Decontamination of nuclear graphite

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 C-14 and Cl-36 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 C-12, this does not seem to be applicable to C-14.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 C-14 [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 C-14 activity of the graphite reflectors from the two German HTR reactors (AVR and THTR) would amount to more than 90% of the total C-14 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 C-14 were emitted into the atmosphere in the form of CO2. The required separation of the C-14 from the off-gas is difficult and not economic because this carbon isotope has the same chemical properties as the C-12 from the graphite matrix. The solidification of the whole amount of CO2 would cancel out the volume reduction advantage of burning.Thus, a process is required which benefits from the inhomogeneous distribution of the C-14 in the graphite matrix leading to C-14-enriched and C-14-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 Cl-36 and I-129, can be removed from graphite by thermal treatment. Even significant parts of the C-14 inventory can be selectively extracted because most of the C-14 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

Behaviour of spent HTR fuel elements in aquatic phases of repository host rock formations

One back-end option for spent HTR fuel elements proposed for future HTR fuel cycles in the EC is an open fuel cycle with direct disposal of conditioned or non-conditioned fuel elements. This option has already been chosen in Germany due to the political decision to terminate the use of HTR technology. First integral leaching investigations at Research Centre Juelich on the behaviour of spent HTR fuel in salt brines, typical of accident scenarios in a repository in salt, proved that the main part of the radionuclide inventory cannot be mobilised as long as the coated particles do not fail. However, such experiments will not lead to a useful model for performance assessment calculations, because a failure of the coatings by corrosion will not occur during experimental times of a few years. In order to get a robust and realistic model for the long-term behaviour in aqueous phases of host rock systems, it is necessary to understand the barrier function of the different parts of an HTR fuel element, i.e. the matrix graphite, the different coating materials, and the fuel kernel.Therefore, our attention is focused on understanding and modelling the barrier performance of the different parts of an HTR fuel element with respect to their barrier function, and on the development of an overall model for performance assessment. In order to understand this behaviour, it is necessary to start with investigations of unirradiated material, and to proceed with experiments with external gamma irradiation to determine the effects of oxidising radiolysis species. Further experiments with irradiated material have to be performed to investigate the influence of the irradiation damage, and finally an investigation has to be made of the irradiated material plus additional gamma irradiation. Experimental data are now available for the diffusive transport of radionuclides in the water-saturated graphite pore system, the corrosion rates of unirradiated graphite with and without external gamma irradiation and unirradiated and irradiated silicon carbide, and for the dissolution rates of UO2 and (Th,U)O-2 fuel kernels with and without external gamma irradiation. All investigations were performed in aquatic phases from salt, granite, and clay host rock. (c) 2006 Elsevier B.V. All rights reserved

Behaviour of Spent HTR Fuel Elements in Aquatic Phases of Repository Host Rock Formations

One back-end option for spent HTR fuel elements proposed for future HTR fuel cycles in the EC is an open fuel cycle with direct disposal of conditioned or non-conditioned fuel elements. This option has already been chosen in Germany due to the political decision to terminate the use of HTR technology. First integral leaching investigations at Research Centre Juelich on the behaviour of spent HTR fuel in salt brines, typical of accident scenarios in a repository in salt, proved that the main part of the radionuclide inventory cannot be mobilised as long as the coated particles do not fail. However, such experiments will not lead to a useful model for performance assessment calculations, because a failure of the coatings by corrosion will not occur during experimental times of a few years. In order to get a robust and realistic model for the long-term behaviour in aqueous phases of host rock systems, it is necessary to understand the barrier function of the different parts of an HTR fuel element, i.e. the matrix graphite, the different coating materials, and the fuel kernel. Therefore, our attention is focused on understanding and modelling the barrier performance of the different parts of an HTR fuel element with respect to their barrier function, and on the development of an overall model for performance assessment. In order to understand this behaviour, it is necessary to start with investigations of unirradiated material, and to proceed with experiments with external gamma irradiation to determine the effects of oxidising radiolysis species. Further experiments with irradiated material have to be performed to investigate the influence of the irradiation damage, and finally an investigation has to be made of the irradiated material plus additional gamma irradiation. Experimental data are now available for the diffusive transport of radionuclides in the water-saturated graphite pore system, the corrosion rates of unirradiated graphite with and without external gamma irradiation and unirradiated and irradiated silicon carbide, and for the dissolution rates of UO2 and (Th,U)O2 fuel kernels with and without external gamma irradiation. All investigations were performed in aquatic phases from salt, granite, and clay host rock

European Programme on High Temperature Reactor Nuclear Physics, Waste and Fuel Cycle Studies

This project represents a collaborative R&D programme in the field of HTR-related nuclear physics, waste and fuel cycle studies (HTR-N) and is funded by the European Commission in the HTR-N and HTR-N1 contracts of the 5th Framework Programme (FP5). The paper illustrates the nuclear physics analyses of first criticality of the Japanese High-Temperature Test Reactor (HTTR) and of the Chinese 10 MWth High-Temperature Reactor (HTR-10). Reasons for discrepancies in the predictions of critical core configuration of HTTR are explained. Further activities deal with different fuel cycles aimed at effective burning of plutonium and minor actinides. First studies, for first generation plutonium (Pu-I) from reprocessed LWR uranium fuel, show that it is possible to reduce the fissile Pu content by up to 90%. Work continues on a symbiotic fuel cycle for LWR and HTR by burning of 2nd generation plutonium (Pu-II) from reprocessed spent LWR MOX fuel in an HTR core still maintaining HTR safety features with regard to nuclear stability and self-acting decay heat removal. Both, block-type and pebble fuel are analyzed. Studies on burnable poison and on uncertainties of nuclear data are included into the HTR-N programme. R&D on waste minimization options and experiments and first results on long-term behavior of spent HTR-fuel is reported, too