Thermodynamic modeling and experimental tests of irradiated graphite molten salt decontamination
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Thermodynamic modeling and experimental tests of irradiated graphite molten salt decontamination Olga Karlina, Michael Ojovan, Galina Pavlova and Vsevolod Klimov Moscow SIA «Radon», 119121, Moscow, 7-th Rostovsky per., 2/14, Russia ABSTRACT Molten salt flameless oxidation of graphite is one of the prospective methods of irradiated graphite waste processing. Molten salts are capable to retain a considerable part of radionuclides, to neutralize acidic off gases, moreover spent salts could be vitrified on completion of the process. We have used thermodynamic modelling to assess the efficiency of molten salt oxidation of graphite. Equilibrium compositions of both the melt and the off gas were calculated depending on graphite content and temperature. The feasibility of decontaminating the irradiated graphite of its near-surface layers using complete molten salt oxidation was investigated in a series of laboratory experiments. As the molten salt medium used to oxidize irradiated graphite we have investigated lithium, potassium and sodium carbonates. Sodium sulphate, boron oxide, barium and potassium chromates were also used as oxidizers. Tests were carried out at 870–1270 Ʉ. The efficiency of decontamination of graphite blocks has been assessed based on the activity of 137Cs and 60ɋɨ in the samples before and after molten salt oxidation. Data obtained demonstrated the feasibility of decontamination by molten salt removal of near surface layers on irradiated graphite blocks. Decontamination rate and efficiency depend on oxidizers used and temperature of process. INTRODUCTION The amount of accumulated irradiated graphite waste is continuing to grow worldwide. Decommissioning of uranium-graphite reactors generates the main part of irradiated graphite waste where the graphite has been used to moderate and reflect neutrons. More than a hundred of such reactors are located within UK, France, former USSR, USA and Spain. Technical solutions and industrial technologies are not yet available to immobilize the irradiated graphite contaminated with nuclear fuel inclusions. Processing methods for irradiated graphite such as cementation, covering and impregnation with resins, melting with low-melt alloys, micro-capsulation, self-propagating high-temperature synthesis, do not provide any volume reduction of waste [1]. Considerable waste volume reduction (up to 1 – 3 % relative to initial volume) can be achieved using incineration. However incineration is typically used to treat low and intermediate level waste only. Besides that, radioactive ashes resulting from incineration, aerosols and acidic off gases are inherent to the combustion process and that makes the process complicated and expensive. Flameless molten salt oxidation (MSO) is one of the most promising methods to treat irradiated graphite waste [2,3]. MSO-based technology does not require fine crushing of waste graphite. Using molten mineral salts as the oxidizing medium provides a number of advantages such as universality, the capability to retain a considerable part of radio
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