A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation.
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10th International Conference on Nuclear Engineering
April 14–18, 2002
Arlington, Virginia, USA
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
0-7918-3597-9
PROCEEDINGS PAPER
Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes
M. J. Holowach,
M. J. Holowach
Pennsylvania State University, University Park, PA
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L. E. Hochreiter,
L. E. Hochreiter
Pennsylvania State University, University Park, PA
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F.-B. Cheung
F.-B. Cheung
Pennsylvania State University, University Park, PA
Search for other works by this author on:
M. J. Holowach
Pennsylvania State University, University Park, PA
L. E. Hochreiter
Pennsylvania State University, University Park, PA
F.-B. Cheung
Pennsylvania State University, University Park, PA
Paper No:
ICONE10-22651, pp. 1035-1040; 6 pages
Published Online:
March 4, 2009
Citation
Holowach, MJ, Hochreiter, LE, & Cheung, F. "Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes." Proceedings of the 10th International Conference on Nuclear Engineering. 10th International Conference on Nuclear Engineering, Volume 3. Arlington, Virginia, USA. April 14–18, 2002. pp. 1035-1040. ASME. https://doi.org/10.1115/ICONE10-22651
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