By application of CFD codes, 3D complex flow field and temperature field in the transfer canal are simulated. It indicates that there is obvious cross-flow in the complex flow path of spent fuel container, and the natural circulation flow brings an inhomogeneous heat-transfer condition to the spent fuel assembly. A model of typical unit of canal is built, and local flow field and temperature field are analyzed with CFD code. The results of thermal hydraulics analysis could be a good reference for nuclear safety research on spent fuel assembly. Considering a conservative operating condition, a case with conservative boundary conditions is investigated by CFD code with 3D model. The result shows that fuel rods in the local region, which are in the container upside and far from holes of wall and roof, get the worst cooling effect, and even maybe boiling occurs in clad surface. Based on the thermal-hydraulics analysis, optimized mechanical designs of container are suggested, the new design would optimize the local flow field, and increase the flow velocity of water, to construct more unhindered convection there, as a result, make spent fuel rods in the bad cooling region get better cooling.
Skip Nav Destination
2013 21st International Conference on Nuclear Engineering
July 29–August 2, 2013
Chengdu, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5580-5
PROCEEDINGS PAPER
Numerical Simulation on 3D Flow in the Fuel Transfer Canal and Local Flow Field Analysis
Qiang Guo
China Nuclear Power Engineering Co., Ltd., Beijing, China
Hui Wang
China Nuclear Power Engineering Co., Ltd., Beijing, China
Paper No:
ICONE21-16341, V003T10A041; 4 pages
Published Online:
February 7, 2014
Citation
Guo, Q, & Wang, H. "Numerical Simulation on 3D Flow in the Fuel Transfer Canal and Local Flow Field Analysis." Proceedings of the 2013 21st International Conference on Nuclear Engineering. Volume 3: Nuclear Safety and Security; Codes, Standards, Licensing and Regulatory Issues; Computational Fluid Dynamics and Coupled Codes. Chengdu, China. July 29–August 2, 2013. V003T10A041. ASME. https://doi.org/10.1115/ICONE21-16341
Download citation file:
8
Views
Related Proceedings Papers
Related Articles
Computational Fluid Dynamic Simulations of Heat Transfer From a 2 × 2 Wire-Wrapped Fuel Rod Bundle to Supercritical Pressure Water
ASME J of Nuclear Rad Sci (January,2018)
Transient Hydrodynamic Phenomena and Conjugate Heat Transfer During Cooling of Water in an Underground Thermal Storage Tank
J. Heat Transfer (February,2004)
Computational Fluid Dynamics Investigation of Supercritical Water Flow and Heat Transfer in a Rod Bundle With Grid Spacers
ASME J of Nuclear Rad Sci (July,2016)
Related Chapters
Utilities’ Perspective of Spent Fuel Storage
Global Applications of the ASME Boiler & Pressure Vessel Code
Insights and Results of the Shutdown PSA for a German SWR 69 Type Reactor (PSAM-0028)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Comparison of the Availability of Trip Systems for Reactors with Exothermal Reactions (PSAM-0361)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)