Extraction mechanisms of uranyl ions by 1-butyl-3-methylimidazolium nonafluorobutanesulfonate containing N-dodecyl-2-pyrrolidone

Extraction of U(VI) in HNO₃ to 1-butyl-3-methylimidazolium nonafluorobutanesulfonate (BMINfO) by using N-dodecyl-2-pyrrolidone as an extractant has been investigated. With increasing the concentration of HNO₃ from 0.01 to 3.0 mol/dm³ (M), distribution ratio decreased. This suggests that uranyl ions are extracted by ion exchange with a cation component of BMINfO, i.e. BMI⁺. The amount of BMI⁺ transferred to aqueous phase accompanied by extraction of U(VI) was evaluated with ¹H NMR measurement. Plots of the amount of U(VI) extracted versus the amount of BMI⁺ transferred from BMINfO phase to the aqueous phase indicated linear relationship. The slope of the line was about 1.0 and 0.63 in the extraction system of U(VI) performed in 0.1 and 1.0 M HNO₃, respectively. This result means that two types of extraction mechanisms exist depending on the concentration of HNO₃. One is the ion exchange mechanism. Another is the ion-pair extraction mechanism.     

Burn up extension with slightly enriched uranium and plutonium fuel in CANDU reactors

A study of fuel burn-up and concentrations of uranium and plutonium isotopes for the three fuel cycles of a CANDU reactor are carried out in the present work. The infinite and effective multiplication factors are calculated as a function of fuel burn up for the natural UO₂ fuel, 1.2% enriched UO₂ fuel and for the 0.45% PuO₂-UO₂ fuel. The amount of ²³⁵U and ²³⁸U consumed and ²³⁹Pu, ²⁴⁰Pu and ²⁴¹Pu produced in the three fuel cycles are also calculated and compared.     

Electrodeposition of metallic uranium at near ambient conditions from room temperature ionic liquid

The electrochemical behavior of U(IV) in the room temperature ionic liquid (RTIL), N-methyl-N-propylpiperidinium bis(trifluoromethylsulfonyl)imide (MPPiNTf₂), was investigated to evaluate the feasibility of using the RTIL for non-aqueous reprocessing application. In this context, the rate of dissolution of uranium oxide (UO₂) in HNTf₂ was studied at 353 K. The dissolution of UO₂ in HNTf₂ was rapid; nearly 50% of UO₂ dissolved within 3 h and more than 95% dissolved in 25 h. The resultant solution was dried, diluted with MPPiNTf₂ and the electrochemical behavior of U(IV) in MPPiNTf₂ was studied at 373 K at platinum, glassy carbon and stainless steel electrodes. The cyclic voltammograms of U(IV) in MPPiNTf₂ at platinum and glassy carbon electrodes consisted of four cathodic waves occurring at a peak potentials of −0.7 V (Fc/Fc⁺), −1.4 V, −2.2 V and −2.7 V. Controlled potential electrolysis of a solution of U(IV) in MPPiNTf₂ at −2.8 V (Fc/Fc⁺) resulted in the deposition of metallic uranium, which was confirmed by X-ray diffraction and scanning electron microscopy.     

Uranium oxide weathering: spectroscopy and kinetics

Particles of UO_2+x (x≅0.16 ± 0.06) exposed to the atmosphere react by oxidation and formation of complexes (hydrates, hydroxides and carbonates). Surface reactions alter and erode the UO₂ particles. This paper outlines results for measurements of oxidation rates on uranium oxide particles using in situ photoluminescence spectroscopy (PL), X-ray photoelectron spectroscopy (XPS) and secondary ion mass spectrometry (SIMS). Phosphorescence spectra observed during oxidation of UO_2+x were attributed to U(VI) in uranyl-type coordination and in octahedral coordination. Uranyl-type spectra formed during wet oxidation of UO_2+x, and U(VI) octahedral spectra formed during dry oxidation of UO_2+x. The uranyl-type species, although more stable, is more kinetically labile for vacuum reduction than is the octahedral U(VI). Oxidation of U(IV) species are diffusion controlled. Vacuum reduction of uranyl U(VI) in UO₃ follows a field-enhanced cationic diffusion rate law, while re-oxidation follows a diffusion rate law. Post-oxidation core and valence band XPS and SIMS measurements provided qualitative and quantitative measures of uranium oxidation states near uranium oxide surfaces.     

