Behaviour of spent HTR fuel elements in aquatic phases of repository host rock formations
Artikel i vetenskaplig tidskrift, 2006

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.


J Fachinger

M den Exter

B Grambow

Stellan Holgersson

C Landesman

M Titov

T Podruzhina

Nuclear Engineering and Design

0029-5493 (ISSN)

Vol. 236 5-6 543-554


Annan materialteknik


Mer information