Numerical and experimental investigation of a new conceptual fluoride salt freeze valve for thorium-based molten salt reactor

Experimental studies have been developed on a new freeze plug concept for safety valves in facilities using molten salt. They are designed to allow the closure of an upstream circuit by solidifying the molten salt in a section of the device and to passively melt in case of a loss of electric power, thus releasing the upper fluid. The working principle of these cold plug designs relies on the control of the heat transfer balance inside the device, which determines whether the salt inside the cold plug solidifies or melts. The device is mainly composed of steel masses that are dimensioned to provide sufficient thermal heat storage to melt the salt and thus open the cold plug after the electric power is stopped. The final goal of the work is to provide useful recommendations and guidelines for the design of a cold plug for the emergency draining system of a molten salt reactor. Some numerical thermal simulations were performed with ANSYS mechanical (Finite Element Method) to be compared with results of the experiments and to make extrapolations for a new component to be used in a reactor.

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Batch-Scale Hydrofluorination of Li27BeF4 to Support Molten Salt Reactor Development

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4030963 ◽

2015 ◽

Vol 1 (4) ◽

Cited By ~ 11

Author(s):

Brian C. Kelleher ◽

Kieran P. Dolan ◽

Paul Brooks ◽

Mark H. Anderson ◽

Kumar Sridharan

Keyword(s):

Molten Salt ◽

Nuclear Reactor ◽

National Laboratory ◽

The United States ◽

Hydraulic Testing ◽

Molten Salt Reactor ◽

Fluoride Salt ◽

Oak Ridge ◽

University Of Wisconsin ◽

Analysis Process

Li 2 BeF 4 , or flibe, is the primary candidate coolant for the fluoride-salt-cooled high-temperature nuclear reactor (FHR). Kilogram quantities of pure flibe are required for repeatable corrosion tests of modern reactor materials. This paper details fluoride salt purification by the hydrofluorination–hydrogen process, which was used to regenerate 57.4 kg of flibe originating from the secondary loop of the molten salt reactor experiment (MSRE) at Oak Ridge National Laboratory (ORNL). Additionally, it expounds upon necessary handling precautions required to produce high-quality flibe and includes technological advancements which ease the purification and analysis process. Flibe batches produced at the University of Wisconsin are the largest since the MSRE program, enabling new corrosion, radiation, and thermal hydraulic testing around the United States.

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Quantify Sodium Fluoride / Beryllium Fluoride Salt Properties for a Liquid Fueled Fluoride Molten Salt Reactor

10.2172/1764854 ◽

2021 ◽

Author(s):

M. Rose ◽

E. Wu ◽

T. Lichtenstein ◽

J. Krueger ◽

S. Thomas ◽

...

Keyword(s):

Molten Salt ◽

Sodium Fluoride ◽

Molten Salt Reactor ◽

Fluoride Salt ◽

Beryllium Fluoride

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An Innovative Spherical Fuel Element to Inhibit the Infiltration of Liquid Fluoride Salt in Molten Salt Reactor

Volume 3: Nuclear Fuel and Material, Reactor Physics and Transport Theory; Innovative Nuclear Power Plant Design and New Technology Application ◽

10.1115/icone25-66639 ◽

2017 ◽

Author(s):

Yajuan Zhong ◽

Jun Lin ◽

Liujun Xu ◽

Haitao Jiang ◽

Zhiyong Zhu

Keyword(s):

Fuel Element ◽

Molten Salt ◽

Pore Diameter ◽

Coefficient Of Thermal Expansion ◽

Operating Temperature ◽

Molten Salt Reactor ◽

Fluoride Salt ◽

Average Pore ◽

Innovative Design ◽

Density Graphite

To inhibit the infiltration of liquid fluoride salt and easy to load and unload, fuel element in molten salt reactor (MSR) was isostatically pressed with an innovative design: A fuel-free low density graphite core of ≤ 30 mm diameter embedded in fuel-zone shell of ≥ 2.5 mm thickness, and then enveloped in a high density graphite shell of ≥ 5 mm thickness. Bulk density of the spherical fuel element can be designed from the range of 1.65–1.80 g/cm3, which is lower than the density of the liquid fluoride salt to make sure the fuel element can float in the MSR to load and unload. Characteristics of mercury infiltration and molten salt infiltration in graphite shell were investigated and compared with A3-3 graphite to identify the infiltration behaviors. The results indicated that the graphite shell has a low porosity about 9%, and an average pore diameter of 100 nm. The fluoride salt occupation of A3-3 was 10 wt% under 6.5 atm, whereas the salt gain did not infiltrate in graphite shell even up to 6.5 atm. It demonstrated that the outside graphite shell could inhibit the infiltration of liquid fluoride salt effectively. At the operating temperature of MSR (700 °C), thermal conductivity of graphite shell was 13.61 W/m K. The coefficient of thermal expansion (CTE) of outside graphite shell lied in 6.01×10−6 K−1 (α⫽) and 6.15×10−6 K−1 (α⊥) at the temperature range of 25–700 °C. The anisotropies factor of graphite shell calculated by CTE maintained below 1.12, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic properties of graphite shell are beneficial for the integrity and safety of the spherical fuel element for a MSR.

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