OpenMC Neutronics Evaluation for Fluoride Salts in Molten Salt Reactors
DOI:
https://doi.org/10.58445/rars.3198Keywords:
Molten Salt Reactor (MSR), Molten Salt Reactor Experiment (MSRE), Fluoride saltsAbstract
Molten Salt Reactors (MSR) can have many different salt carriers. Different fluoride salt carriers have been suggested from the beginning of the Molten Salt Research Experiment (MSRE) at Oak Ridge National Laboratory in 1965. While development and study of different salt carriers has been barred by economical restraints around reactor maintenance and construction, different fuel carriers can be simulated using an open source Monte Carlo code OpenMC. Through differing salt carriers in a pin cell configuration of a Molten Salt Reactor, such as NaF-ZrF4, BeF2, NaF-BeF2, LiF-NaF-BeF2, the neutronics can be studied using the tally methods of OpenMC. More specifically, we will look into k-effective values and flux spectra of the different MSR salts. The outcome of this research paper was that criticality calculations (k-eff values) were different for each pin cell geometry modeled, which suggests an optimal ratio between the size of the moderator and fuel cell exists.
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