The two-phase flow instability of forced convection has been experimentally investigated in a vertical narrow channel with the hydraulic diameter of 2.857mm and aspect ratio of 20. Transparent, metallic and conductive films on external surfaces of the test section can provide visualization and uniform heating for deionized water. The heat flux is 6–18.2 kW · m−2. When the instability occurs at low vapor quality, a series of parameters are measured and visualized images are obtained by a high-speed camera. The results show that the large amplitude of pressure drop between the inlet and outlet in the test section is due to the elongated bubble, and the value of pressure drop is positively correlated with the volume of the bubble. The oscillation period of pressure drop decreases with the increase of heat flux, and the period can be determined by the method of the Fast Fourier Transform. The backflow phenomenon is analyzed, which has a greater effect on the oscillation of pressure drop than bubble nucleation, bubble growth, bubble coalescence and recoiling of bubble boundary.
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2018 26th International Conference on Nuclear Engineering
July 22–26, 2018
London, England
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5149-4
PROCEEDINGS PAPER
Experimental Investigation and Flow Visualization of the Two-Phase Flow Instability at Low Vapor Quality in a Vertical Narrow Channel
Liqiang Pan,
Liqiang Pan
Tsinghua University, Beijing, China
Search for other works by this author on:
Yefei Liu
Yefei Liu
Tsinghua University, Beijing, China
Search for other works by this author on:
Liqiang Pan
Tsinghua University, Beijing, China
Yang Liu
Tsinghua University, Beijing, China
Weihua Li
Tsinghua University, Beijing, China
Yefei Liu
Tsinghua University, Beijing, China
Paper No:
ICONE26-82052, V06BT08A026; 7 pages
Published Online:
October 24, 2018
Citation
Pan, L, Liu, Y, Li, W, & Liu, Y. "Experimental Investigation and Flow Visualization of the Two-Phase Flow Instability at Low Vapor Quality in a Vertical Narrow Channel." Proceedings of the 2018 26th International Conference on Nuclear Engineering. Volume 6B: Thermal-Hydraulics and Safety Analyses. London, England. July 22–26, 2018. V06BT08A026. ASME. https://doi.org/10.1115/ICONE26-82052
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