Development of synthetic graphite for reactor fuel, moderators, and nuclear reactor components: A literature review

Abstract

Graphite has been used as a moderator material in nuclear power stations and is considered a potential material for use in future Generation IV advanced reactors. Graphite has excellent thermal-mechanical and neutron moderation properties, and therefore, it is currently used as a structural core component and nuclear fuel in High Temperature Gas Cooled Reactors (HTGR). Synthetic graphite with higher specific capacity and superior cyclic life is prepared by high-temperature graphitization of anthracite coupled with effective catalysts. In the graphitization process, thermodynamically unstable carbon atoms regularly change from a disordered structure to a graphite crystal structure through thermal activation. However, the graphitization process needs to be carried out at a high temperature above 2800 °C. Common raw materials include pyrolysis coal tar and carbonized and calcined carbon materials. Literature studies show that the selection of coal as the raw material to be synthesized affects the final properties of graphite, such as purity, homogeneity and microscopic structure. This comparison of raw materials provides insight into which materials are most efficient and economical to use in the production of synthetic graphite. A spectrum fitting methodology is developed. Next, SEM observations of the irradiated materials are performed. to compare the reactor performance by evaluating the critical numbers for systems using a two-phase composite moderator (magnesia matrix with a moderated phase containing beryllium or retained hydrides) relative to the case of a reference graphite moderator. Depending on their properties, different nuclear grades of graphite are used for HTGR. Given their excellent mechanical properties, IG-110 and PGX graphite grades are used for core structural components such as fuel blocks and permanent reflector blocks, while A3-3 heavy fraction graphite/resin mixtures are used as matrix components in TRISO fuel compaction due to their high absorption of fission products such as Cs and Sr. The accumulation of strong neutron absorbers in graphite lowers the cavity reactivity coefficient, but it remains positive. Despite the loss of excess reactivity in the core, the presence of strong neutron absorbers in graphite is beneficial from a safety point of view