Moderator Optimisation in Core Design of Small Modular – Molten Salt Reactor for Military Submarines - Hasibuan, Muhammad Hafiz; Wijanarko, Fakhri; Afiva, Rahmatul; Alfany, Dini; Alwan, Mochammad Ibnu; Meireni, Mutia; Dwijayanto, R Andika Putra; Harto, Andang Widi (MUNISI: Military Mathematics and Natural Sciences, 2024)
The Molten Salt Reactor (MSR) has the potential to be a 4th generation nuclear reactor design due to its high thermal-electric conversion efficiency, integrated safety features, and online reprocessing capacity. The Molten Salt Reactor (MSR) does not require shutdown during operation because the fuel is reprocessed online, making it suitable for submarine propulsion. In this study, beryllium oxide and graphite are used as moderators to increase neutrons in the reactor. Beryllium oxide and graphite are two types of moderators typically employed in nuclear reactors, one of which is in the MSR. Good moderation properties produce many slow neutrons for nuclear fission reactions. The reactor is powered by LiF-U233F3-ThF4 salt in a virtual one-and-a-half liquid configuration. All reactor core design calculations were performed using the MCNP 6.2 code with ENDF/B-VII.0 neutron cross-section. In the context of neutron analysis, the calculated parameters are the effective multiplication factor and the temperature coefficient of reactivity (TCR). The value of is obtained by using the ratio of Th : U, . TCR values were obtained as for graphite moderator and for the beryllium oxide moderator. These initial values indicate that the MSR can reach the critical level with the existing safety characteristics of its features. Using the MSR design, the submarine can achieve a much better range than diesel–powered submarines.