EARTH SCIENCES AND HUMAN CONSTRUCTIONS
Print ISSN: 2944-9154, E-ISSN: 2944-9006 An Open Access International Journal of Earth Sciences and Human Constructions
Volume 5, 2025
Study on Some Neutronics Behavior of Low Enriched Uranium Salt Composition Proposed for a Molten Salt Reactor (MSR) Using the OpenMC Monte Carlo Method
Authors: , , ,
Abstract: The molten salt reactor (MSR) is a type of GEN-IV advanced reactor that uses melt combinations of heavy metal elements and molten salt as fuel and coolant. Molten salt reactors (MSRs) are fourth-generation reactors that are built to be safe, have no risk of core meltdown, and can be fed and processed online. This study examines the neutronics properties of a conventional MSR using Monte-Carlo and OpenMC codes. MSR cores with varying low-enriched 235U coolant salt compositions were tested to determine the optimal fuel salt composition. To assess non-proliferation, neutronics and safety were tested on low-enriched uranium fuel with different coolant salt compositions for MSR. OpenMC was used to create and simulate eight reactor cores with various fuel compositions. These computations were completed in 35 cycles, with 3000 particles per cycle, while 5 cycles were skipped to avoid statistical errors. For fission rates, temperature reactivity feedback, conversion ratio, and neutron spectra calculations, statistical errors were reduced to 1.9% and 63 pcm for keff values, respectively. All computations are performed using the nuclear data libraries JEFF3.3 and END/B VIII.0. The burnup of fissionable materials and neutron toxicity were investigated. The fission rates of U-235, Pu-239, Xe-135 and Pu-241 were investigated in relation to burnup. The neutronic evaluation of standard fuel salt composition for the ORNL molten salt reactor was performed using OpenMC during normal operation and compared to the experimental value in terms of effective multiplication factor for validation (keff), which was 0.06%. MSRs are passively safe because of the negative temperature reactivity coefficient of fuel salt. Because of their increased atom density, conversion ratio, and FoM, cooling salts like 73%LiF-27%UF4 may be suitable carrier salts. This study outlines the problems, challenges, and development trends for MSR multi-physics models in order to guide future research. This work serves as a reference for molten salt reactor core design using an ideal fuel salt composition of 73% LiF-27%UF4.
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Pages: 37-45
DOI: 10.37394/232024.2025.5.4