Michel Ougier

Electroseparation of Actinides from Lanthanides on Solid Aluminum Electrode in LiCl-KCl Eutectic Melts

University of Grenoble, ENSEEG, 38402 Saint Martin d’He`res, FranceThis work presents a study on the electrochemical behavior of actinides (An 5 Am and Pu! and lanthanides (Ln 5 La and Nd!onto solid aluminum cathodes in a molten LiCl-KCl eutectic at 733 K. Cyclic voltammetry of these elements onto Al workingelectrode is carried out to estimate the reduction potentials of An and Ln and to predict the efficiency of an An/Ln separation byelectrolysis. Results show that the reduction of Am

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.

Study of Molten Salt Electrorefining of U-Pu-Zr Alloy Fuel.

Electrorefining of unirradiated metal alloy fuel, U-20Pu-10Zr (% by weight), was carried out in an Ar atmosphere cell to obtain a basic knowledge of spent fuel treatment. Before electrorefining, the Pu ion concentration in LiCl-KCl electrolyte was adjusted to 4 wt.% using pure Pu metal and CdCl2. Cylindrical alloy fuel was anodically dissolved in the salt electrolyte. Electrotransport to either liquid Cd cathode or solid cathode was carried out several times in galvanic mode. The deposit on the solid cathode mostly consisted of metal U and entrained salt and took place at high current efficiency. The recovery of Pu and Am into a liquid cadmium cathode was also demonstrated with high current efficiency. The anodic dissolution of the fuel alloy was found to progress from the outside whilst leaving a dense salt layer on the alloy surface. The peculiar fluctuation of anode potential during U-Pu-Zr dissolution was explained by the competitive oxidation of U-Pu and Zr.

Low-Burnup Irradiation Behavior of Fast Reactor Metal Fuels Containing Minor Actinides

Abstract Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and/or rare earths (REs) have been irradiated in the fast reactor PHÉNIX to examine the effects of adding those elements on metal fuel irradiation behavior. In this experiment, two MA-containing metal fuel pins, in which the test alloys U-19Pu-10Zr-2MA-2RE and U-19Pu-10Zr-5MA/U-19Pu-10Zr-5MA-5RE (wt%) were loaded into part of a standard U-19Pu-10Zr alloy fuel stack, and a reference fuel pin of U-19Pu-10Zr alloy without MAs or REs was set in an irradiation capsule. Two other capsules with this same configuration are also irradiated. Postirradiation examinations are conducted at ~2.5, ~7, and ~11 at.% burnup. For the low-burnup fuel pins, nondestructive tests after irradiation have been performed, and the integrity of the pins was confirmed. The irradiation behavior of MA-containing metal fuels up to 2.5 at.% burnup was analyzed using the ALFUS code. The calculation results, such as the axial swelling distribution of a fuel slug or the extrusion behavior of bond sodium to the gas plenum, are consistent with the measurement data regardless of the addition of MAs and REs to the U-Pu-Zr alloy fuels. This observation result indicates that the macroscopic irradiation behavior of U-Pu-Zr fuels containing MAs and REs of 5 wt% or less is similar to that of U-Pu-Zr fuels up to ~2.5 at.% burnup.

Equilibrium Distribution of Actinides Including Cm Between Molten LiCl-KCl Eutectic and Liquid Cadmium

Equilibrium distribution of actinides both in molten LiCl-KCl eutectic and liquid cadmium were measured from the concentration data obtained in electrorefining tests and reductive extraction tests. Separation factors for U, Np, Am, Cm against Pu were derived in the practical temperature range of 700 K to 783 K. The derived separation factors are consistent with the reported values measured at 773 K and 723 K. The temperature dependence for Cm is different compared to the other actinides (U, Np and Am). This behavior remains unclear and additional experimental measurements of distribution coefficient of Cm are required before ruling on the real behavior.

Recovery of Transuranium Elements from Real High-Level Liquid Waste by Pyropartitioning Process

A pyropartitioning process is under development to recover minor actinide elements from high-level liquid waste (HLLW) generated by Purex reprocessing. This pyropartitioning process consists of a denitration step that converts various elements in the HLLW into oxides, a chlorination step that converts the oxides into chlorides, and a reductive-extraction step that separates the actinide elements from fission products (FPs) in the chlorination product. The feasibility of each step was confirmed using simulating FPs and unirradiated transuranium elements (TRUs). In the present study, approximately 520 g of real HLLW was prepared to demonstrate the feasibility of the pyropartitioning process. Almost 100% of each TRU originally contained in the HLLW was recovered in the liquid cadmium phase in the reductive-extraction step, which showed that the expected chemical reactions were completed and the mass loss of TRUs was negligible in the denitration, chlorination, and reductive-extraction steps. The separation behaviors of actinide elements, including americium and curium, from FPs in the reductive-extraction step were quite similar to those observed in previous experiments using unirradiated materials. Hence, the pyropartitioning process was successfully verified.

Development of Pyrochemical Separation Processes for Recovery of Actinides from Spent Nuclear Fuel in Molten LiCl-KCl

Abstract Effective fuel utilization and waste minimization are required to provide a sustainable and safe energy generation for the future of nuclear reactors. To obtain an efficient fuel cycle it is essential to develop an efficient and selective recovery of the key elements from spent nuclear waste. This necessitates that Am and Cm can be separated selectively from lanthanide fission products. This chapter describes two pyrochemical (dry) processes. Both processes are based on the dissolution of metallic spent fuel in a fused LiCl-KCl eutectic mixture. In the first process, U is collected selectively on a solid iron cathode; U, Pu, and minor actinides are then recovered simultaneously using a liquid Cd cathode. The technique is also used to recover actinides from oxide fuel—the oxide ions are oxidized to CO2 and CO at a carbon anode. The second process is based on electrorefining in a molten LiCl-KCl bath using a reactive solid aluminum cathode. The fuel is dissolved electrochemically into the carrier melt, forming a mixture of actinides and fission product ions. All actinides are selectively recovered together from the melt by electrodeposition on solid aluminum cathodes in the form of solid actinide–aluminum alloys. Data concerning the basic electrochemistry of actinides in a LiCl-KCl eutectic are given. Advantages and drawbacks of the pyrochemical processes are discussed.