Studies on the Liquid Fluoride Thorium Reactor: Comparative Neutronics Analysis of MCNP6 Code with SRAC95 Reactor Analysis Code Based on FUJI-U3-(0)


The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.


Nuclear Engineering and Radiation Science

Keywords and Phrases

Codes (symbols); Fused salts; Molten materials; Nuclear reactors; FUJI-U3; Graphite-moderated; MCNP6; Neutronics; ORIGEN2; Reactor analysis; Reactor physics; Thorium reactor; Molten salt reactor; MSR

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Article - Journal

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© 2017 Elsevier, All rights reserved.

Publication Date

01 Apr 2017