Study of uranium and iodine behavior in the molten fluorides LiF-ThF₄ at 650 C

The Molten Salt Fast Reactor (MSFR) is an innovative concept of GEN IV developed by the CNRS since 2004. It is currently studied in the framework of the European project SAMOFAR of H2020. The MSFR operates with a liquid nuclear fuel consisting of a mixture of fluoride salts LiF-ThF₄- (UF₄/UF₃) (77.5-20-2.5) mol% melted at high temperature (700-900 C). This reactor is particularly advantageous for the thorium fuel cycle (²³²Th-²³³U). This concept proposes an integrated reprocessing of the nuclear fuel based on pyrochemical methods in order to extract the fissile material and to separate the actinides from the fission products. A scheme for the treatment of the fuel salt, proposed in a previous European project (EVOL, FP7), is based on the redox and acido-basic properties of the elements produced by the fission and capture reactions occurring in the reactor core. The baseline for this scheme was initially thermodynamic. Experimental validation is currently under way to study the chemical and electrochemical behavior of the molten salt and the elements solubilized therein. Previous studies on molten salt reactors can only be partially used for this concept because the composition of the MSFR salt defined by the European EVOL project is different from the composition of the salts proposed up to now for this type of reactor. However, the diffusion and activity coefficients depend on the physicochemical properties of the molten salt (in particular solvation) and in previous studies we have shown that the solvation properties of molten salts are strongly dependent on their nature and their composition.The objectives of this thesis are the electrochemical characterization of the molten mixture LiF-ThF₄ and the study of the electrochemical behavior of uranium and iodine.The electrochemical study of the behavior of uranium shows the stability of two soluble species (UF₄ and UF₃) of this element in the molten medium and the possibility of its reduction to the metallic state. This point is important because the co-existence of these two compounds will make possible to control the potential of the fuel salt in the core of the reactor in order to limit the corrosion reactions with the structural materials. The activity coefficients of U(IV) and U(III) were determined. The values obtained show that the solvation of uranium to the degree of oxidation (IV) by fluoride ions is much greater than that of uranium to degree (III), which is in agreement with subsequent observations in other fluoride salts. Our choice for the study of fission products in the fuel salt has focused on iodine. In the core of the reactor, the stable form of the iodine is the soluble halide form I^- and in the general scheme of treatment of the fuel salt, it is planned to extract iodine by a fluorination step in order to produce the gaseous compound I₂. The electrochemical study shows the contribution of a chemical reaction to the electrochemical oxidation of iodide ions in gaseous iodine. This redox chemical reaction corresponds to the oxidation of the iodide ions by oxygen. This reaction is explained for the existence of a soluble thorium oxyfluoride ThOF₂. Extraction efficiencies of I₂ (g) greater than 95% were obtained by electrolysis at controlled potential. These electrolysis, which simulate fluorination process, make it possible to validate the method for the extraction of the iodine in the reprocessing scheme.This research has led to a better understanding of salt stability and of the chemical and electrochemical behavior of several compounds (U and I) in the molten salt. (author)