Dry storage of nuclear fuel is compromised by threats to the cladding integrity, such as creep and hydride reorientation. To predict these phenomena, spent fuel simulation codes have been developed. In spent fuel simulation, temperature information is the most influential factor for creep and hydride formation. Traditional fuel simulation codes required a user-defined temperature history input which is given by separate thermal analysis. Moreover, geometric changes in nuclear fuel, such as creep, can alter the cask’s internal subchannels, thereby changing the thermal analysis. This necessitates the development of a coupled thermal and nuclear fuel analysis code. In this study, we integrated the 2D FDM nuclear fuel code GIFT developed at SNU with COBRA -SFS. Using this, we analyzed spent nuclear stored in TN-24P dry storage cask over several decades and identified conditions posing threats due to phenomena like creep and hydrogen reorientation, represented by the burnup and peak cladding temperature at the start of dry storage. We also investigated the safety zone of spent nuclear fuel based on burnup and wet storage duration using decay heat.

Spent nuclear fuels should be safely stored until being disposed and dry storage system is predominantly used to retain the fuels. Thermal analysis to estimate temperatures of spent nuclear fuel and the storage system should be performed to evaluate whether the temperatures exceed safety limit. Recently, thermal hydraulic analysis with CFD codes is widely used to investigate the temperature of spent nuclear fuel in dry storage. COBRA-SFS is a legacy code based on subchannel analysis code, and its fidelity is verified for evaluating the thermal analysis for licensing a dry cask system. Herein, thermal analysis result based on CFD and COBRA-SFS codes is compared and the Dry Cask Simulator (DCS) is assessed as a benchmark experiment in this study. Extended Storage Collaborating Program (ESCP) led by the Electric Power Research Institute (EPRI) is organized to address the degradation effects of spent nuclear fuel during long-term dry storage, and DCS is the first phase of the program. The dry storage system, containing a single BWR assembly in a canister, was designed to produce validation-quality data for thermal analysis model. ANSYS FLUENT was used to simulate DCS. Simulations were conducted in various decay heat and helium pressure inside the canister. In realistic conditions of decay heat and helium pressure of actual dry cask system, CFD and COBRA-SFS analysis results gave good agreement with experimental measurement. Peak temperatures of channel can, basket, canister and shell predicted by CFD simulation also showed good prediction and the discrepancies were less than 7 K while measurements uncertainty was 7 K. In high decay heat and high pressure condition, however, CFD and COBRA-SFS underestimated peak cladding temperature than experimental results.

Detailed temperature distributions of the spent fuel are required to evaluate the long-term integrity of the dry storage system. In this study, a subchannel analysis method was established to obtain the detailed temperatures of a spent fuel using the COBRA-SFS code. The SAHTT (Single Assembly Heat Transfer Test) model was selected as the subchannel analysis. It was developed at the PNL to investigate heat transfer characteristics of spent PWR fuel under dry storage conditions. The SAHTT has a 15×15 rod array with simulated rods 0.42 in. (10.7 mm) in diameter. Control rod thimbles were modeled with unheated rods. The COBRA-SFS input consists a detailed subchannel model with 256 subchannels, 225 rods, and 8 slab nodes. The heat generation rate was axially uniform with total power of 1.0 kW. Subchannel analyses were performed for the vertical orientation under three different backfills of air, helium, and vacuum. For the vacuum backfill, the peak temperature was the highest and temperature gradients the sharpest only due to the radiation heat transfer effect. For the helium backfill, peak temperature was lowest and the axial profiles flattest due to the higher conductivity and lower density of helium. Subchannel analyses were also performed to evaluate the effect of thermal parameters such as surface emissivity, convective heat transfer coefficients, and flow resistance coefficients on the PCT (Peak Cladding Temperature). The PCT was affected by the emissivity of the fuel rod and the basket, and in particular, the basket emissivity had a greater effect. The PCT was affected by the Nusselt number, but the range of the Nusselt number is around 3.66. Therefore, the effect of the Nusselt number on the PCT will not be significant. As a result of the analysis according to the flow resistance coefficients, the PCT was affected by the wall friction factor, but the loss coefficients from the space grid had little effect. Subchannel technique obtained from this work can be used to predict the detailed temperature distributions of spent fuel assembly.