An integrated model of tritium transport and corrosion in Fluoride Salt-Cooled High-Temperature Reactors (FHRs) – Part I: Theory and benchmarking

Abstract

The Fluoride Salt-Cooled High-Temperature Reactor (FHR) is a pebble bed nuclear reactor concept cooled by a liquid fluoride salt known as “flibe” (7LiF-BeF2). A model of TRITium Diffusion EvolutioN and Transport (TRIDENT) was developed for use with FHRs and benchmarked with experimental data. TRIDENT is the first model to integrate the effects of tritium production in the salt via neutron transmutation, with the effects of the chemical redox potential, tritium mass transfer, tritium diffusion through pipe walls, tritium uptake by graphite, selective chromium attack by tritium fluoride, and corrosion product mass transfer. While data from a forced-convection polythermal loop of molten salt containing tritium did not exist for comparison, TRIDENT calculations were compared to data from static salt diffusion tests in flibe and flinak (0.465LiF-0.115NaF-0.42KF) salts. In each case, TRIDENT matched the transient and steady-state behavior of these tritium diffusion experiments. The corrosion model in TRIDENT was compared against the natural convection flow-loop experiments at the Oak Ridge National Laboratory (ORNL) from the 1960s and early 1970s which used Molten Salt Reactor Experiment (MSRE) fuel-salt containing UF4. Despite the lack of data required by TRIDENT for modeling the loops, some reasonable results were obtained. The TRIDENT corrosion rates follow the experimentally observed dependence on the square root of the product of the chromium solid-state diffusion coefficient with time. Additionally the TRIDENT model predicts mass transfer of corrosion products from the hot to the cold leg (as was observed in the experiments with salts containing UF4). In a separate paper the results of TRIDENT simulations in a prototypical FHR are presented.