The Institute for Transuranium Elements ͑ITU͒ is building up an accurate database of actinide behavior in chloride melts in support of its nuclear fuel reprocessing development program. The electrochemical properties of uranium, dissolved in LiCl-KCl eutectic melt, were investigated by transient electrochemical techniques, such as cyclic voltammetry ͑CV͒ and chronopotentiometry on an inert tungsten electrode. It was shown that U 4+ is reduced to U 0 by a two-step mechanism corresponding to U 4+ /U 3+ and U 3+ /U 0 transitions. In the 400-550°C ͑673-823 K͒ range, the diffusion coefficients of U 3+ and U 4+ were similar and equal to: D͑U 3+ ͒ = 13.7 ϫ 10 −4 exp͕−24.2 ϫ 10 3 /RT͑K͖͒ and D͑U 4+ ͒ = 5.25 ϫ 10 −4 exp͕−19.8 ϫ 10 3 /RT͑K͖͒ cm 2 s −1 . The apparent standard potentials of U 4+ /U 3+ and U 3+ /U 0 redox systems were E o* ͑U 4+ /U 3+ ͒͑V͒ = −1.902 + 0.0006104T ͑K͒ vs. Cl 2 /Cl − and E o* ͑U 3+ /U 0 ͒͑V͒ = −3.099 + 0.0007689T ͑K͒ vs. Cl 2 /Cl − , respectively. Some thermochemical properties ͑⌬G * ,␥͒ of uranium solutions were also derived from the electrochemical measurements. The Gibbs free energies of dilute solution of UCl 3 and UCl 4 in the LiCl-KCl were determined to be: ⌬G * = −897.09 + 0.226T ͑K͒ and −183.53 + 0.0589T ͑K͒ in kJ mol −1 , respectively. In the 400-550°C ͑673-823 K͒ range, the activity coefficients ␥ of UCl 3 and UCl 4 range between 0.3 − 4.5 ϫ 10 −3 and 17.1 − 12.7 ϫ 10 −3 , respectively.Pyrochemical reprocessing is a promising method for recycling spent nuclear fuel. Uranium is the major element in most nuclear fuel cycles, thus the assessment of accurate thermochemical data for this element in molten salts is of high importance. Uranium behavior has been studied in many chloride-based melts: for example, MgCl 2 -NaCl-KCl, 1 NaCl-KCl, 2,3 LiCl-NaCl-CaCl 2 -BaCl 2 , 4 and LiCl-KCl eutectic. [5][6][7][8][9][10][11][12][13][14][15] In the literature, the first electrochemical studies on uranium behavior in molten chlorides appeared in 1959. 1,7,8 It was established that the reduction of UCl 4 to uranium metal proceeds via a two-step mechanism, i.e., U 4+ is reduced into U 3+ before forming uranium metal. In a MgCl 2 -NaCl-KCl mixture, Hill et al., 1 who used a platinum needle microelectrode, reported a threeelectron system for the U 3+ /U 0 transition. The apparent standard potential was close to −2.467 V vs. Cl 2 /Cl − at 450°C ͑723 K͒. In the NaCl-KCl eutectic, by electromotive force ͑emf͒ measurements, Flengas 2 measured E o* ͑U 3+ /U 0 ͒͑V͒ = 0.000592T ͑K͒ − 2.973522 vs. Cl 2 /Cl − and E o* ͑U 4+ /U 3+ ͒͑V͒ = 0.000492T ͑K͒ − 1.808311 vs. Cl 2 /Cl − in the 670-850°C ͑943-1123 K͒ range. In the NaCl-KCl eutectic, Serrano et al. 3 showed by cyclic voltammetry ͑CV͒ that the reduction U 3+ to uranium metal is reversible and proceeds through a three-electron mechanism. The diffusion coefficient of U 3+ was determined equal to 5.5 ϫ 10 −5 cm 2 s −1 at 715°C ͑983 K͒. Recently, in LiCl-NaCl-CaCl 2 -BaCl 2 baths, Poa et al. 4 studied the electrochemical reduction of uranium U 3+ to uranium metal b...

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This work presents a study on the electrochemical behavior of actinides (An ϭ Am and Pu͒ and lanthanides (Ln ϭ La and Nd͒ onto solid aluminum cathodes in a molten LiCl-KCl eutectic at 733 K. Cyclic voltammetry of these elements onto Al working electrode is carried out to estimate the reduction potentials of An and Ln and to predict the efficiency of an An/Ln separation by electrolysis. Results show that the reduction of Am 3ϩ , Pu 3ϩ , La 3ϩ , and Nd 3ϩ ions leads to the formation of An-Al or Ln-Al surface alloys. Moreover, the difference in reduction potentials between An and Ln elements was found to be larger than on the classical molten cadmium or bismuth electrodes. Pu/Nd and Am/Nd electroseparations were carried out using Al foam as cathode and a plutonium rod as anode. Efficient and faradaic reduction of actinides was achieved at a cathodic potential preventing Nd coreduction. After electrolysis, a compact and adherent deposit covered the cathode surface, and scanning electron microscopy, energy-dispersive X-ray analysis of the deposit indicated the existence of Al 4 -Pu or mixed Am-Pu-Al alloys. Analysis of the dissolved Al cathodes, after electrolytic separation of Am from Nd, showed that 536 mg of An ͑94% of the total Am present in the initial salt phase͒ had been deposited compared to less than 3 mg Nd. This work represents a breakthrough in the selective partitioning of An by electrorefining methods.

In this work a hot demonstration of the DIAMEX process using the new reference molecule DM-DOHEMA (N,N -DiMethyl-N,N -DiOctylHexylEthoxyMalon-Amide) and genuine high-level PUREX raffinate as feed is reported. The continuous counter-current experiment was carried out in a 16-stage centrifugal extractor battery, installed in a hot cell. In order to produce a representative High Level Liquid Waste (HLLW) a PUREX process was applied on dissolved fuel using the same equipment. It was demonstrated that 5 extraction stages were sufficient to achieve feed decontamination factors above 2200 and 320 for Am and Cm, respectively. Co-extraction of molybdenum and zirconium, as well as of palladium, were efficiently prevented using oxalic acid and HEDTA scrubbing, respectively. However, technetium, yttrium, and to some extent ruthenium, were coextracted. The back-extraction proved to be very efficient, yielding in 4 stages more than 99.9% recovery of Am and Cm. The extraction profiles were modeled with computer code simulations and the results compared with data obtained from an experiment using the former reference molecule DMDBTDMA (DiMethylDiButylTetraDecylMalonAmide).