Graphite is used in gas-cooled reactors (e.g. AGR, MAGNOX, HTR) and Russian RMBK reactors as a moderafor and reflector. About 250,000 Mg of irradiated graphite (i-graphite) has to be considered as radioactive waste in the next few centuries. Fission products and activation of impurities in the graphite contaminate this graphite during reactor operation. Key nuclides for waste management are Co-60 during decommissioning, if decommissioning is performed immediately after reactor shutdown, and the long living radionuclides (14)C and (36)Cl for long-term safety in the case of direct disposal. Most radioisotopes can, in principle, be removed by using the purification methods already applied during the manufacture of nuclear graphite. However, due to the same chemical behaviour as (12)C, this does not seem to be applicable to (14)C. Contaminated graphite cannot be stored in low-level surface disposal facilities such as Centre de L'Aube, in France, due to the long half-life of (14)C [Millington, D.N.. Sneyers, A., Mouliney, M.H., Abram, T., Brucher, H., 2006. Report on the international Regulation as regards HTR/VHTR Waste Management, Deliverable D-BF1.1 of the Raphael Project, EC Contract 516508, Confidential report]. Furthermore, the (14)C activity of the graphite reflectors from the two German HTR reactors (AVR and THTR) would amount to more than 90% of the total (14)C activity licensed for the underground disposal site Konrad in Germany for non-heat-generating radioactive waste [Brennecke, P., October 1993. Anforderungen an endzulagernde radioaktive Abfalle (Vorlaufige Endlagerbedingungen, Stand: April 1990 in der Fassung vom Oktober 1993) - Schachtanlage Konrad -, BfS-ET-3/90-REV-2, Salzgitter, p. 51]. The burning of nuclear graphite would be an efficient method for volume reduction, but would not be accepted by the public as long as all the (14)C were emitted into the atmosphere in the form of CO(2). The required separation of the (14)C from the off-gas is difficult and not economic because this carbon isotope has the same chemical properties as the (12)C from the graphite matrix. The solidification of the whole amount of CO(2) would cancel out the volume reduction advantage of burning. Thus, a process is required which benefits from the inhomogeneous distribution of the (14)C in the graphite matrix leading to (14)C-enriched and (14)C-depleted off gas streams (Schmidt, P.C., 1979. Alternativen zur Verminderung der C-14-Emission bei der Wiederaufarbeitung von HTR-Brennelementen, JUL-1567]. Tritium and other radioisotopes, including (36)Cl and (129)I, can be removed from graphite by thermal treatment. Even significant parts of the (14)C inventory can be selectively extracted because most of the (14)C may be adsorbed on the surface of the crystallites in the pore structure and not integrated into the crystal lattice. This has recently been demonstrated in principle by the HTR-N/N1 project. As an accompaniment to thermal treatments, steam reforming is an alternative method for decontaminating graphite from radionuclides. The decontamination rates are even higher in comparison to pure thermal treatment in an inert atmosphere, as was first evidenced by basic experiments in the HTR-N/N1 project. (C) 2008 Johannes Fachinger. Published by Elsevier B.V. All rights reserved.
