Very High temperature gas-cooled reactor (VHTR), especially the pebble-bed core type reactor, is inevitable to take place the wear of graphite components and generate the graphite dust in the core. The graphite dust was taken away by helium coolant and deposited on the surface of the primary circuit, and the fission products may be absorbed on the dust. Since it is possible that the fission products are released with dust under the accident conditions such as depressurization events, they have a potential hazard of radiation exposure to the environment. In this paper, VHTR as the research object, a testing platform is to be built with the purpose of investigating the behavior of graphite dust emission during the accident conditions. Circuit loop design is used to simulate the primary system and nitrogen is used as the working substance. Experiments of graphite dust deposition and resuspension study, as well as the study of graphite dust emission behavior during the accident conditions in pebble-bed type design VHTR will be conducted on the testing platform. The experimental data will be used for the development of VHTR source term analysis modeling.
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18th International Conference on Nuclear Engineering
May 17–21, 2010
Xi’an, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4934-7
PROCEEDINGS PAPER
Design and Analysis of Testing Platform for Study of Graphite Dust Emission Behavior During Accident Conditions
Zhipeng Chen,
Zhipeng Chen
Tsinghua University, Beijing, China
Search for other works by this author on:
Suyuan Yu
Suyuan Yu
Tsinghua University, Beijing, China
Search for other works by this author on:
Zhipeng Chen
Tsinghua University, Beijing, China
Suyuan Yu
Tsinghua University, Beijing, China
Paper No:
ICONE18-30312, pp. 349-354; 6 pages
Published Online:
April 8, 2011
Citation
Chen, Z, & Yu, S. "Design and Analysis of Testing Platform for Study of Graphite Dust Emission Behavior During Accident Conditions." Proceedings of the 18th International Conference on Nuclear Engineering. 18th International Conference on Nuclear Engineering: Volume 6. Xi’an, China. May 17–21, 2010. pp. 349-354. ASME. https://doi.org/10.1115/ICONE18-30312
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