Fast Reactor Operating on Uranium–Plutonium Oxide Fuel in a Self-Regulatable Regime

Mathematical simulation is used to show that it is possible to develop a fast reactor operating on uranium–plutonium oxide fuel (UO₂)_1–x (PuO₂) _x , the same for all fuel elements in the core, and with uranium carbide in breeding elements with heavy coolant (PbBi eutectic). A self-regulatable regime is obtained in the reactor. This enhances safety while minimizing control. Tailings uranium with 0.1% ²³⁵U and a mixture of plutonium isotopes, which is obtained from spent fuel, making it possible to conduct operation in an actinide-closed fuel cycle, is used in the fuel and uranium carbide. ²³⁸U is actually consumed in the reactor, but most fission products are produced from ²³⁹Pu.     

UO2 dissolution in Boom Clay conditions

The solubility of uranium dioxide (UO₂) was measured in real and synthetic Boom Clay waters with varying concentrations of humic acids and carbonate under reducing conditions at 20 °C. Uranium concentrations in function of time suggest the reduction of U(VI) to U(IV) by the humic acids which is occurring faster in real clay water than in synthetic clay waters. Humic acids induce also a competition to complex U(VI) in carbonate-containing solution, but they are not able to control the uranium concentration at high bicarbonate concentration (0.02 mol dm⁻³). Nevertheless they may play a role at low carbonate concentration. In our experimental conditions, the geochemical calculations indicate that two uranium secondary phases (U₄O₉ and UO_2(c)) are susceptible to control the uranium concentration in solution. These calculations are in good agreement with results of the X-ray photoelectron spectroscopy. At the end of tests, uranium concentrations reach steady-state values between 3 × 10⁻⁸ and 5 × 10⁻⁸ mol dm⁻³ in the bicarbonate-rich solutions. Although these concentrations are considered as conservative, they are 10-100 times higher than in natural Boom Clay. The consequence is that spent fuel could slowly dissolve in the interstitial clay water undersaturated with respect to UO₂/UO_2+x of the fuel.     

Electrochemical Studies on Uranium in the Presence of Organic Acids

We examined electrochemical redox reactions of UO_{2 2+} in perchlorate and organic acid (oxalic, malonic, succinic, adipic, L-malic, and L-tartaric acids) solutions using cyclic voltammetry to reveal the effects of complex formation with organic acids on the redox behavior. In the perchlorate and organic acid solutions, a redox reaction of UO_{2 2+}/UO_{2 +} and an oxidation reaction of U(IV) produced by a disproportionation of UO₂ ⁺ were observed. The peak potentials of the UO₂ ²⁺ reduction showed a good linear relationship with the stability constants of 1:1 UO₂ ²⁺-organic complexes. In the presence of malonic acid, the redox potential for UO₂ ²⁺/UO₂ ⁺ was constant at pH 1-2 and 5-6 while it decreased with an increase in pH from 2 to 5. Additionally, it was independent of malonate concentration at 0.1–0.5 M while it decreased with an increase in the concentration from 0.005 to 0.1 M. Based on the experimental and the speciation calculation results, we determined the redox reactions of UO₂ ²⁺-malonate complexes as a function of pH and malonate concentration. We also determined the redox reactions of UO₂ ²⁺-oxalate complexes in the same way.     

In-situ Measurement of UO2 2+ Concentration in Molten NaCl-2CsCl by Differential Pulse Voltammetry

In order to assess the applicability of the Differential Pulse Voltammetry technique to the in-situ measurement of UO₂ ²⁺ concentration in the oxide electro-winning process, DPV measurements for UO₂Cl₂ in molten NaCl-2CsCl were studied. DPV measurement of UO₂ ²⁺ in NaCl-2CsCl at 923 K, with a set of optimized parameters (potential sweep rate ?0.1 V/s, pulse cycle 0.1 s, pulse width 10 ms and pulse potential height 50 mV), showed a clear current peak at ?0.9 V vs. Cl₂/Cl^?. This was attributed to the reduction of UO₂ ²⁺ to UO₂ ⁺. The relation between the current peak height and the analytical concentration of the UO₂ ²⁺ showed good proportionality in the concentration region up to 0.06 mol.l ^?1, and the applicability of UO₂ ²⁺ concentration measurement by DPV was confirmed up to 0.4 mol.l ^?1. In order to assess the interference by the coexisting fission product elements to the measurement of UO₂ ²⁺ concentration, DPV measurements of UO₂ ²⁺ concentration in molten NaCl-2CsCl containing PdCl₂, NdCl₃, SmCl₃ and CeCl₃ were also performed. Even before removing Pd, the current peak at ?0.9 V vs. Cl₂/Cl^? by the reduction of UO₂ ²⁺ to UO₂ ⁺ was found to be distinguishable from the reduction currents of Pd²⁺ to Pd at ?0.7 V vs. Cl₂/Cl^?. As a result, the application of DPV measurement technique to the in-situ monitoring of UO₂ ²⁺ concentration in the oxide electro-winning method requires the improvement of the DPV measuring condition or the electrode structure on higher UO₂ ²⁺ concentration condition.     

