Thermodynamic Simulation and Experimental Study of Irradiated Reactor Graphite Waste Processing with REE Oxides
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Thermodynamic Simulation and Experimental Study of Irradiated Reactor Graphite Waste Processing with REE Oxides Olga K. Karlina, Vsevolod L. Klimov, and Galina Yu. Pavlova Moscow SIA “Radon”, 2/14, 7-th Rostovsky per., Moscow, 119121, Russian Federation Michael I. Ojovan Immobilisation Science Laboratory, Department of Engineering Materials, University of Sheffield, S1 3JD, UK ABSTRACT Thermochemical processing of reactor graphite waste is based on self-sustaining reaction 4Al + 3TiO2 + 3C = 3TiC + 2Al2O3 which chemically binds 14C from the irradiated graphite in the titanium carbide. Thermochemical processing was investigated to analyse the behaviour of rare earth elements (REE), where REE = Y, La, Ce, Nd, Sm, Eu and Gd. Both thermodynamic simulations and laboratory scale experiments were used. The REEs in the irradiated reactor graphite are formed as activation products of impurities and spread over the graphite bricks surfaces as well as arise from fission of nuclear fuel. REEs can be used also to substitute for waste actinides as well as to increase the durability of carbide-corundum ceramics relative to waste actinides. Thermodynamic calculations and X-ray diffraction analysis of ceramic specimens synthesized revealed that durable REE’s aluminates with perovskite, β-alumina and garnet structures are formed by interaction of REE oxides with the Al2O3 melt during the selfpropagating reaction of ceramic formation. The porous carbide-corundum ceramics synthesized have a high hydrolytic durability, e.g. the normalised leaching rates of 137Cs, 90Sr and Nd are of the order of 10–7 – 10–8 g/(cm2⋅day). INTRODUCTION Operation of uranium-graphite reactors resulted in production of significant amounts of irradiated graphite waste in the form of dust, powder, chips and lumps. This waste resulted both from technological operations and incidents and contains inclusions of metallic and ceramic nuclear fuel and other reactor components. This waste contains activation products such as 10Ве, 14 С, 36Cl, 41Ca and 59Ni, actinides in fuel particles such as 234,236,238U, 237Np, 239,240,242Pu, 243Am and 242,243,244Cm, decay products such as 134,135,137Cs, 90Sr, 151Sm and 154,155Eu . Similar waste can be produced when cleaning-decontaminating surfaces of graphite blocks heavily contaminated by radionuclides. Among the radionuclides present in irradiated graphite the long-lived radionuclide 14С with half-life of 5730 years is particularly hazardous. 14С is readily incorporated into organic matter molecular structure of living species including humans, moreover it enters into RNA and DNA molecules [1].
Disperse radioactive wastes intended for long-term storage and disposal in Russia must be immobilised (consolidated) accordingly to Russian regulatory requirements, which were developed accounting for IAEA recommendations [2]. To solidify disperse irradiated graphite waste the self-sustaining high-temperature synthesis (SHS) based on exothermic chemical reaction 3C(graphite) + 4Al + 3TiO2 = 3TiC + 2Al2O3 was suggested in Russia [3–5]. Th
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