The dry storage system (DSS) is a major interim storage method for the spent nuclear fuel (SNF). The passive cooling concept and sealed fuel canister design provide the better safety and retrievability capability for SNF management. However, the sealed design of DSS causes that the internal status of transportable storage canister (TSC) is difficult to monitor for operation requirement. The issue means that many kinds of safety events, such as the re-filling gas leakage issue of TSC, half-blockage and fully-blockage cannot be early detected by present measurement method. In this study, a validated CFD simulation methodology has been employed to investigate the relationship between the natural convection behavior of DSS and shell temperature distribution of the TSC. All possible accidents (e.g., re-filling gas leakage, half-blockage, and fully-blockage) are considered in this investigation. The results show that those accidents will influence the convection behavior and further change the temperature profile of the TSC shell. The quantified comparison also shows that the minimum deviation of temperature is large enough to measured and will increased with heat load increasing. This results implying that the change of status of DSS can be determined by the temperature values and profile of TSC. Finally, a two measure point method has been purposed and proofed in this study. It proof that the major accident of DSS, such as the re-filling gas leakage, half-blockage and full blockage for a DSS can be confirmed through the purposed measured method.
- Nuclear Engineering Division
Investigating the Natural Convection Behavior in the Transportable Storage Canister of Dry Storage Systems Through CFD Simulation
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Tseng, Y, Chen, C, Ferng, Y, & Shih, C. "Investigating the Natural Convection Behavior in the Transportable Storage Canister of Dry Storage Systems Through CFD Simulation." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 4: Radiation Protection and Nuclear Technology Applications; Fuel Cycle, Radioactive Waste Management and Decommissioning; Computational Fluid Dynamics (CFD) and Coupled Codes; Reactor Physics and Transport Theory. Prague, Czech Republic. July 7–11, 2014. V004T10A010. ASME. https://doi.org/10.1115/ICONE22-30330
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