5 January 2018

UO2 fuel behavior at very high burnup

Miscellaneous Energy and Environment Materials Research Nuclear Power Plant Safety


The investigation of the nuclear fuel at very high burnup is critical for evaluating the safety margin for the evaluated fuel in normal as well as in accidental conditions. PSI is one of the very few hot laboratories which possesses access to irradiated UO2 fuel with very high burnup from commercial reactors. The application of relevant tools for the investigation, handling and analysis of those highly irradiated materials emphasize the necessary expertise. The shielded instruments of the laboratory have been used for detailed analyses of the structure and composition of different fuel pellets irradiated in commercial reactors with burnups over 70 MWd/kgHM and up to 105 MWd/kgHM. The results show a high fission gas release up to 42%. Detailed investigations of the radial distribution of xenon and the distribution of fission gas between bubbles and the fuel matrix show that the release from the fuel at intermediate radial positions was mainly responsible for the high fission gas release. In the case of this investigation, even when it is accepted that only part of the gas released from the fuel matrix is contained in the gas pores (the rest being released to the rod free volume) then the gas pressure in the pores is up to 70 times higher than the equilibrium pressure. It can be concluded that gas release from high burnup fuel will increase considerable during an accident if thermal release is induced in an area of the fuel containing the high burnup structure. Furthermore the fission gas release from the high burnup structure, below the threshold temperature for thermal release, has to be considered at very high local burnup.
Contact
Matthias Martin
Head of ARM
Paul Scherrer Institut
OHLA/134
Telephone: +41 56 310 41 79
E-mail: Matthias.Martin@psi.ch
Original Publication
*Journal of Nuclear Materials *
(R. Restani, M. Horvath, W. Goll, J. Bertsch, D. Gavillet, A. Hermann, M. Martin, C.T. Walker Journal of Nuclear Materials 481 (2016) 88-100)