The Release of C-14 of HTR Graphite Spheres during High Temperature Oxidation

Abstract

As the key components of the HTGR fuel, the intact TRISO-coated particles contain almost all fission products under the normal and accident conditions in the reactor. On the other hand, the high temperature oxidation of irradiated fuel pebbles is one of feasible processing methods for HTR spent fuel pebbles. The knowledge of the C-14 release from irradiated fuels and the migration behavior during high temperature oxidation is of great importance for the post-irradiation examinations (PIE) and long-term storage and reprocessing of HTR-PM fuel pebbles. In this work, the C-14 release behavior from one irradiated HTR graphite sphere, which was used as a substitute for a irradiated HTR fuel, was determined by out-of-pile examination. The irradiated HTR fuel was oxidized in the hot lab of INET, and the substantial migration of fission products from graphite matrix was adsorbed and systematically studied by γ-scanning. Specifically, the C-14 release during oxidation was analyzed and illustrated. This work will provide theory and experiment basement for spent fuel disposal of HTGR, as well as a new sample preparation method for C-14 measurement by liquid scintillation counter (LSC).