Bulk hydrogen retention in graphite at elevated temperature

Abstract

Nuclear graphite and graphite matrix are both high-purity, fine-grain, graphitized materials designed for use in Fluoride-Salt Cooled High Temperature Reactors (FHRs), High Temperature Gas Cooled Reactors (HTGRs), and Molten Salt Reactors (MSRs). Their ability to chemisorb radioisotopes (such as tritium) and their chemical stability is dependent upon the density of reactive carbon sites (RCS), which varies among grades of graphite and changes with neutron irradiation. While nuclear graphites have been extensively characterized, much less data is available for graphite matrix. The current study performs a comparative analysis of IG-110 nuclear graphite and A3 graphite matrix, by X-ray Diffraction (XRD), Raman, and H2 uptake at 700 °C with H2 partial pressure range of 300 Pa and 22 kPa. At 10 kPa, the RCS occupied by H2 uptake per number of carbon atoms at the edge of a crystallite, H/Cedge, is 600(300) appm for IG-110 and 2300(200) appm for A3. Prior empirical studies for nuclear graphite establish a linear relationship between H/C and Cedge/C and degree of graphitization. The present study shows that these correlations no longer hold when extrapolating to graphite matrix; the type of defects and the defect density at the crystallite surface are different between IG-110 and A3 due to differences in raw materials and manufacturing method.