Moderator Optimisation in Core Design of Small Modular – Molten Salt Reactor for Military Submarines

Muhammad Hafiz Hasibuan(1*), Fakhri Wijanarko(2), Rahmatul Afiva(3), Dini Alfany(4), Mochammad Ibnu Alwan(5), Mutia Meireni(6), R Andika Putra Dwijayanto(7), Andang Widi Harto(8),

(1) Department of Physics, Faculty of Military Mathematics and Natural Sciences, Republic of Indonesia Defense University, Bogor, Indonesia
(2) Department of Physics, Faculty of Military Mathematics and Natural Sciences, Republic of Indonesia Defense University, Bogor, Indonesia
(3) Department of Physics, Faculty of Military Mathematics and Natural Sciences, Republic of Indonesia Defense University, Bogor, Indonesia
(4) Department of Physics, Faculty of Military Mathematics and Natural Sciences, Republic of Indonesia Defense University, Bogor, Indonesia
(5) Department of Physics, Faculty of Military Mathematics and Natural Sciences, Republic of Indonesia Defense University, Bogor, Indonesia
(6) Department of Physics, Faculty of Military Mathematics and Natural Sciences, Republic of Indonesia Defense University, Bogor, Indonesia
(7) Research Center for Nuclear Reactor Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency, Tangerang Selatan, Indonesia
(8) Department of Nuclear Engineering, Faculty of Engineering, Gadjah Mada University, Yogyakarta, Indonesia
(*) Corresponding Author

Abstract


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.  

Keywords


Molten Salt Reactor; MCNP; Graphite; Beryllium Oxide; Nuclear Submarine

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