Thermal-hydraulic analyses of the TN-24P cask loaded with consolidated and unconsolidated spent nuclear fuel

This paper presents the results of comparisons of COBRA-SFS (spent fuel storage) temperature predictions with experimental data from the TN-24P (Transnuclear) spent fuel storage cask loaded with unconsolidated and consolidated spent PWR fuel. Peak cladding temperature predictions using the COBRA-SFS code are compared with test data and predicted axial and radial temperature distributions are compared with measured temperature profiles. The pre-tested accuracy of the COBRA-SFS code in predicting temperature distributions (before the experimental data were obtained) is discussed, along with the effect of post-test model improvements on temperature predictions. This paper also briefly describes the COBRA-SFS code, which is designed to accurately predict flow and temperature distributions in spent nuclear fuel storage and transportation systems