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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Summary. The selective partitioning (P) of long-lived minor actinides from highly active waste solutions and their transmutation (T) to short-lived or stable isotopes by nuclear reactions will reduce the long-term hazard of the high-level waste and significantly shorten the time needed to ensure their safe confinement in a repository. The present paper summarizes the on-going research activities at Forschungszentrum Jülich (FZJ), Karlsruher Institut für Technologie (KIT) and Institute for Transuranium Elements (ITU) in the field of actinide partitioning using innovative solvent extraction processes. European research over the last few decades, i.e. in the NEWPART, PARTNEW and EUROPART programmes, has resulted in the development of multi-cycle processes for minor actinide partitioning. These multi-cycle processes are based on the co-separation of trivalent actinides and lanthanides (e.g. by the DIAMEX process), followed by the subsequent actinide(III)/lanthanide(III) group separation in the SANEX process. The current direction of research for the development of innovative processes within the recent European ACSEPT project is discussed additionally. This paper is focused on the development of flow-sheets for recovery of americium and curium from highly active waste solutions. The flow-sheets are verified by demonstration processes, in centrifugal contactors, using synthetic or genuine fuel solutions. The feasibility of the processes is also discussed.