EPR Study of Uranium (V) Species in Photo-and Electrolytic Reduction Processes of UO2 NO3)2-2TBP

Abstract

Electron Paramagnetic Resonance (EPR) and optical spectra of uranium(V) species were observed in both processes of photo- and electrolytic reduction of UO₂(NO₃)₂-2tributylphosphate (TBP) in 80%TBP-n-dodecane solution. The formation of U(V) species was detected by an optical spectrum (λ_max: 770, 970 and 1,420 nm). EPR signal with the value of ff-factor –2.3 and a linewidth of approximately 1,100 Gs was observed during the electrolytic reduction. On the other hand, during the photoreduction the signal with the value of fil-factor –1.94 was observed and there was found a superhyperfine structure with the intensity ratio of 1:2:1, that is caused by the superhyperfine coupling with nuclear spin, I=1/2, of the strongly coordinated ³¹P to the central uranium through oxygen atom. The superhyperfine coupling constant was estimated to be 27 Gs. Moreover, the signal with the value of g-factor –2.00 due to an organic radical was observed. The residue after the thermo-gravimetric analysis of UO₂(NO₃)₂-2TBP was identified as α-UP₂O₇ by the powder X-ray diffraction analysis, indicating the strong coordination of TBP to the central uranium atom.     

Finite element analysis of local overheating within plutonium enriched UO2 fuel rods caused by PuO2 islands

Within natural UO₂ fuel elements enriched with plutonium, this last material should form PuO₂ solid solutions inside the UO₂ pellets, in a wide range of concentrations. If the solutions are obtained by mechanical mixing of the oxides, PuO₂ islands are formed in the UO₂ matrix. These islands may be the source of several problems in the fuel behaviour, the most important being the overheating of the matrix in the neighbourhood of the particles. It is caused by the large fission cross section of plutonium compared with that of uranium.A detailed study of the thermal effects produced by PuO₂ particles in the UO₂ matrix and the cladding is then important for the specification of their permissible size. A portion of the fuel rods with spherical particles in the most significant places was studied. In order to obtain the dimensionless overheating of the fuel and cladding produced by the presence of those particles, the spacial distribution of temperature was calculated, solving the stationary and linear bidimensional equation of heat conducting using a finite element code. Several geometrical variables and material properties have been taken as dimensionless parameters. A satisfactory convergence of the numerical results to an asymptotic limit with a wellknown exact solution, has been obtained.     

Swelling results from helium-pressurized 304L stainless steel tubes

Thermodynamic properties of uranium oxides with O/U atomic ratios between 2.04 and 2.34, principally in the nonstoichiometric UO_2+x single-phase region, were determined at temperatures in the range 500°C to 1100°C by electromotive force measurements with cells of the type Ni-NiO¦ZrO₂ (+CaO)¦uranium oxide. The relative partial molar free energies of oxygen for the single-phase region UO_2+x and the two two-phase coexisting regions UO_2+x-U₄O_9?y and U₄O₉-U₃O_8?z were obtained with adequate precision. Several statistical models for the defect structure of UO_2+x have been reviewed. The variation of the relative partial molar entropy of oxygen with composition in the UO_2+x phase was derived on the basis of the 2:2:2 defect complex model proposed from the neutron diffraction study of Willis. The entropies were computed to fit the experimental data assuming that the excess oxygen atoms entered into the fluorite structure in pairs and that the available sites for such oxygen pairs and U⁵⁺ ions would be partially restricted for occupancy.     

Direct electrochemical reduction of solid uranium oxide in molten fluoride salts

The direct electrochemical reduction of UO₂ solid pellets was carried out in LiF-CaF₂ (+2 mass.% Li₂O) at 850 °C. An inert gold anode was used instead of the usual reactive sacrificial carbon anode. In this case, oxidation of oxide ions present in the melt yields O₂ gas evolution on the anode. Electrochemical characterisations of UO₂ pellets were performed by linear sweep voltammetry at 10 mV/s and reduction waves associated to oxide direct reduction were observed at a potential 150 mV more positive in comparison to the solvent reduction. Subsequent, galvanostatic electrolyses runs were carried out and products were characterised by SEM-EDX, EPMA/WDS, XRD and microhardness measurements. In one of the runs, uranium oxide was partially reduced and three phases were observed: nonreduced UO₂ in the centre, pure metallic uranium on the external layer and an intermediate phase representing the initial stage of reduction taking place at the grain boundaries. In another run, the UO₂ sample was fully reduced. Due to oxygen removal, the U matrix had a typical coral-like structure which is characteristic of the pattern observed after the electroreduction of solid oxides.     

