Characterization of Tritium Transport in the Flibe-Graphite System, for In-Situ Tritium Absorption by the Fuel Elements of the Fluoride-Salt-Cooled High-Temperature Reactor (FHR)

摘要

Tritium control in the Fluoride-Salt-Cooled High Temperature Reactor (FHR) is an area of great interest because tritium is produced in the primary coolant (FLiBe) by neutron irradiation, and readily diffuses through metallic heat exchanger tubes at FHR operating temperatures. This paper provides an update on the tritium transport research at the University of Wisconsin - Madison in the FHR salt-fuel system. It is expected that the matrix-graphite fuel elements in the FHR will provide a significant removable sink for tritium control. Operational data from the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL) indicated that the graphite had a strong affinity for the tritium produced in FLiBe. Experimental data for tritium retention in graphite is available from the MSRE and from research with application to fusion systems. Data from the MSRE will be integral to benchmarking tritium retention expectation. Other experiments, which expose graphite to fixed pressures of tritium gas may not properly characterize the salt-graphite system because it may not account for conditions in the FHR primary coolant. This paper focuses on the mechanisms for transport and retention in matrix graphite fuel within the FLiBe-matrix graphite system. The differences affecting tritium retention and kinetics between matrix graphite and nuclear graphite are proposed and discussed. In future work, benchmark models for tritium retention in graphite and experimental techniques to image tritium within graphite will be investigated.