Since its first successful use in the CP-1 nuclear reactor in 1942, nuclear graphite has played an important role in nuclear reactors especially the high temperature gas-cooled type (HTGRs) owing to its outstanding comprehensive nuclear properties. As the most promising candidate for generation IV reactors, HTGRs have two main designs, the pebble bed reactor and the prismatic reactor. In both designs, the graphite acts as the moderator, fuel matrix, and a major core structural component. However, the mechanical and thermal properties of graphite are generally reduced by the high fluences of neutron irradiation of during reactor operation, making graphite more susceptible to failure after a significant neutron dose. Since the starting raw materials such as the cokes and the subsequent forming method play a critical role in determining the structure and corresponding properties and performance of graphite under irradiation, the judicious selection of high-purity raw materials, forming method, graphitization temperature and any halogen purification are required to obtain the desired properties such as the purity and isotropy. The microstructural and corresponding dimensional changes under irradiation are the underlying mechanism for the changes of most thermal and mechanical properties of graphite, and irradiation temperature and neutron fluence play key roles in determining the microstructural and property changes of the graphite. In this paper, the basic requirements of nuclear graphite as a moderator for HTGRs and its manufacturing process are presented. In addition, changes in the mechanical and thermal properties of graphite at different temperatures and under different neutron fluences are elaborated. Furthermore, the current status of nuclear graphite development in China and abroad is discussed, and long-term problems regarding nuclear graphite such as the sustainable and stable supply of cokes as well as the recycling of used material are discussed. This paper is intended to act as a reference for graphite providers who are interested in developing nuclear graphite for potential applications in future commercial Chinese HTGRs.

The effects of temperature on the oxidation behavior of the A3-3 matrix graphite (MG) in the temperature range 798-973 K in air with a flow rate of 100 ml/min to burn-offs of 10-15 wt%, were investigated by a home-made thermo-gravimetric experimental setup. The oxidation rate (OR) increases significantly with the temperature. The OR at 973 K is over 70 times faster than at 798 K. The oxidation kinetics of A3-3 MG in air at temperatures up to 973 K is in the reaction control regime, where the activation energy is 176 kJ/mol and the Arrhenius equation could be described as: OR=2.9673×108·exp(-21124.8/T) wt%/min. The relatively lower activation energy of MG than that of structural nuclear graphite indicates that MG is more easily oxidized.

The coated particles were first invented by Roy Huddle in Harwell 1957. Through five decades of development, the German UO2 coated particle and US LEU UCO coated particle represent the highly successful coated particle designs up to now. In this paper, current status as well as the failure mechanisms of coated particle so far is reviewed and discussed. The challenges associated with high temperatures for coated particles applied in future VHTR are evaluated. And future development prospects of advanced coated particle suited for higher temperatures are presented. According to the past coated fuel particle development experience, it is unwise to make multiple simultaneous changes in the coated particle design. Two advanced designs which are modifications of standard German UO2 coated particle (UO2∗ herein) and US UCO coated particle (TRIZO) are promising and feasible under the world-wide cooperations and efforts.