Electrolysis of Burnup-Simulated Uranium Nitride Fuels in LiCl-KCl Eutectic Melts

The electrochemical behavior of burnup-simulated uranium nitride fuels containing representative solid fission product elements, UN+Mo (Mo = 2.84 wt%), UN+Pd (Pd = 4.6 wt%) and (U, Nd)N (NdN = 8.0 wt%), was investigated in the molten LiCl-KCl eutectic salt with 0.54 wt% UCl₃ in order to clarify the effects of fission products on the dissolution of actinide nitrides and the behavior of FPs in the electrorefining of spent nitride fuel. The rest potentials of burnup-simulated UN pellets were similar to that of pure UN. The electrochemical dissolution of UN began at about _0:75V vs Ag/AgCl reference electrode in all samples as well as that of pure UN. After the electrolyses at the constant anodic potential of ?0:65––0:60V vs Ag/AgCl, most of UN was dissolved into LiCl-KCl as UCl₃ at the anode, and U was recovered in the liquid Cd cathode in all samples. Furthermore, Nd was dissolved at the anode and accumulated into LiCl-KCl as NdCl₃, while Mo and Pd were not dissolved but remained at the anode.     

微氧条件下硫酸盐还原菌颗粒污泥处理废水中铀(Ⅵ)的实验研究

通过驯化培养得到了以硫酸盐还原菌(SRB)为优势菌属的活性颗粒污泥,研究了微氧条件(氧浓度为0.6~1.0mg/L)下其对废水中U(Ⅵ)的去除性能,利用傅里叶红外光谱(FTIR)、扫描电镜(SEM)和X射线能谱(EDS)分析了其对U(Ⅵ)的去除机理,采用连续提取法考察了铀在颗粒污泥上的沉积形态。实验结果表明,微氧条件下SRB颗粒污泥结合并转化U(Ⅵ)的过程主要分为两步:初期(前30min)吸附和后期微生物还原沉淀作用,SRB颗粒污泥对U(Ⅵ)的去除率达98.89%。FTIR、SEM和EDS分析表明,颗粒污泥表层上酰胺基、羧基、羟基、磷酸基等均参与了对U(Ⅵ)的沉积,同时Na+、Mg2+等与UO2+2存在离子交换作用。形态分析表明,在颗粒污泥上,铀主要以残渣态形式沉积,生物有效性差,可迁移能力弱,不易造成二次污染。     

Study of homogenisation and cation interdiffusion in mixed UO2-PuO2 compacts by X-ray diffraction

Homogenisation in mixed UO₂-PuO₂ compacts has been studied by X-ray diffraction. It is observed that the homogenisation proceeds, mainly, by the assimilation of UO₂ into PuO₂. This near one-way flow of material, from UO₂ to PuO₂, is shown to be due to high activity (large BET surface area) of the PuO₂ powder as compared with that of the UO₂ powder.An X-ray line profile analysis method of determining various mixed composition fractions in sintered mixed compacts has been used to evaluate homogeneity in terms of the fraction of UO₂ that has gone into PuO₂. A concentric core-shell diffusion model, in which UO₂ forms a solute core and PuO₂ forms a solvent shell, was used to determine cation interdiffusion coefficients from the homogenisation data. The temperature dependence of cation diffusivity in the range 1573–1873 K is obtained as $\text{D = 2.55 × 10}{}^{\mathrm{?11}}\text{exp(?2.22 × 10}{}^5\text{/8.31 T)}$ m²/s. The low value (222 kJ/mole) of activation energy for cation interdiffusion is attributed to the hypostoichiometry of the mixed compacts studied.The diffusivity values at 1573 and 1673 K separately give an activation energy of 126 kJ/mole, which suggests grain-boundary diffusion as the primary mechanism of homogenisation in this temperature range.     

