Enhancing Thermal-Hydraulic Modelling in Dual Fluid Reactor Demonstrator: The Impact of Variable Turbulent Prandtl Number
In response to the growing demand for advanced nuclear reactor technologies, this study addresses significant gaps in thermal-hydraulic modelling for dual fluid reactors (DFRs) by integrating Kays correlation to implement a variable turbulent Prandtl number in the Reynolds-averaged Navier–Stokes (RA...
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| Main Authors: | , , |
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| Format: | Article |
| Language: | English |
| Published: |
MDPI AG
2025-01-01
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| Series: | Energies |
| Subjects: | |
| Online Access: | https://www.mdpi.com/1996-1073/18/2/396 |
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| Summary: | In response to the growing demand for advanced nuclear reactor technologies, this study addresses significant gaps in thermal-hydraulic modelling for dual fluid reactors (DFRs) by integrating Kays correlation to implement a variable turbulent Prandtl number in the Reynolds-averaged Navier–Stokes (RANS) simulations. Traditional approaches employing a constant value of the turbulent Prandtl number have proven inadequate, leading to inaccurate heat transfer predictions for low Prandtl number liquids. The study carefully selects the appropriate formula for the turbulent Prandtl number in the DFR context, enhancing the accuracy of thermal-hydraulic modelling. The simulations consider Reynolds numbers between 15,000 and 250,000, calculated based on the hydraulic diameters at different diameter pipes of the fuel and coolant loops. The molecular Prandtl number is equal to 0.025. Key findings reveal that uneven flow distributions within the fuel pipes result in variable temperature distribution throughout the reactor core, confirming earlier observations while highlighting significant differences in parameter values. These insights underscore the importance of model selection in CFD analysis for DFRs, revealing potential hotspots and high turbulence areas that necessitate further investigation into vibration and structural safety. The results provide a framework for improving reactor design and operational strategies, ensuring enhanced safety and efficiency in next-generation nuclear systems. Future work will apply this modelling approach to more complex geometries and flow scenarios to optimise thermal-hydraulic performance. |
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| ISSN: | 1996-1073 |