Feasibility of immobilizing fluorinated pyrochemical reprocessing salts in a glass-ceramic matrix
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Feasibility of immobilizing fluorinated pyrochemical reprocessing salts in a glass-ceramic matrix Agnès Grandjean CEA Valrhô – Marcoule DTCD/SCDV/LEBV ABSTRACT Spent fuel reprocessing by an innovative reductive extraction process in a molten fluoride medium (LiF/AlF3) is now being evaluated; in this hypothesis, all the unrecoverable fission products would be conditioned as fluorides. A preliminary study was undertaken to assess the feasibility of incorporating these fluorides by melting in a glass-ceramic matrix. The containment matrix for the fluorinated waste stream was selected after examining the consequences of fluorinated compounds on the vitreous state and on the physical and chemical properties of the melt and the solidified glass. The presence of fluorinated compounds in the raw materials used to produce the vitreous material raises the problem of the volatility of some fluorides, of their solubility in the melt, and of possible crystallization of the material. INTRODUCTION Reprocessing of spent nuclear fuel with group extraction of the actinides by an innovative process in fluoride media is currently being evaluated. The process involves separation by reductive salt-metal extraction. After dissolving the fuel or target in a salt bath, the noble metal fission products are first extracted by contacting them with a slightly reducing metal (e.g. zinc); the actinides are then selectively separated from the remaining fission products (mainly lanthanides) by contact with a reductive alloy (chemical reducing agent + metal solvent). The salt/metal system optimized to enhance An/Ln separation should be LiF-AlF3 / Al- [1]. Aluminum is used both as a reducing agent and as a metal solvent. During extraction, the fluoride bath becomes concentrated with radionuclide fluorides including alkali metals (Cs, Na, etc.), alkaline earth metals (Sr, Ba, etc.) and a few rare earth elements. These fission products modify the initial properties of the salt, for example by raising its melting point and modifying its thermochemical characteristics. The fluorinated salt must then be regenerated or replaced. The degraded salt containing notably 137Cs and 90Sr is a highlevel wasteform consisting mainly of the LiF/AlF3 solvent and fission product fluorides. We therefore examined the feasibility of immobilizing this waste in a borosilicate matrix. The development of glass (or glass-ceramic) compositions suitable for a given wasteform must take three criteria into account: • Waste loading in the matrix, i.e. the mass fraction of waste incorporated into the containment matrix. The higher the waste loading, the lower the cost of the ultimate waste package. The objective is to maximize the waste loading while maintaining compatibility with the other two criteria. • The feasibility of fabricating the matrix using the selected melting process. The process temperature must be a tradeoff between a temperature high enough to enhance the homogeneity of the melt and a temperature low enough to minimize the radionuclide volatility.
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