Study of Tritium Transport in Nuclear-Grade Graphite and Molten Fluoride Salt Systems
High-temperature salts are proposed as coolant in recent reactor technologies including Fluoride Salt-Cooled High Temperature Reactor (FHR) with fluoride salt as primary and secondary loop coolant, Molten Salt Reactors (MSRs) with fuel salt, and high-magnetic-field fusion machines with salt coolant blanket. The baseline coolant is {sup 7}Li{sub 2}BeF{sub 4} (flibe) due to its good thermal-hydraulic and neutronic properties. Tritium produced in flibe due to neutron irradiation makes the issue of tritium transport an area of great interest. In the temperature range of 427- 704 deg. C, tritium readily diffuses through metallic reactor structural materials, which makes tritium control and capture a key technical challenge. It is anticipated that tritium can be captured by graphite, as graphite material showed a high affinity for tritium in the Molten Salt Reactor Experiment (MSRE) due to the porous and irregular crystallite microstructure. Two types of nuclear grade graphite which used as core components for FHR: matrix graphite, which encapsulates the TRISO (Tri-structural-isotropic) fuel particles in FHR, and high purity, isotropic nuclear grade graphite IG-110 which can be used as moderator and reflector in reactor core for both fission and fusion reactor. The microstructure of matrix graphite and IG-110 is different from that of matrix graphite, due to the raw material and manufacturing processes. There are three phenomena important for predicting tritium transport in graphite-salt systems: tritium species diffusion in molten flibe; tritium species transport at the salt-graphite interface; and tritium species transport in graphite. In this summary microstructure characterization of IG-110 and matrix graphite, characterization of graphite-salt interface as well as techniques for trace tritium detection in graphite are discussed.