Neutronics and thermal-hydraulics coupling analyses on transient and accident behaviors of molten chloride salt fast reactor

ABSTRACT

Neutronics and thermal-hydraulics coupling transient analyses have been carried out to investigate the intrinsic safety characteristics and fuel temperature ranges of a molten chloride salt fast reactor. The analysis system includes a primary heat transport system and a secondary heat transport system connected by aheat exchanger and a decay heat removal system connected to the secondary system. The neutronics and thermal-hydraulics coupling analysis was performed with the RELAP5-3D code to analyze the entire plant behavior of each event, the temperature and velocity of the reactor inlet flow were obtained, and the reactor power was analyzed in detail following the FLUENT code. Through these analyses, the most severe events for the reactor were revealed. Analysis using the two codes showed that the reactor power calculated by RELAP5-3D was in good agreement with that of FLUENT although the detailed flow inside the core could not be reproduced.

Graphical Abstract

KEYWORDS:

• Molten salt fast reactor
• neutronics-thermal hydraulics coupling analysis
• FLUENT
• RELAP5-3D
• inherent safety

Nomenclature

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Acknowledgments

The author wishes to express his sincere thanks to Idaho National Laboratory for the RELAP5-3D code licensing. This research used Idaho National Laboratory computing resources, which are supported by the Office of Nuclear Energy of the US Department of Energy and the Nuclear Science User Facilities under Contract No. DE-AC07-05ID14517.

Disclosure statement

No potential conflict of interest was reported by the author(s).