Behaviour of spent HTR fuel elements in aquatic phases of repository host rock formations
Journal article, 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.