Studies on the physicochemical properties of borosilicate glass developed for radioactive waste

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To simultaneously vitrify Ion Exchange Resin, Zeolite, and Dry Active Waste generated from Korean Nuclear Power Plant, a borosilicate glass system was formulated. Viscosity and electrical conductivity of the glass were measured and within the desired ranges at the processing temperature. Those activation energies were evaluated as 152 and 70.46 kJ/mol within a temperature range of 1223 to 1623 K, respectively. Time-Temperature-Transformation study was performed using data from heat treatment. The hematite crystal was found within a temperature range of 823 to 1123 K. Mossbauer spectroscopy showed about 42% Fe^<2+> state existed in the glass produced from operation of the pilot-scale plant. Product Consistency Test performed from 7 to 120 d in the glass showed the leach rates of B, Na, Li and Si were much less than those of the benchmark glass. International Organization for Standardization test was performed at 363 K for 1022 d and shown that Cumulative Fraction Leached values of Na, Li, and Si were saturated below the fraction of 0.4 except that the leaching of B increases continuously. About 50 μm thickness layer was observed to be as a protective layer against continuous corrosion. According to Vapor Hydration Test, the corrosion rate of the glass was 2 g/m^2/d and met the specification (50 g/m^2/d). Based on Soxhlet leaching accomplished at 371 K for 30 d, the weight loss of the glass was determined as 106.8 g/m^2 which was lower than those of other HLW (High-Level radioactive Waste) glasses.
社団法人日本セラミックス協会の論文
2008-03-01

Studies on the physicochemical properties of borosilicate glass developed for radioactive waste

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