Definition of pre-disposal conditioning for radioactive graphite blocks from reactor decommissioning
The decommissioning of gas-graphite reactors in the EC (e.g. French UNGGs, British Magnox reactors and AGRs, and reactors in Spain and in Italy) will produce large amounts of graphite blocks. The aim of the study is to examine the behaviour of graphite waste and to develop a conditioning technique which makes this waste acceptable for shallow land disposal sites. 18kg of graphite core samples with an outside diameter of 74mm were removed from the G2 gas-cooled reactor at Marcoule. Their radioactivity was highly dependent on their position inside the reactor. Measured results indicate an activity range of 100-400 MBq/kg with 90% tritium, 5% 14C, 3% 60Co, 1.5% 63Ni. Repeated porosity analyses showed that open porosity ranging from 0 to 100 micron exceeded 23 vol% in the graphite. Water penetration kinetics were investigated in unimpregnated graphite and resulted in impregnation by water of 50-90% of the open porosity. Preliminary lixiviation tests on the crude samples showed quick lixidegree of Cs (several per cent) and of 60Co, and 133Ba at a lesser degree. The proposed conditioning technique does not involve a simple coating but true impregnation by a tar-epoxy mixture. The blocks recovered intact from the core by robot services will be placed one by one inside a cylindrical metallic container. Even if this container corrodes and the blocks become fragmented in the future, the normally porous graphite will be unaffected by leaching, since it is proved that all pores larger than 0.1 micron are filled with the tar-epoxy mixture. This is a true long-term waste packaging concept.
Bibliographic Reference: EUR 12815 FR (1990) 78 pp., FS, ECU 7.50
ISBN: ISBN 92-826-1530-8
Record Number: 199011106 / Last updated on: 1994-12-01
Original language: fr
Available languages: fr