Perspective Schemes of Conditioning of Liquid Radioactive Waste of Ukrainian Nuclear Power Plants

The existing world experience of practical use of sorption technology and technology of cementing liquid borne radioactive waste of nuclear power plants (NPP) with water-water energetic reactors (WWER) to obtain a product 1suitable for transfer to disposal facilities is considered. It has been concluded that salt cake accumulated in NPP storage facilities is a major problem that determines the further choice of the development and implementation of conditioning technologies. Currently, 18,000 salt cake containers stored at the Zaporizhzhia NPP and Khmelnitskiy NPP storage facilities have exceeded their design life. A possible solution is to change the classification of the salt cake and to classify it as solid radioactive waste. It is noted that the existing tax system for the generation of radioactive waste in Ukraine does not contribute to the choice of conditioning technologies aimed at minimizing the volume of the final product. The prospect of application of the technology of melting in the “cold crucible” for one-stage formation from a evaporator bottoms and salt cake borosilicate glass, guaranteed capable in the conditions of surface disposal to ensure the isolation of radionuclides during the time required for decay to a safe level of radioactivity. It is proposed to create a melting unit according to the modular scheme, when several parallel crucibles with capacity up to 20 kg/h with cheaper highfrequency generators with a capacity of 160 kW are connected to common ventilation system. The urgency of carrying out technical and economic analysis to determine the optimal 56 ISSN 2311-8253 Nuclear Power and the Environment № 3 (18) 2020 solutions for the introduction of effective and economically sound technologies for the processing of evaporator bottoms and salt cake at Ukrainian NPPs is emphasized.

Advanced Vitreous Wasteforms for Radioactive Salt Cake Waste Immobilisation

Salt cake radioactive waste is a remnant solid salt concentrate after deep evaporation of radioactive evaporator concentrate at WWER NPP’s. The traditional cementing of borate-containing liquid radioactive waste, to which the salt cake belongs, leads to a significant increase in the volume of the final product. This work describes borosilicate vitreous wasteforms developed to immobilize radioactive salt cake waste and comprises data on both glass synthesis and characterization. The composition of glass selected for the purpose of immobilisation of the salt cake radioactive waste allows to include up to 40 wt. % of the oxides contained in the salt cake and to reduce the volume of the final product by more than 2 times compared with the cement compound. The batches were melted in a cold crucible melter at 1200 °C. The normalized cesium leaching rate of the vitrified wasteform product was within range 3.0·10-5 – 3.7·10-6 g/(cm2·day).