The spacer grid is one of the structural components for the fuel assembly. In order to increase or extend the fuel life cycle, a spacer grid which has a much higher performance from the thermal/hydraulic and mechanical/structural point of view will be needed. From the thermal/hydraulic viewpoint, the CHF margin is very important in order to extend its life. Particularly, the mixing flow or cross flow between the subchannels have to be reinforced for this purpose. From the mechanical/structural viewpoint, the critical strength and the fuel rod support behaviour of a spacer grid are the same as the TH performance improvement for the next generation fuel. A computational fluid dynamics (CFD) analysis was performed to investigate the coolant mixing in a nuclear fuel bundle that is promoted by the mixing vane on the grid spacer. Single and multiple subchannels of one grid span of the fuel bundle were modeled to simulate a 5by5 rod array experiment with the mixing vane. The three-dimensional CFD models were generated by a structured multi-block method. The standard k-ε turbulence model was used in the current CFD simulation since it is practically useful and converges well for the complex turbulent flow in a nuclear fuel bundle. The CFD predictions of the axial and lateral mean flow velocities showed a somewhat larger difference from the experimental results near the spacer but represented the overall characteristics of the coolant mixing well in a nuclear fuel bundle with the mixing vane. Comparison of the single and multiple subchannel predictions shows a good agreement for the flow characteristics in the central subchannel of the rod array. The simulation of the multiple subchannels shows a slightly off-centered swirl in the peripheral subchannels due to the external wall of the rod array. It also shows no significant swirl and crossflow in the wall subchannels and the corner subchannels. In addition to this, the impact and the stress analysis of a spacer grid are accomplished by the FE method. The FE model was created using I-DEAS , and the ABAQUS/explicit version 6.3 commercial code was used for the solver. The FE analysis procedure was established, the FE analyses results were verified by the experimental method. The developed spacer grid will be evaluated from the thermal/hydraulic and mechanical/structural design criteria.
- Nuclear Engineering Division
Developed a Spacer Grid for the Future PWR Fuel Assembly by Considering the Thermal/Hydraulic and Mechanical/Structural Performance
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Yoon, K, In, W, Kang, H, & Song, K. "Developed a Spacer Grid for the Future PWR Fuel Assembly by Considering the Thermal/Hydraulic and Mechanical/Structural Performance." Proceedings of the 12th International Conference on Nuclear Engineering. 12th International Conference on Nuclear Engineering, Volume 2. Arlington, Virginia, USA. April 25–29, 2004. pp. 223-231. ASME. https://doi.org/10.1115/ICONE12-49106
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