Xiangwen Zhou

Nuclear graphite for high temperature gas-cooled reactors

Abstract 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 oxidation behavior of A3-3 matrix graphite

Abstract 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 =2.9673×10 8 ·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.

Study on the Comprehensive Properties and Microstructures of A3-3 Matrix Graphite Related to the High Temperature Purification Treatment

At the beginning, a comparative analysis was made on the oxidation corrosion rate and ash content of A3-3 matrix graphite (MG) pebbles lathed before and after high temperature purification (HTP) treatment. Their oxidation corrosion rate and ash contents were almost identical, which indicated that the HTP process was to purify the entire MG pebbles and not limited on the surfaces. Furthermore, the multiple mechanical and thermal properties of MG treated without and with the treatment of HTP at ~1900°C were compared and their microstructure features were characterized as well. As the crush strength, oxidation corrosion rate, and erosion rate of MG without HTP treatment did not satisfy the specifications, the comprehensive properties and purity of MG with HTP were improved in various degrees through the HTP process so that all performances met the requirements of the A3-3 MG. The improvement of crush strength and erosion rate of MG in the HTP process could be mainly attributed to the upgradation of ordered microstructure and corresponding increase of density. However, the enhancement of oxidation corrosion rate was due to the synergistic effects of microstructural optimization and reduction of impurity elements, especially the transition metal elements of MG in the HTP process.

Oxidation Behavior of Matrix Graphite and Its Effect on Compressive Strength

Matrix graphite (MG) with incompletely graphitized binder used in high-temperature gas-cooled reactors (HTGRs) is commonly suspected to exhibit lower oxidation resistance in air. In order to reveal the oxidation performance, the oxidation behavior of newly developed A3-3 MG at the temperature range from 500 to 950°C in air was studied and the effect of oxidation on the compressive strength of oxidized MG specimens was characterized. Results show that temperature has a significant influence on the oxidation behavior of MG. The transition temperature between Regimes I and II is ~700°C and the activation energy ( ) in Regime I is around 185 kJ/mol, a little lower than that of nuclear graphite, which indicates MG is more vulnerable to oxidation. Oxidation at 550°C causes more damage to compressive strength of MG than oxidation at 900°C. Comparing with the strength of pristine MG specimens, the rate of compressive strength loss is 77.3% after oxidation at 550°C and only 12.5% for oxidation at 900°C. Microstructure images of SEM and porosity measurement by Mercury Porosimetry indicate that the significant compressive strength loss of MG oxidized at 550°C may be attributed to both the uniform pore formation throughout the bulk and the preferential oxidation of the binder.