Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Thermal dehydration tests of FLiNaK salt for thermal-hydraulic experiments

  • Shuai Che (Department of Nuclear Engineering & Radiological Sciences, University of Michigan) ;
  • Sheng Zhang (Department of Nuclear Engineering & Radiological Sciences, University of Michigan) ;
  • Adam Burak (Department of Nuclear Engineering & Radiological Sciences, University of Michigan) ;
  • Xiaodong Sun (Department of Nuclear Engineering & Radiological Sciences, University of Michigan)
  • Received : 2023.06.29
  • Accepted : 2024.01.24
  • Published : 2024.03.25
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Abstract

Fluoride-salt-cooled High-temperature Reactor (FHR) is a promising nuclear reactor technology. Among many challenges presented by the molten fluoride salts is the corrosion of salt-facing structural components. Higher moisture contents, in the FLiNaK (LiF-NaF-KF, 46.5-11.5-42 mol%) salt, aggravate intergranular corrosion and pitting for the given alloys. Therefore, several thermal dehydration tests of FLiNaK salt were performed with a batch size suitable for thermal-hydraulic experiments. Thermogravimetric Analysis (TGA) was performed for the three constituent fluoride salts individually. Preliminary thermal dehydration plans were then proposed for NaF and KF salts based on the TGA curves. However, the dehydration process may not be required for LiF since its low mass loss (<1.3 wt%). To evaluate the performance of these thermal dehydration plans, a batch-scale salt dehydration test facility was designed and constructed. The preliminary thermal dehydration plans were tested by varying the heating rates, target temperature, and holding time. The sample mass loss data showed that the high temperatures (>500 ℃) were necessary to remove a significant amount of moisture (>1 wt%) from NaF salt, while relatively low temperatures (around 300 ℃) with a long holding time (>10 h) were sufficient to remove most of the moisture from KF salt.

Keywords

Acknowledgement

This research was performed using funding received from the U.S. Department of Energy (DOE) Office of Nuclear Energy's Nuclear Energy University Program (NEUP), award number DE-NE0008977. The authors appreciate the financial support from the DOE NEUP Office. The authors would also like to thank Dr. Russell Bornschein of the University of Michigan, Prof. Minghui Chen and Mr. Yuqi Liu of the University of New Mexico for their valuable commentary and technical support of this work.

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