Chemical corrosion of highly radioactive borosilicate nuclear waste glass under simulated repository conditions

This review summarizes the results of the joint Japanese (Central Research Institute of Electric Power Industry, CRIEPI, Tokyo), Swiss (National Cooperative for the Storage of Radioactive Waste, NAGRA, Baden), Swedish (Swedish Nuclear Fuel and Waste Management Company, SKB, Stockholm) international 'JSS' project on the determination of the chemical durability of the French nuclear waste borosilicate glass, which was completed in 1988. Radioactive and non-radioactive glass specimens were investigated. A data base was created with results from glass corrosion tests performed with different water compositions, pH values, temperatures, sample surface areas (S), solution volumes (V), and flow rates. Glass corrosion tests were performed with and without bentonite and/or steel corrosion products present. Variation of the glass composition was taken into account by including the borosilicate glass 'MW' in the investigations, formulated by British Nuclear Fuels, plc. An understanding was achieved of the glass corrosion process in general, and of the performance of the French glass under various potential disposal conditions in particular. A special effort was made to establish a corrosion data base, using high S/V ratios in the experiments in order to understand the glass durability in the long term. A computer program, GLASSOL, was developed, based on a dissolution-precipitation model, to calculate the glass water reaction. Fair agreement between observations and the model calculations was achieved. Use of a constant (time-independent) long-term rate was justified by observations on naturally altered basaltic glasses of great age, which were compatible with what was inferred from experiments with nuclear waste glasses in the laboratory. A fractured glass block would not be altered within 10 000 years at 90 degree C (flow rate <100 L/year)