In order to better understand the stresses to which the flow distribution device (FDD) is subjected for the pressure fluctuation, we need to improve our knowledge of the fluid flow inside the reactor pressure vessel (RPV). The flow field of the reactor lower plenum which is associated with a typical pressurized water reactor (PWR) is simulated by using ANSYS CFX code. Calculations have been carried out from reactor pressure vessel inlet to the core outlet. Grid sizes of million nodes with the k-epsilon turbulent model have been used with a porous zone approach for the reactor core space. Predictions of the steady-state pressure and velocity field have been made. The results are compared with the scaled experiment data in order to verify the accurate description of the fluid flow. Based on this verified turbulent model, a sub-domain is extracted from the lower plenum for the transient two-way Fluid-Structure Interaction (FSI) simulation which is limited by the computer capability and computing time. This transient analysis of fluid-structure coupling system is conducted by using CFX and ANSYS in numeric calculation of flow field and structure, with an exchanging platform MFX-ANSYS/CFX which can transfer fluid pressure and structure displacement between computational fluid dynamics (CFD) and computational structure dynamics (CSD) grid systems. The loose coupling method is used to investigate the transient dynamic response of the flow distribution device which is immerged in the bottom plenum. Dynamic stress and strain of the flow distribution device are given and discussed. This analysis practice can be guidance for the optimization design of reactor and improve our understanding of reactor components flow induced vibration phenomena.
- Nuclear Engineering Division
Fluid-Structure Coupling Analysis for the Flow Distribution Device of Nuclear Reactor Internals
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Zhang, M, Duan, Y, Ran, X, & Liu, Y. "Fluid-Structure Coupling Analysis for the Flow Distribution Device of Nuclear Reactor Internals." 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. V003T10A009. ASME. https://doi.org/10.1115/ICONE21-15180
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