Dimensional and physical property changes of high temperature-reactor graphites, pyrocarbons and fuel bodies irradiated at temperatures in the range 600–1200°C

Abstract

The 20 MW(th) experimental helium cooled reactor “Dragon” designed and built by the OECD High Temperature Reactor Project at Winfrith, England uses an all graphite core with coated particle fuel embedded in a resin bonded graphite powder matrix. This paper describes the irradiation behaviour of the carbonaceous core components. Measurements have been carried out of the dimensional and physical property changes of extruded and pressed Gilsocarbon graphite, irradiated to fast neutron doses of up to 5 × 10 21 n cm −2 Ni DIDO equivalent, and the results are compared with two earlier nuclear graphites. Differences in the shrinkage rates of the various Gilsocarbon graphites are related to their initial Young's Modulus and strength. The results indicate that for a given basic type of graphite (gilsonite) a higher strength material will shrink faster than a comparable low strength material. Comparison of the volume changes of the graphites tested suggests that the expansion rate after the shrinkage maxima is more rapid for graphites irradiated at 900°C than at 1200°C. This rapid volume expansion at 900°C is accompanied by a loss of strength already shown by the fall in Young's Modulus of an earlier fine grain nuclear graphite used in the Dragon Reactor. A range of pyrolytic carbon specimens having a range of densities was irradiated up to fast neutron doses of 3 × 10 21 n cm −2 Ni DIDO equivalent. The results show that the dimensional changes of these specimens depend primarily on their initial density. Measurements of the outer diameter and length changes of coated particle matrix type fuel compacts irradiated at temperatures between 900°C–1200°C, to a maximum fast neutron dose of 1.7 × 10 21 n cm −2Ni DIDO equivalent, reveal a relatively isotropic behaviour of these composite bodies. However they shrink more rapidly than most nuclear graphites. The influence of materials selection on some reactor design problems is briefly discussed.