Long-Term Dissolution Behavior of Spent Fuel in Compacted Bentonite and Synthetic Granitic Groundwater
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Long-Term Dissolution Behavior of Spent Fuel in Compacted Bentonite and Synthetic Granitic Groundwater K.S. Chun, S.S. Kim and C.H. Kang Korea Atomic Energy Research Institute, P.O. Box 105, Yusong, Taejon, 305-600, Republic of Korea ABSTRACT The long-term dissolution behavior of spent PWR fuel in synthetic granitic groundwater has been investigated since June of 1998 in order to identify the release mechanism of spent fuel in contact with domestic Ca-bentonite, which was compacted as a density of 1.4g/cm3. Several spent fuel specimens were cut to around 3-mm thick, and these specimens were then loaded into each leaching cell. Every half year, one specimen was collected and the bentonite block was sliced into several pieces. The element distribution on the specimen surface was measured by EPMA and an optical microscope, and the gamma activities in the sliced bentonite pieces were measured. All the leachates were sampled and then the gamma and gross alpha activities were measured. By the results, coming from leaching up to about 1.5 years, the fraction of radiocesium released through the compacted bentonite layer from the spent fuel is approximately a hundredth lower than that without the bentonite layer. By the depth profile of the gamma activity on the bentonite, the activity gradually decreased in accordance with more distance from the specimen and the fractional release rates of cesium were nearly the same. On the other hand, the EPMA results indicate that more time and more information is required in order to identify the surface alteration of spent fuel by leaching. This experiment will therefore be continued at least into the year 2006.
INTRODUCTION Many experimental results have been reported for the dissolution behavior of spent fuel and unirradiated UO2 in water under various conditions [1-7]. Most of the studies on the dissolution of UO2 pellets with bentonite have been concerned with the formation of an alteration phase and precipitates which are a solubility-limiting solid phase of uranium in bentonite water. A lot of second phases and solubilities of radionuclides in bentonite water have recently been identified.
A few experiments have been tried for the identification of the dissolution mechanism of spent fuel in repository conditions [8.9], but it is not well known with the lack of real data. Release and transport models of radionuclides in compacted bentonite are assumed to be dominated by diffusion for performance assessment. Data from a uranium ore body on the release of dissolved species has being been collected for comparison with a near field release model. Most of the experiments with clay for buffer material have been done for the short-term period and/or in bentonite water, that bentonite is added to water. However, a few experimental studies have been performed for the interaction of spent fuel, compacted clay and groundwater. The purpose is to get information on the corrosion behavior of spent fuel and the release rate of radionuclides from fuel within domestic bentonite and synthetic gr
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