirical potential based molecular dynamics. Deposition energies in the range 10-40 eV have been considered so as to model typical deposition energies of magnetron sputtering. Defects formed as a function of both the deposition energy and deposition species have been studied.The results show that in the majority of cases Ti interstitial atoms are formed, irrespective of whether Ti was contained within the deposited cluster. Furthermore that the majority of these interstitials are formed by displacing a surface Ti atom into the interstitial site. O surface atoms are also relatively common, with Ti and TiO₂ surface units often occurring when the deposited cluster contains Ti but becoming less frequent as the deposition energy is increased. Structures that would give rise to the growth of further layers of rutile are not observed and in the majority of the simulations the energy barriers for diffusion of the end-products is high.     

In-pile creep study on mixed oxide (U,Pu)O2

The in-pile creep of a mixed oxide UO₂-PuO₂ under compression was studied up to fission rate of $\text{6 × 10}{}^{13}\text{f cm}{}^{\mathrm{?3}}\text{s}{}^{\mathrm{?1}}$, for stresses up to 26.5 MN m^?2, at temperatures ranging from 700 to 900°C. The results obtained agree with those of other authors. The creep rate is proportional to the applied stress and to the fission yield. However, it is athermal within the temperature range explored and is not affected by the burn-up, which has so far reached 30000 MWd t^?1 (3.6% FIMA). When the sample is under compression the fuel swells under the action of the fission products formed in the oxide during its irradiation. The swelling rate is about that commonly accepted for a clad fuel element. Finally it seems that the oxide swells more when free from stress than when subjected to a stress field, but this point has to be confirmed.     

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.     

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 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.     

Thermal conductivity of (U,Pu,Np)O2 solid solutions

The thermal conductivities of (U_,Pu_,Np)O₂ solid solutions were studied at temperatures from 900 to 1770 K. Thermal conductivities were obtained from the thermal diffusivity measured by the laser flash method. The thermal conductivities obtained below 1400 K were analyzed with the data of (U,Pu,Am)O₂ obtained previously, assuming that the B-value was constant, and could be expressed by a classical phonon transport model, λ = (A + BT)⁻¹, A(z₁, z₂) = 3.583 × 10⁻¹ × z₁ + 6.317 × 10⁻² × z₂ + 1.595 × 10⁻² (m K/W) and B = 2.493 × 10⁻⁴ (m/W), where z₁ and z₂ are the contents of Am- and Np-oxides. It was found that the A-values increased linearly with increasing Np- and Am-oxide contents slightly, and the effect of Np-oxide content on A-values was smaller than that of Am-oxide content. The results obtained from the theoretical calculation based on the classical phonon transport model showed good agreement with the experimental results.     

Molecular dynamics simulations of C2, C2H, C2H2, C2H3, C2H4, C2H5, and C2H6 bombardment of diamond (1 1 1) surfaces

The sticking and erosion of C₂H_x molecules (where x=0-6), at 300 and 2100 K onto hydrogenated diamond (1 1 1) surfaces was investigated by means of molecular dynamics simulations. We employed both quantum-mechanical and empirical force models. Generally, the sticking probability is observed to somewhat increase when the radical temperature increases and strongly decrease with increasing number of H atoms in the molecule.     

Interferences in electron emission from O2 by 30 MeV O impact

Interference structures in the ejected electron spectra for 30 MeV O^5,8+ + O₂ are investigated. The measured electron yields were studied for electron energies from 5 to 400 eV and observation angles of 30°, 60°, 90°, 120° and 150° with respect to the incident beam direction. Experimental molecular cross-sections were normalized to theoretical molecular one-center cross-sections revealing oscillatory structures suggestive of secondary interferences as evidenced by the independence on the observation angle. An oscillation interval for 30 MeV O^5,8+ + O₂ of Δk ∼ 4 a.u. is found, a value two times larger than that previously observed for 3 MeV H⁺ + N₂. No obvious evidence for primary Young-type interferences was seen.     

Crack formation,crack healing and porosity redistribution during irradiation of UO2 and (U,Pu) O2

The irradiation-induced void volume redistribution in the fuel was analysed. The radial crack volume and porosity distributions, the central radii and the radial gap width were measured after irradiation and compared with the calculated values. Short-time (He-loop experiments in the FR2 reactor), medium-time (bundle irradiation in the BR2 reactor) and long-time (trefoil-irradiation in the DFR reactor) irradiated fuel pins were examined. The model of pore migration, used in the computer code SATURN-la, is based on the evaporation-condensation mechanism. Measured swelling rates were extrapolated to higher temperatures and used. The crack volume distribution was calculated on the basis of a multifractured fuel model. One can conclude from the comparison between calculated and measured void volume distributions that several mechanisms redistribute void volume. These are crack formation, crack healing, migration of sinter pores and fission gas bubbles, gas swelling, evaporation-condensation phenomena in the region of the central void, irradiation-induced sintering and increase in diameter of the cladding.     

The mechanisms of fission gas release from (Th,U)O2

The releases of xenon from three (Th, U)O₂ specimens with different U contents were measured over a wide range of fission dose from $\text{2.9 × 10}{}^{19}$ to $\text{2.2 × 10}{}^{22}$ fissions m^?3 by using a post-irradiation technique. The releases were found to decrease with dose and to level off at higher doses. Measurements of the changes in lattice parameter and specific surface area of the same specimens enabled one to conclude that the decrease in release originates in the trapping of xenon by the vacancies and vacancy clusters induced by fission fragments. And the release mechanisms of fission gas were proposed based on the proper evaluation of the observation on radiation damage and recovery in oxide fuel.