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Conference Article: Heat treatment of graphite and resulting tritium emissions

J. Wörner, W. Botzem, S.D. Preston

Abstract

Pile I still retains significant amounts of Wigner Energy within graphite bricks not affected by, the 1957 fire. A recently undertaken survey could clarifv the status of the graphite with respect to stored energy. Based on the history of Pile I it is not possible to present a unique set of release curves valid for all investigated channels. Furthermore peripheric blocks not submitted to any survey, due to the lack of being accessible by trepanning machinery are unknown with respect to possible stored energy.

Graphite samples of several blocks show release behaviour that exceeds the heat capacitv of unirradiated graphite. StatisticalIy this characteristic correspondence is given only for less than one quarter of the material. Three quarters do not represent any risk in the sense that either handling or later encapsulation of the graphite will result in unforeseeable energy release.

For the purpose of planning all stages of the graphite-disposal process the question is addressed whether an inadvertent release of stored energy can occur during handling and storage and whether deliberate or partial annealing of the material is a requirement. On the basis of a multi-activation energy concept it could be cleared that annealing between 200 and 250oC is able to clear all sites eventually being activated by the grouting process. This temperature range was examined with respect to the efficiency of a deliberate annealing procedure. 80 to 90percent of the stored energy releasable tip to 500oC can be released if the annealing procedure is executed between 250 and 300oC.

As far as tritium release from the graphite during an annealing procedure is concerned measurements of the desorption of tritium as effect of heating of grounded Pile I graphite material were executed. Apparently less than 0.5 percent of the total tritium content are released at the proposed annealing temperatures.

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key words: Gas Cooled Reactor, Nuclear Technology
Reference:
IAEA Technical Committee Meeting on "Nuclear Graphite Waste Management", held from 18-20 October 1999 in Manchester, United Kingdom
International Atomic Energy Agency, Vienna (Austria)
TCM-Manchester99, pp:65-76