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Conference Article: Behaviour of spent HTR fuel elements in aquatic phases of repository host rock formations
J. Fachinger, S. Holgerson, M. Titov, T. Podruhzina, Research Centre Jülich, Jülich, GERMANY; M. den Exter, Nuclear Research and Consultancy Group-NRG, Petten, THE NETHERLANDS; B. Grambow; C. Landesmann, Ecole de Mines de Nantes, Nantes Cedex, FRANCEAbstract
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 the Research Centre Juelich on the behaviour of spent HTR fuel in salt brines, typical for 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 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 a HTR fuel element, which are the matrix graphite, the different coating materials and the fuel kernel. Therefore our is focused on the understanding and modelling of the barrier performance of the different parts of a HTR fuel element with respect to their barrier function, and on the development of an overall model for performance assessment. In order to understand the 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 damages, and finally the investigation of 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 without and with external gamma-irradiation and unirradiated and irradiated silicon carbide, and for the dissolution rates of UO2 and (Th,U)O2 fuel kernels without and with external gamma-irradiation. All investigations have been performed in aquatic phases from salt, granite and clay host rock.
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key words: Gas Cooled Reactor, Nuclear Technology
- Reference:
- Proceedings of the Conference on High Temperature Reactors, Beijing, China, September, 22-24, 2004
- International Atomic Energy Agency, Vienna (Austria)
- HTR-2004, pp:1-22
- International Atomic Energy Agency, Vienna (Austria)