Oxidative Resistance of Irradiated Graphite Coated with Silicon Carbide

Abstract

Oxidation-induced damage in nuclear reactor materials is one of the important safety issues in accidents. In particular, graphite materials are used for the internal components of a high temperature gas-cooled reactor (HTGR) due to its excellent neutron, thermal and mechanical properties. Graphite is used as a structural material for fuel elements, as well as neutron moderator and reflector. In the case of an accident with ingress of air or water, one of the most serious problems would be the oxidation of graphite material. To eliminate this, it is proposed to use the graphite having an oxidation-resistant protective coating based on silicon carbide (SiC). Studies of the impact of high temperature and neutron irradiation to oxidation resistance properties of the graphite coated with SiC were carried out at INP (Kazakhstan) jointly with JAEA (Japan). The specimens from four Japanese manufacturers were irradiated in the WWR-K reactor for 200 EFPDs. The maximum fast neutron fluency (En> 0.18 MeV) comprised 1.1·1025 m-2. After irradiation, the specimens were tried for oxidation in the environment of helium and 20% of oxygen at 1200°C. The release of carbon dioxide (CO2) was monitored during the test. Only one specimen, from manufacturer “X”, was found to to be resistant to oxidation – no CO2 emission was not detected. Three other specimens were oxidized. The work is devoted to this test.