Distribution of fission and activation products in the graphite sleeves of HTGR fuel rods: first and second OGL-1 fuels

Abstract In order to develop the fuel of a high temperature gas-cooled reactor, axial and radial profiles of fission and activation products in the graphite sleeves of the first and second fuel assemblies irradiated in the in-pile gas loop OGL-1 were measured by means of lathe sectioning, gamma spectrometry, and ion-exchange separation. Several sharp axial profile peaks of 90 Sr, 134 Cs, and 137 Cs were observed for the second fuel sleeves. The peaks are assigned to the failed coated fuel particles; the rest of the profiles to the contamination uranium contained in the fuel compact matrix. Fission products with lower diffusivity such as 90 Sr, 106 Ru, and 144 Ce were completely retained in the second fuel sleeve, whereas similar effectiveness in retention was not observed for 134 Cs and 137 Cs. Diffusion coefficients of cesium and strontium have been roughly estimated through the comparison between the measured and the simply calculated profiles in the sleeve.