3f4a901d-c0f2-49a9-b672-22e82b99ae8a20251119124614221wseas:wseasmdt@crossref.orgMDT DepositEARTH SCIENCES AND HUMAN CONSTRUCTIONS2944-90062944-915410.37394/232024https://wseas.com/journals/eshc/662025662025510.37394/232024.2025.5https://wseas.com/journals/eshc/2025.phpMd Nazirul HudaAnikDepartment of Nuclear Science and Engineering, Military Institute of Science and Technology (MIST) Dhaka, BANGLADESHMd NaibHassanDepartment of Nuclear Science and Engineering, Military Institute of Science and Technology (MIST) Dhaka, BANGLADESHMd RajinRahmanDepartment of Nuclear Science and Engineering, Military Institute of Science and Technology (MIST) Dhaka, BANGLADESHA. S.MollahDepartment of Nuclear Science and Engineering, Military Institute of Science and Technology (MIST) Dhaka, BANGLADESHThe 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%UF4.111920251119202537454https://wseas.com/journals/eshc/2025/a08eshc-003(2025).pdf10.37394/232024.2025.5.4https://wseas.com/journals/eshc/2025/a08eshc-003(2025).pdfRenault, C, Hron, M, Konings, R, and Holcomb, D E. (2009). "The Molten Salt Reactor (MSR) in generation 4: overview and perspectives." NEA. 10.1016/j.pnucene.2014.02.014Jérôme Serp, Michel Allibert, Ondřej Beneš, Sylvie Delpech, Olga Feynberg, Véronique Ghetta, Daniel Heuer, David Holcomb, Victor Ignatiev, Jan Leen Kloosterman, Lelio Luzzi, Elsa Merle-Lucotte, Jan Uhlíř, Ritsuo Yoshioka, Dai Zhimin, (2014). The molten salt reactor (MSR) in generation IV: Overview and perspectives, Progress in Nuclear Energy, Volume 77, Pages 308-319, ISSN 0149-1970, https://doi.org/10.1016/j.pnucene.2014.02.014. 10.1016/j.anucene.2013.09.004Ignatiev, V., Feynberg, O., Gnidoi, I., Merzlyakov, A., Smirnov, V., Surenkov, A., Tretiakov, I., Zakirov, R., Afonichkin, V., Bovet, A. and Subbotin, V., (2007). Progress in development of Li, Be, Na/F molten salt actinide recycler & transmuter concept. In Proceedings of ICAPP (Vol. 7, pp. 13-18). 10.1016/j.nucengdes.2009.12.033LeBlanc, David. (2010). Molten Salt Reactors: A New Beginning for an Old Idea. Nuclear Engineering and Design. 240. 1644-1656. 10.1016/j.nucengdes.2009.12.033 Ondřej Chvála., (2014). Proceedings of ICAPP 2014 Charlotte, USA, Paper 14187 10.1007/s10512-012-9537-2Ignatiev, V.V., Feynberg, O.S., Zagnitko, A.V., Merzlyakov, A.V., Surenkov, A.I., Panov, A.V., Subbotin, V.G., Afonichkin, V.K., Khokhlov, V.A. and Kormilitsyn, M.V. (2012). Molten-salt reactors: new possibilities, problems and solutions. Atomic energy, 112, pp.157-165. 10.2172/4654707R. C. ROBERTSON, “MSRE Design and Operations Report Part I: Description of Reactor Design,” ORNL-TM0728,” Oak Ridge National Laboratory (1965). 10.1080/00295639.2021.1880850Dan Shen,a Germina Ilas,b Jeffrey J. Powers,b and Massimiliano Fratoni, Reactor Physics Benchmark of the First Criticality in the Molten Salt Reactor Experiment, Nuclear Science and Engineering, Vol. 195, 825-837, 2021, DOI: https://doi.org/10.1080/00295639.2021.1880850 Thoma, R. E. (Ed.). (1959). Phase diagrams of nuclear reactor materials (Vol. 2548). Oak Ridge National Laboratory. 10.2172/4622532Thoma, R. E. (1968). CHEMICAL FEASIBILITY OF FUELING MOLTEN SALT REACTORS WITH PuF (No. ORNL-TM-2256). Oak Ridge National Lab., Tenn. 10.2172/4492893Cantor, S. (1968). PHYSICAL PROPERTIES OF MOLTEN-SALT REACTOR FUEL, COOLANT, AND FLUSH SALTS (No. ORNL-TM-2316). Oak Ridge National Lab., Tenn. 10.2172/4521233Briggs, R.B., (1966). Molten-Salt Reactor Program, Semiannual Progress Report For Period Ending February 28, 1966, ORNL-3936, ORNL. 10.2172/1036564Williams, D. F., Toth, L. M., and UT-Battelle, L. L. C. (2005). Chemical considerations for the selection of the coolant for the Advanced High-Temperature Reactor. ORNL/GEN4/LTR-05-011, Oak Ridge National Laboratory, Oak Ridge, TN.