Irradiation-induced creep of mechanically blended porous UO2-PuO2 fuel

With regard to the behaviour of fast breeder reactor fuel, irradiation creep of mechanically blended, porous U_natO₂-15% PuO₂ was investigated. Some results for UO₂ are also quoted to clarify the dependence of creep rate on stress and temperature. The sintered density of the UO₂-PuO₂ samples amounted to 86% TD and 93,5% TD, their irradiation temperatures were between 300 and 1000°C, the stress in the samples at 15 and 40 MN/m², the fission rates between 2.5 and $\text{5 × 10}{}^{\mathrm{?9}}$ f/(U + Pu)-atom · s, and the maximum burnup at about 1%. The creep rates of UO₂-PuO₂ are much higher than previously measured on UO₂ of high density, but there was a good correspondence of the stress and temperature dependence. The difference of the creep rates cannot be explained only by the porosity of the UO₂-PuO₂ samples. Therefore the PuO₂ portion of the fuel, whose distribution is heavily inhomogeneous, is treated as additional “effective” porosity. By this means a suitable interpretation is obtained for the results below about 650°C. At higher temperatures, UO₂-PuO₂ of 86% TD showed a rapid initial densification up to about 93% TD, apparently together with a simultaneous homogenization of the fission-density distribution. The results measured could be interpreted without considering an influence of the Pu-content as such.     

Neutron Diffraction Studies of NaUO3

Neutron diffraction studies have been made on a powdered NaUO₃ sample. It has been shown that an octahedron composed of six oxygen atoms around pentavalent uranium in this compound is distorted and four of six oxygen atoms are much closer to an uranium atom (U-O distance of 2.15Å) than the other two (U-O distance of 2.24Å). A unit of the UO₂ ⁺ ion does not exist in the crystal, though the pentavalent uranium has been expected to be in the form of UO₂ ⁺ as in the case of UO₂ ²⁺ of a hexavalent uranium Ion. A neutron diffraction pattern at temperature of liquid helium has shown no appreciable difference from that at temperature of liquid nitrogen.     

Electrochemical and Spectroelectrochemical Studies on Uranyl Carbonato and Aqua Complexes

Abstract

The electrochemical reduction of uranyl ion has been studied by the cyclic voltammetry (CV) and the spectroelectrochemical method using optically transparent thin-layer electrode (OTTE) in basic carbonate and acidic aqueous solutions. The CV measurements showed that U(V) was formed quasireversibly by the reduction of uranyl ion in both basic carbonate and acidic Perchlorate solutions. Uranium(V) was found to be much more stable in the carbonate media and its absorption spectrum was measured by the OTTE cell under the applied potentials varying stepwise from ?0.7 to ?0.9V vs. Ag/AgCl (3 M NaCl). On the other hand, in the acidic solution the absorption spectra recorded at 20s interval by the OTTE cell under the fixed applied potential at -0.4 V vs. Ag/AgCl (3 M NaCl) exhibited only the spectral change from uranyl ion to the final reduction product U(IV). It has been suggested based on the present electrochemical study that uranyl is stable for the redox reaction in the deep groundwater, which is in reductive environment of —0.3 V (vs. NHE) and contains carbonate ion, but is reduced to U(IV) in the acidic groundwater in the same reductive environment.     

Influence of certain halogen compounds on the oxidation of U(IV) in a sulfuric medium

The work presented here showsthat, by adding the salts NaI, NaF, NaBr, and NaC1 to the mixture during the oxidation of UO₂ by manganese dioxide and ammonium persulfate the degree of solution of the uranium is increased. The fluoride ion is outstanding inthis, as it brings about the complete dissolution of the UO₂. In this article we consider possible mechanisms of the interaction between U(IV) and the fluoride ion. It is shown that Cu²⁺ catalyzes the UO₂ oxidation by ammonium persulfate and atmospheric oxygen in an acid medium.Translated from Atomnaya Énergiya, Vol. 16, No. 2, pp. 130–134, February, 1964     

Helium behavior in oxide nuclear fuels: First principles modeling

UO₂ and (U, Pu)O₂ solid solutions (the so-called MOX) nowadays are used as commercial nuclear fuels in many countries. One of the safety issues during the storage of these fuels is related to their self-irradiation that produces and accumulates point defects and helium therein.We present density functional theory (DFT) calculations for UO₂, PuO₂ and MOX containing He atoms in octahedral interstitial positions. In particular, we calculated basic MOX properties and He incorporation energies as functions of Pu concentration within the spin-polarized, generalized gradient approximation (GGA) DFT calculations. We also included the on-site electron correlation corrections using the Hubbard model (in the framework of the so-called DFT + U approach). We found that PuO₂ remains semiconducting with He in the octahedral position while UO₂ requires a specific lattice distortion. Both materials reveal a positive energy for He incorporation, which, therefore, is an exothermic process. The He incorporation energy increases with the Pu concentration in the MOX fuel.