Open AccessA Comparative Study of Th-U233 and Th-Pu Molten-Salt Reactor (MSR) with 2-region core
Author(s) -
Melati Ifthacharo,
Sidik Permana,
Geby Saputra
Publication year - 2020
Publication title -
journal of physics. conference series
Language(s) - English
Resource type - Journals
SCImago Journal Rank - 0.21
H-Index - 85
eISSN - 1742-6596
pISSN - 1742-6588
DOI - 10.1088/1742-6596/1493/1/012015
Subject(s) - fissile material , nuclear engineering , nuclear reactor core , uranium , plutonium , core (optical fiber) , enriched uranium , molten salt reactor , thorium fuel cycle , nuclear data , neutron flux , neutron , molten salt , mox fuel , nuclear physics , materials science , physics , engineering , metallurgy , composite material
Sustainable closed fuel cycle and inherent safety are one of the key aspects of Generation IV reactor design. Hence Molten salt reactor (MSR) is chosen as one of the candidates of Generation IV reactor concepts. FUJI-U2 or MSR-2R is a 2-region core MSR design that was proposed as a possible simplification of FUJI-U3 (3-region core). Taking into consideration that uranium −233 doesn’t exist naturally, FUJI-U2 utilize both plutonium and uranium −233 as a fissile material. To understand the characteristics further, a comparative study of 2-region core MSR with Th-U 233 and Th-Pu as fuel has been carried out. Nuclear analysis code SRAC2006 with the nuclear data library of JENDL4.0 was used, and CITATION was employed to validate the core. The design parameters set for each region are fuel composition, core radius, and fuel fraction. The simulation burn up calculation is set to 2000 days in total, with 21 step each for 100 days cycle. Neutron multiplication factor, fuel elemental, burn up calculation, conversion ratio, and neutron flux in each core will be discussed.