Metallic insulation is commonly used in reactor vessel because of its resistance to radiation and corrosion. Since the main mode of heat loss of reactor vessel is thermal radiation, the ability to prevent radiation heat transfer is important for metallic insulation. But the thermal conductivity of metallic insulation is difficult to calculate owing to their complex geometry. This article uses FLUENT 14.0 to obtain the important parameter “view factor”, and then develops a computational model of effective conductivity of metallic insulation. Heat transfer test of metallic insulation was done, and the numerical simulation of metallic insulation was also performed. Based on results of test and simulation, the computational model is modified. The modified model can fit the test result better. Based on the modified model, the effective conductivity of metallic insulation increases with the increase of temperature of hot side and cold side, among which the temperature of hot side influences more. And when the temperature is high, the effective conductivity increases much faster.
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2014 22nd International Conference on Nuclear Engineering
July 7–11, 2014
Prague, Czech Republic
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4589-9
PROCEEDINGS PAPER
A Study of Heat Transfer Calculation Method of Reactor Vessel Metallic Insulation
Yan Dapeng,
Yan Dapeng
Nuclear Power Institute of China, Chengdu, Sichuan, China
Search for other works by this author on:
Ying Luo
Ying Luo
Nuclear Power Institute of China, Chengdu, Sichuan, China
Search for other works by this author on:
Yan Dapeng
Nuclear Power Institute of China, Chengdu, Sichuan, China
Ying Luo
Nuclear Power Institute of China, Chengdu, Sichuan, China
Paper No:
ICONE22-30446, V001T03A016; 5 pages
Published Online:
November 17, 2014
Citation
Dapeng, Y, & Luo, Y. "A Study of Heat Transfer Calculation Method of Reactor Vessel Metallic Insulation." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 1: Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant; Nuclear Fuel and Materials; Plant Systems, Structures and Components; Codes, Standards, Licensing and Regulatory Issues. Prague, Czech Republic. July 7–11, 2014. V001T03A016. ASME. https://doi.org/10.1115/ICONE22-30446
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