Electrochemical Reduction of (U,Pu)O2 in Molten LiCl and CaCl2 Electrolytes

The electrochemical reduction of UO₂-PuO₂ mixed oxides (MOX) was performed in molten LiCl at 923 K and CaCl₂ at 1,123 K to evaluate the behavior of the plutonium quantitatively and to define the optimum conditions for the electrochemical reduction of those materials.

In LiCl, excess deposition of lithium metal can be avoided and the MOX was smoothly reduced at ?0.65 V vs. Bi-35 mol% Li reference electrode. The reduction ratio calculated from the mass change of the samples taken during the electrochemical reduction and the ratio evaluated by gas-burette method were in good agreement. The cathodic current efficiency remained 30–50% mainly due to the deoxidation of tantalum cathode basket. Although dissolution of plutonium and americium into the electrolyte was found by the chemical analysis, the dissolved amount was negligible and had no immediate influence on the feasibility of the electrochemical reduction process.

In CaCl₂, reduction of the MOX occurred in whole range of the tested cathode potential (?0.15 V to ?0.40 V vs. Ca-Pb reference electrode). The cathodic current efficiency was around 30%. Although the MOX was completely reduced at ?0.25 V, the reduction was interrupted by formation of the surface barrier made of the reduced material and the vacancy between the reduced and the non-reduced areas at ?0:30 V. Plutonium and americium dissolved also into the CaCl₂ electrolyte to slightly higher concentrations than those observed in LiCl electrolyte. The analyses for the reduction products showed that the amount of those actinides lost from the cathode was much larger than that found in the electrolyte, probably due to the formation of mixed oxide precipitate.     

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.     

Fabrication of high-density UO2 and (U0.75Pu0.25)O2 by hot pressing

The hot pressing behavior of hyperstoichiometric UO₂ and (U, Pu)O₂ powders has been evaluated. Specimens with densities in excess of 99% $\text{ρ}{}_{\mathrm{th}}$ can be fabricated with this technique at temperatures of 1000°C or less.     

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 behavior of (U,Pu) mixed-oxide fuels

Transpiration experiments were performed at temperatures between 1000 and 1700°C to study the thermodynamics and defect structure of hypostoichiometric UO₂-20 wt % PuO₂ solid-solution systems. The oxygen partial pressures were established by using flowing $\text{H}{}_2\text{/H}{}_2\text{O}$ mixtures. After equilibration, the quenched products were analyzed by chemical, X-ray, neutron-diffraction and metallographic techniques. The versus temperature curves for the mixed oxide with different oxygen-to-metal ratios were plotted.Based on X-ray, neutron diffraction and metallographic data, it was concluded that the 20 wt % PuO₂ mixed oxide exists as a single phase under normal conditions, even at an oxygen-to-metal ratio as low as 1.92.The data from density measurements and neutron-diffraction analysis indicated that the predominant defects in the hypostoichiometric UO₂-20 wt % PuO₂ are anion vacancies.     

Dimensional Changes in Irradiated (Th,U)O2

Effects of irradiation on the dimension and microstructure in (Th,U)O₂ pellets were examined by measurements of lattice parameter and bulk density changes, and observations of pore structures. The concentrations of fission-induced defects and the damage volume were estimated by a simple model. Both macroscopic and microscopic dimensional changes were found to increase initially with fission dose and then fall off. The difference between macroscopic and microscopic ingrowths increased with dose, suggesting that fission-induced interstitials would cluster or go to sinks and the concentration of vacancies would be in excess of that of interstititials. The damage volume for vacancies was estimated to be about 1x10^?22m³·fiss.^?1, and almost agreed with that for fission Xe release. Observations of the pore structure indicated that the volume fraction of pores smaller than 2–3 μm decreases with irradiation and the distribution of pore size shifts toward the larger side.     

Investigation of ThO2 and (U, Th)O2

The effect of the properties of ThO₂ and (U, Th)O₂ powders, prepared with different technological regimes, on the properties of the finished items is investigated. The work includes detailed investigations of ThO₂ and (U, Th)O₂ powders (x-ray phase analysis, electron-microscope investigation) and sintered fuel pellets (determination of density, study of microstructure, thermophysical investigations). The temperature dependences of the crystal lattice parameters and the sizes of the crystallites in ThO₂ and (U, Th)O₂ powders with different UO₂:ThO₂ ratio are obtained. The temperature dependences of the thermal conductivity of sintered ThO₂ and (U, Th)O₂ pellets with different UO₂:ThO₂ ratio are studied.     

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.     

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.     

Thermal conductivity of near-stoichiometric (U, Ce)C and (U, Pu, Ce)C solid solutions

The thermal conductivities of near-stoichiometric (U, Ce)C and (U, Pu, Ce)C solid solutions containing CeC up to 10 mol% were determined in the temperature range from 740 to 1600 K by the laser flash method. The thermal conductivity decreased with the cerium content in the solid solutions. The electrical resistivities were also measured for the purpose of analyzing the heat conduction mechanism. It was found that the decrease of electronic heat conduction caused by the addition of cerium resulted in decreasing the thermal conductivities of (U, Ce)C and (U, Pu, Ce)C compared with UC and (U, Pu)C.     

H2O2对Pu(Ⅴ)的还原动力学研究

在0.1 mol/L NaClO4溶液中研究了Pu(Ⅴ)与H2O2反应的动力学。测定了Pu(Ⅴ)与H2O2的反应速率。探讨了温度以及Fe2 ,SO42-,HCO3-,F-等无机离子的存在对反应的影响。实验结果表明,反应对Pu(Ⅴ)与H2O2呈一级,对溶液中H 呈-1级;速率方程可表示为:-dc(Pu(Ⅴ))dt=(3.93±1.93)×10-9c(Pu(Ⅴ))c(H2O2)c(H )。随着温度升高,反应速率明显加快,根据Arrhenius规律,计算出了反应的活化能为Ea=84 kJ/mol。地下水中Fe2 ,SO42-,HCO3-,F-等离子的存在,有利于Pu(Ⅴ)的还原。