Decontamination of Nuclear Graphite

Graphite is used in gas-cooled reactors as moderator and reflector. Fission products and activation of impurities of the graphite contaminate this graphite during reactor operation. Most radioisotopes can principally be removed by using the purification methods already applied during manufacture of nuclear graphite. But due to the same chemical behaviour as ¹²C, this seems not to be applicable to ¹⁴C, which is a key nuclide for waste management. Contaminated graphite can not be stored in low level surface disposal facilities such as e.g. Centre de L'Aube, in France, due to the long half live of ¹⁴C [1]. Furthermore, the ¹⁴C activity of the graphite reflectors from the two German HTR reactors (AVR and THTR) only would already cover more than 90% of the ¹⁴C activity licensed for the underground disposal site Konrad in Germany for non heat generating radioactive waste [2]. Therefore, alternative waste management strategies should be developed for nuclear graphite. The total burning of nuclear graphite requires a separation of the ¹⁴C from the off-gas. However, this carbon isotope has the same chemical properties as the ¹²C from the graphite matrix and the solidification of the whole amount of CO₂ would negate the volume reduction benefit of burning. Physical separation methods like cryogenic distillation or ultra centrifugation are not economic for the large amounts of CO₂. Thus, a process is requested which takes benefit from the inhomogeneous distribution of the ¹⁴C in the graphite matrix leading to ¹⁴C enriched and depleted offgas streams [3]. Pyrolysis in an inert atmosphere with a low amount of oxygen impurities or oxidation in a steam atmosphere seem to be the most promising process options. First experiments had been performed with nuclear graphite from the thermal column of the research reactor FRJ-1 and from the graphite core structure of the high temperature reactor AVR. ¹⁴C could be removed from the graphite with high selectivity under an inert atmosphere. An even higher ¹⁴C decontamination factor could be obtained in a steam atmosphere. Next step will be a combination of these processes to optimise the ¹⁴C release rates combined with high selectivity allowing for an industrial process development. Furthermore, an off gas treatment process for the ¹⁴C enriched fraction has to be develop to solidify the ¹⁴C for disposal reflecting the geochemical conditions of repositories. An advanced optional off gas treatment could also be a re-conversion into products for reuse in nuclear industry. (author)