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Persistent URL http://purl.org/net/epubs/work/47947065
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Record Id 47947065
Title Modelling the Draining of a Molten Chloride Salt Reactor
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Abstract Molten salt reactors (MSR) are one of the advanced reactor concepts being developed under the guidance of the Generation IV International Forum, with the aim of improving the economic and environmental sustainability of nuclear power, as well as reactor reliability and safety. In order to ensure the safety of MSR concepts, the safe shutdown of the reactor must be demonstrated. One strategy to safely shutdown a molten salt reactor under accident conditions is to drain the liquid corium into a series of subcritical tanks. A preliminary study is being performed that examines the amount of fuel-salt which must be removed from a spherical molten salt reactor core to achieve a fast shutdown and how long the procedure takes to remove a sufficient amount. Numerical simulations of the neutronics of the molten salt reactor arrangement have been performed using the Monte Carlo neutron transport code SERPENT. In this study, the chloride based fuel-salt mixtures considered for the neutronics calculations were lithium chloride – uranium chloride and sodium chloride – uranium chloride. Uranium enrichment and reactor volume were adjusted to achieve a critical configuration for the proposed salt compositions to support a future salt choice. The optimal reactor configuration was then selected to perform further simulations to determine the liquid level at which the reactor core is considered sufficiently subcritical. Based on this data, initial fluid dynamic calculations are performed with the solver Code_Saturne to examine how quickly the molten salt can be withdrawn from the core. The volume of fluid module in Code_Saturne was applied to an isothermal reactor with constant physical properties applying conditions that simulated gravitational, pressured or pumped evacuation.
Organisation STFC , SCI-COMP , SCI-COMP-EE
Keywords chloride salt , liquid fuel draining , liquid metal coolant , molten salt reactor
Funding Information
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Language English (EN)
Type Details URI(s) Local file(s) Year
Paper In Conference Proceedings In 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH 2019), Portland, Oregon, 18-23 Aug 2019, (2019): 4470-4483. 2019