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Abstract

Studies on thermal tritium release in tritiated Zircaloy embedded in cement mortar showed a better retention of tritium as compared with WAK cement stone, especially at increased temperatures. Lixiviation attempts on active Zircaloy fuel cans in WAK cement stone using H-2O, saturated NaCl solution and quinary alkaline have been completed. When compared with clear fuel cans, cemented fuel cans show with H-2O an increased tritium leaching, and a reduced (by several orders of magnitude) lixiviation of actinides. The useful time of WAK cement stone in contact with quinary alkaline at RT amounts to approximately 1-1,5 years before decomposition of the cement matrix occurs; afterwards, an increased level of tritium release is to be expected. A calorimeter was developed for the determination of the capacity of nuclear heat sources from active waste embedded in cement stone in bundles of 200 l. The design of the calorimeter was optimized for an integral heat capacity of 10 W. An electrically heated barrel simulator was used for calibration of the calorimeter. It was possible to simulate the effect of local nuclear hot-spots in different positions within the barrel by connecting heating cartridges. The experiments showed that the capacity determination is not affected by local overheating in the barrel.

Additional information

Authors: GRABNER H, KERNFORSCHUNGSZENTRUM KARLSRUHE GMBH (GERMANY);KAPULLA H, KERNFORSCHUNGSZENTRUM KARLSRUHE GMBH (GERMANY);FROTSCHER H KERNFORSCHUNGSZENTRUM KARLSRUHE GMBH (GERMANY), KERNFORSCHUNGSZENTRUM KARLSRUHE GMBH (GERMANY)
Bibliographic Reference: EUR 9108 DE (1984) FS, 45 P., BFR 200, EUROFFICE, LUXEMBOURG, POB 1003
Availability: Can be ordered online
Record Number: 1989124023200 / Last updated on: 1987-01-01
Category: PUBLICATION
Available languages: de
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