The behavior of the graphite dust has an important effect on the safety analysis of High Temperature Gas-cooled Reactors. The graphite dust with the large size will deposit at the bottom of the reactor core by gravity, while the graphite dust with the small size will flow in the primary circuit by carrying of helium. These suspended dusts can deposit in primary loop surface and influence the surface’s feature such as fouling resistance. Moreover, fission products released by fuel elements would enter primary loop and combine with dust, resulting in making the maintenance and repair of steam generator difficult. On the other hand, in the loss-of-coolant accident, when the discharge pipe of fuel pellets was broken, helium will be rejected with the speed of sound. The graphite dust with radioactive contamination will be carried by helium into the environment. The present study experimentally investigates the resuspension of graphite dust in the process of depressurization accident. The effect of both the initial pressure and the velocity of the gas on the resuspension of graphite dust were studied.
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2013 21st International Conference on Nuclear Engineering
July 29–August 2, 2013
Chengdu, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5579-9
PROCEEDINGS PAPER
Study of Graphite Dust Releasing Behavior in a Depressurization Accident of HTR
Ya-nan Zhen,
Ya-nan Zhen
Tsinghua University, Beijing, China
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Xiao-yong Yang,
Xiao-yong Yang
Tsinghua University, Beijing, China
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Su-yuan Yu
Su-yuan Yu
Tsinghua University, Beijing, China
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Wei Peng
Tsinghua University, Beijing, China
Ya-nan Zhen
Tsinghua University, Beijing, China
Xiao-yong Yang
Tsinghua University, Beijing, China
Su-yuan Yu
Tsinghua University, Beijing, China
Paper No:
ICONE21-15324, V002T03A008; 6 pages
Published Online:
February 7, 2014
Citation
Peng, W, Zhen, Y, Yang, X, & Yu, S. "Study of Graphite Dust Releasing Behavior in a Depressurization Accident of HTR." Proceedings of the 2013 21st International Conference on Nuclear Engineering. Volume 2: Plant Systems, Construction, Structures and Components; Next Generation Reactors and Advanced Reactors. Chengdu, China. July 29–August 2, 2013. V002T03A008. ASME. https://doi.org/10.1115/ICONE21-15324
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