Tori Hiatt

Computational Nuclear Physics / Engineering project (exact title TBD)

Started July 2, 2025

Abstract or project description

Molten Salt Reactors (MSR) can have many different salt carriers. Different fluoride salt carriers have been suggested from the beginning of the Oak Ridge National Laboratory experiment 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.