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Comparison Between Thermal Analysis Results With FVM and Sub-channel Codes in a Single Assembly Heat Transfer Test
Dry storage is a predominantly used method as a spent nuclear fuel storage system after spent nuclear fuel is cooled in the spent fuel pool. Spent nuclear fuel is highly radioactive and it generates heat called decay heat originated by fission products and radiation. Therefore, temperature of spent nuclear fuel should be predicted whether its cladding temperature is maintained under 400°C, which is the allowable temperature limit of cladding in a dry storage. ANSYS Fluent and COBRA-SFS are predominantly used computational method to investigate the temperature of spent nuclear fuels in a dry storage. Herein, thermal analysis results with the methods were compared based on a Single Assembly Heat Transfer Test, which is a heat test with an electrically heated model of a single PWR fuel assembly in a dry cask performed at the Pacific Northwest Laboratory. Decay heat was 1kW and backfill gas was air. Fix temperature boundary condition is applied to inner shell according to measured temperature. In case of peak cladding temperature (PCT), Fluent predicted 240–284°C, while COBRA-SFS gave 243–292°C. The discrepancy between the codes is under 2.5%. The location where PCT took place was 3.65 m from the bottom of the assembly in both results. However, temperature difference between the upper and lower region of the assembly based on the Fluent was smaller than the temperature difference based on the COBRA-SFS. It means the heat was well transferred in an axial direction with Fluent compared to COBRA-SFS. In lower plenum region where air was naturally circulated, COBRASFS had disadvantages compared to Fluent because it modeled the lower plenum by single node, so it was hard to simulate convection heat transfer by natural circulation. natural circulation speed of air in a center region of the assembly was 0.07–0.1 m·s−1 in both cases.
