Modeling the Behavior of Tritium in a Molten Salt Nuclear Reactor

A liquid fuel reactor based on lithium and beryllium fluoride salts produces a significant amount of tritium as a result of nuclear reactions of the main constituents of the fuel salt with neutrons. At high operating temperatures (600–750°C), tritium can diffuse through the metal walls of the reactor and pipelines, thus affecting the environmental situation. In this article, a method for calculating tritium distribution in this type of reactor is developed and verified. The calculation method is based on differential equations of tritium mass balance in the system elements taking into account the change in the ratio of chemical forms of tritium due to the redox potential of the salt. The calculation method is verified on the experimental data obtained in the 10 MW(t) Li,Be,Zr,U/F Molten Salt Reactor Experiment (MSRE) carried out at the Oak Ridge National Laboratory (ORNL, United States).