Preliminary Thermal Modeling of HI-STORM 100 Storage Modules at Diablo Canyon Power Plant ISFSI

Cuta, Judith M.; Adkins, Harold E.

doi:10.2172/1134526

Cited by 2 publications

(3 citation statements)

References 4 publications

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“…The ANSYS/Fluent software can simulate the thermal behaviors and capture the important features of a vertical storage cask. Based on the validation results, it is indicated that a reasonable tighter convergence criteria is needed to obtain a more accurate temperature profiles and thus peak temperatures, yet in an with measurement data [26] and those obtained in simulations by Cuta et al [25]. 0 degrees 27 acceptable simulation time cost.…”

Section: Discussionmentioning

confidence: 99%

“…As of December 2013, the total decay heat load of the HI-STORM 100 storage cask at the Diablo Canyon ISFSI was 17.05 kW, which was distributed among the 32 PWR spent fuel assemblies as shown in Figure 13 [25,26]. A heat load of 4.26 kW was used to represent the quarter region in x(4,6)-y(3,1) in Figure 13, which is approximately a quarter of the total decay heat of 17.05 kW.…”

Section: Heat Loads and Boundary Conditionsmentioning

confidence: 99%

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Thermal modeling of a vertical dry storage cask for used nuclear fuel

Liu

2016

Nuclear Engineering and Design

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Thermal modeling of temperature profiles of dry casks has been identified as a high-priority item in a U.S. Department of Energy gap analysis. In this work, a three-dimensional model of a vertical dry cask has been constructed for computer simulation by using the ANSYS/FLUENT code. The vertical storage cask contains a welded canister for 32 Pressurized Water Reactor (PWR) used-fuel assemblies with a total decay heat load of 34 kW. To simplify thermal calculations, an effective thermal conductivity model for a 17x17 PWR used (or spent)-fuel assembly was developed and used in the simulation of thermal performance. The effects of canister fill gas (helium or nitrogen), internal pressure (1-6 atm), and basket material (stainless steel or aluminum alloy) were studied to determine the peak cladding temperature (PCT) and the canister surface temperatures (CSTs). The results showed that high thermal conductivity of the basket material greatly enhances heat transfer and reduces the PCT. The results also showed that natural convection affects both PCT and the CST profile, while the latter depends strongly on the type of fill gas and canister internal pressure. Of particular interest to condition and performance monitoring is the identification of canister locations where significant temperature change occurs after a canister is breached and the fill gas changes from high-pressure helium to ambient air. This study provided insight on the thermal performance of a vertical storage cask containing high-burnup fuel, and helped advance the concept of monitoring CSTs as a means to detect helium leakage from a welded canister. The effects of blockage of air inlet vents on the cask's thermal performance were studied. The simulation were validated by comparing the results against data obtained from the temperature measurements of a commercial cask. 3 Nomenclature abbreviations AMP aging management program CFD computational fluid dynamics CISCC chloride-induced stress corrosion cracking CST canister surface temperatures DCSS dry cask storage system DOE Department of Energy GWd/MTU gigawatts day/metric ton uranium ISFSI independent spent fuel storage installations MPC multiple purpose canister SIMPLE semi-implicit method for pressure linked equations PCT peak cladding temperature PWR pressurized water reactor SSC structures, systems, and components SST k- shear stress transport k-ω turbulence model UDF user defined function

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Section: Discussionmentioning

confidence: 99%

Section: Heat Loads and Boundary Conditionsmentioning

confidence: 99%

Thermal modeling of a vertical dry storage cask for used nuclear fuel

Liu

2016

Nuclear Engineering and Design

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“…The total number of power irradiation, flux irradiation and decay commands in the ORIGEN input file must be less or equal to 150. Although the last updating of ORIGEN was in 2002, it is still used as a standalone code [4,5,6] or in combination with other codes [7]. ORIGEN is still used especially in safety analysis because it is more conservative than some other codes, such as ORIGEN-ARP [8].…”

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confidence: 99%

Cooling Period Calculation of Evolutionary Power Reactor Spent Fuel for Dry Management Safety

Youssef

Sultan

2016

Nucl. and Rad. Safe.

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The calculation of the evolutionary power reactor (EPR) spent fuel (SF) cooling period (CP) was performed. The CP was determined by comparing the heat load of a cask with the calculated value of EPR decay heat (DH). The EPR DH was calculated by the ORIGEN computer code based on the EPR parameters. For conservatively study, the EPR and ORIGEN parameters that lead to higher DH values were selected and safety margins were considered. The fitting tool was utilized in the calculation of CP to overcome the ORIGEN limitation. The resultant values of CP will maintain the peak cladding temperature (PCT) of SF lower than 400°C during storage, transport, and disposal. The results show that -for normal operation- the SF of EPR should stay in the pool at least 4.75 years before it is loaded to the passively cooled dry casks.

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Preliminary Thermal Modeling of HI-STORM 100 Storage Modules at Diablo Canyon Power Plant ISFSI

Cited by 2 publications

References 4 publications

Thermal modeling of a vertical dry storage cask for used nuclear fuel

Thermal modeling of a vertical dry storage cask for used nuclear fuel

Cooling Period Calculation of Evolutionary Power Reactor Spent Fuel for Dry Management Safety

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