Distribution of fission products in irradiated graphite materials of HTGR fuel assemblies: Third and fourth OGL-1 fuels

Abstract Axial, circumferential and radial distributions of fission products in the graphite sleeve, inner-tube and block of irradiated high temperature gas-cooled reactor (HTGR) fuel assemblies were measured by gamma spectrometry, lathe sectioning and beta counting with ion-exchange separation. Some distinctive peaks of 144 Ce and 125 Sb in their axial profiles, together with the very high activity level of fission products are ascribed to the failure of coated fuel particles. The effective retention capability of the graphite sleeve was observed for 90 Sr, 106 Ru, 125 Sb, 144 Ce and 155 Eu; whereas not for 134 Cs and 137 Cs. Silver-110m was detected in graphite materials of the fourth OGL-1 fuel assembly with an increased burnup of 1.96% fissions per initial metal atom (FIMA). Effective in-pile diffusion coefficients of 90 Sr, 125 Sb and 144 Ce in the graphite sleeves have been estimated using the Fickian diffusion theory.