A METHOD FOR THE STUDY AND CORRELATION OF FISSION-GAS-RELEASE BEHAVIOR OF FUEL MATERIALS DURING IRRADIATION

摘要

Carbon-coated and silicon carbide-coated 1.5-in.-diameter graphite spheres fueled with natural UC and UC2 particles were irradiated at 130 to 1900 F at relatively low flux [<5 x 10 n/(cm )(sec)] in an in-pile study of fission-gas release. A stream of helium flowed continuously over the spheres and was sampled periodically for fission gases. Five species, krypton-85m, krypton-87, krypton-88, xenon-133, and xenon-135, were measured.nMoisture or air in the sweep gas apparently oxidized the carbon coating and exposed the fueled matrix of the carbon-coated sphere. The fission-gas release from this sphere, was high at all temperatures and increased tenfold over the range from 1000 to 1900 F. Diffusion of fission recoils from the graphite crystallites is believed to explain the temperature coefficient of release.nThe siliconized silicon carbide-coated sphere showed excellent retention of fission gas, exhibiting a release that was 10 -4 that of the exposed matrix of the carbon-coated sphere at 1900 F with little or no effect of temperature. Surface uranium contamination in the coating is believed to explain the release and absence of a temperature effect.nGeneralized expressions were developed to describe the diffusion of fission products under a variety of conditions. Clad-element and equivalent-sphere concepts were utilized.