Swedish Nuclear Fuel Research Articles

The Äspö Hard Rock Laboratory—a step toward the Swedish final repository for high-level radioactive waste

Abstract Sweden's energy consumption produces about 250 metric tons of spent nuclear fuel annually. In order to meet the country's growing need for high-level radioactive waste disposal, the Swedish Nuclear Fuel and Waste Management Company (SKB) has developed a waste management system that will ensure safe handling of all of Sweden's radioactive waste. To prepare for the siting and licensing of the final disposal site, SKB is constructing an underground research laboratory, the Aspo Hard Rock Laboratory, which will replace the underground laboratory at Stripa Mine that has been in operation since 1977. This article describes the R & D objectives of the Hard Rock Laboratory, as well as site characterization, layout and construction of the laboratory, which began in October 1990.

Modelling of bedrock stability (abstract)

"Modelling of bedrock stability (abstract)." Geologiska Föreningen i Stockholm Förhandlingar, 112(4), p. 327

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 nonradioactive 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 highS/Vratios in the experiments in order to understand the glass durability in the long term.

The Swedish Repository for Low and Intermediate Reactor Waste- SFR. Radioactivity Release and Transport Calculations

The Swedish repository for low and intermediate level waste- SFR was ready for operation in early 1988. The license to start operation for the low level waste part was given in the spring of the same year. A very detailed safety analysis of the repository has been made by the SWEDISH NUCLEAR FUEL AND WASTE MANAGEMENT CO.- SKB. The main issues of this work are presented in this paper.

Issues in atmospheric and oceanic modeling

The Swedish Nuclear Fuel and Waste Management Company (SKB) is conducting a series of long term tests at the Äspö Hard Rock Laboratory to assess the behaviour of the bentonite buffer under conditions similar to those expected in a KBS-3 repository for high level nuclear waste. The LOT A2 experiment consists of a vertical borehole with a central heater inside a copper tube surrounded by compacted bentonite. During four years, the temperature of the copper tube was maintained at 130 °C, while the bentonite was progressively water saturated by the injection of groundwater. During this period, physical and hydro-geochemical data were collected. By using the code TOUGHREACT, a model was made to simulate the processes of solute transport which control the chemical and the mineralogical distribution observed in the bentonite at the end of the test. Additionally, a series of sensitivity analyses was performed to assess the influence of key parameters controlling the thermal–hydro-geochemical processes. Numerical results indicate that, within the first year, the heated bentonite blocks are completely water saturated, which agree with the measured data. The simulated transport of chloride, the dissolution/precipitation of Ca sulphate minerals, and the cation redistribution in the montmorillonite interlayer also agree with data measured at the end of the experiment.

Effects on Fissure Calcite by Recharge Water - Some Problems Related to 14C-Dating of Groundwater

AbstractThe modelling of water flow in the bedrock is of especial importance for safetyassessementof radioactive waste disposal in crystalline rocks. Age dating of the water has been tried for the validation of ground water transport models. Thus, several tritium and 14C-analyses have been carried out on water samples from deep boreholes drilled within SKB's (Swedish Nuclear Fuel and Waste Management Company) study sites. However, several processes will influence the final content of 14C in the groundwater measured and correction for enrichment and dilution are hard to carry out. Another contribution to the understanding of recharge water transport by using frequency and stable isotope analyses of calcite fissure fillings is suggested.

Radiolysis of Groundwater from HLW Stored in Copper Canisters

ABSTRACTA large number of computer calculations of the radiolysis of ground water outside copper canisters have been carried out within the Swedish Nuclear Fuel safety Project.At dose rates higher than 5 × 10−2rad/s the hydrggen concentration is constant, approximately 10−3mol/dm3, and independent of time and dose rate. Therefore the rate of difiusion out of the system is constant, approximately 7 × 10−4mol/year. At dose rates lower than 5 × 10−4rad/s the logarithm of the hydrogen production is proportional to the logarithm of the dose rate. The hydrogen production depends on fuel type, burn-up and canister thickness only to the extent that the dose rate varies with these parameters. The hydrogen production is independent of the volume of water in which the energy is assumed to be absorbed (5 or 15 cm layers outside the surface of the canister).The hydrogen production depends on the concentration of Fe2+ions in the water. In the basic calculations the concentration was assumed to be 5 ppm. An increase to 50 ppm increases the total hydrogen production after one million years outside a 1 cm thick canister from 360 to 1000 mol. A decrease to 0.5 ppm decreases the same yield to 80 mol.