[en] The molten salt fast reactor (MSFR) is one of the six nuclear reactor concepts retained during the Forum GEN IV in 2001. The particularity of this concept is to use a liquid fuel consisting of a molten salt, LiF-ThF₄-UF₄ /UF₃ (77-19-4 mol%) and to have an integrated spent fuel treatment process. This treatment consists of successive chemical separation steps based on redox and acid-base properties of the elements produced in the reactor by nuclear reactions: soluble and gaseous fission products, metals elements and soluble minor actinides. One of the major steps of the treatment method is a reducing extraction which consists to contact the molten salt and a liquid metal, bismuth, containing the reducing element, lithium. This step allows separating the minor actinides and lanthanides. Minor actinides are reintroduced in the nuclear reactor to be burned while the lanthanides are confined in deep storage. The work in this thesis had two objectives: (i) assess the feasibility of reducing extraction of actinides and lanthanides, a step that had previously only been validated on the basis of thermodynamic calculations and (ii) study the chemistry of molten fluoride salts (and especially the fuel salt) by developing a methodology for the determination of fundamental data such as the activity coefficients in fluorides media, coefficients activities which quantify the solvation properties. To experimentally realize a reducing extraction, the first step is to prepare a metal layer of liquid Bi-Li with predefined composition. An electrolysis technique in molten salt LiCl-LiF at 550 deg. C was chosen to achieve these metal solutions. We have shown that only this molten medium could be used for the manufacture of such metal alloys. Extraction tests were then carried out by contact between LiF-ThF₄ (with UF₄ and NdF₃ are introduced to simulate respectively the actinides and lanthanides) and Bi-Li at 650 deg. C. The main results show that the extraction of neodymium and uranium was obtained with yields of around 1% and 15% respectively in the best conditions. These values are low compared to previous thermodynamic calculations. Low efficiency of the extraction is due to a simultaneous extraction of thorium in the liquid metal phase which forms intermetallic compounds at the metal/salt interphase and blocks the transfer. Methods have been developed to achieve fundamental data that are lacking in molten fluoride medium, in particularly the solvation properties. Speciation of some metallic cations by fluoride ions with high temperature was particularly studied and calculation of complexation constants by simulated experimental results was done. Carried out for two lanthanides, neodymium and lanthanum, two actinides, thorium and uranium, and also for a transition metal, nickel, this study achieves to calculate the activity coefficients of these elements in different fluoride molten salt. The study of the speciation of thorium was an important step to understand the chemistry of the fuel salt LiF-ThF₄. We were able to calculate the activity coefficient of the fluoride ion in this environment at 650 deg. C. Finally, all of this work allows giving a first estimate of the reactivity of each element of the periodic table (present in the nuclear reactor after operation) at each stage of the treatment of the spent fuel salt. (author)