COBRA-SFS (Spent Fuel Storage): A thermal-hydraulic analysis computer code: Volume 3, Validation assessments

Lombardo, N.J.; Cuta, Judith M.; Michener, T.E.; Rector, David R.; Wheeler, C.L.

doi:10.2172/6883584

Cited by 8 publications

(11 citation statements)

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“…The results of the thermal analysis, which were obtained using the COBRA-SFS code (Rector 1986), are presented in Section 4.1. Source term calculations are presented in Section 4.2, followed by the shielding calculations in Section 4.3.…”

Section: Analytical Resultsmentioning

confidence: 99%

“…A lumped rod-and-channel arrangement was used to model the whole assemblies. Modeling studies (Rector et al 1986) have shown that this type of modeling gives adequate resolution of temperature gradients within an unconsolidated spent fuel assembly. This model requires only 57 channels and 105 rods to represent the assembly.…”

Section: Cask Model Used For Thermal Analysismentioning

confidence: 99%

See 1 more Smart Citation

Evaluation of Effect of Fuel Assembly Loading Patterns on Thermal and Shielding Performance of a Spent Fuel Storage/Transportation Cask

Cuta¹,

Jenquin²,

McKinnon³

2001

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Executive SummaryThe licensing of spent fuel storage casks is generally based on conservative analyses that assume a storage system uniformly loaded with design basis fuel. The design basis fuel typically assumes a maximum assembly enrichment, maximum burnup, and minimum cooling time. These conditions set the maximum decay heat loads and radioactive source terms for the design. Recognizing that reactor spent fuel pools hold spent fuel with an array of initial enrichments, burnups, and cooling times, this study was performed to evaluate the effect of load pattern on peak cladding temperature and cask surface dose rate.In 1991, DOE pursued a cooperative program with the Sacramento Municipal Utility District (SMUD). Part of the cooperative program was to demonstrate the effect of cask loading on the thermal and shielding performance of a cask containing spent nuclear fuel. At the time the cooperative agreement was established between DOE and SMUD, SMUD had fuel with cooling times ranging from 5 to 17 years, burnups of 8 to 38 GWd/MTU, and enrichments of 2.01 to 3.43% at its Rancho Seco nuclear power plant. It was thought that there was enough variability in the fuel to determine the effect of load pattern on dose rates. The effect of load pattern on the thermal performance of the casks was to be determined analytically using a computer code that had been validated extensively. With the termination of the DOE/SMUD cooperative agreement prior to cask loading, an analytical method was chosen to determine the effect of load pattern on dose rates. The codes were validated using data generated in other DOE cooperative efforts.The cask used in the analysis was similar to the TN-24P spent fuel storage cask. The TN-24P spent fuel storage cask is designed to hold 24 spent fuel assemblies with a total heat load of 24 kW. The cask body is forged steel surrounded by a resin layer for neutron shielding and a steel outer shell. The overall cask length is 16 ft (5.0 m), and the outer diameter is 7. The influence of load pattern on the thermal performance of the cask was determined through analysis with the COBRA-SFS code. The same COBRA-SFS model previously developed for analyzing the TN-24P cask was used in this parametric study. Three radial power distributions were considered. Each load pattern had an average decay heat output of 1 kW per fuel assembly for a total output of 24 kW from the fully loaded cask. Seventeen different load patterns were selected, and three backfill conditions were considered for each load pattern. The backfill media selected were helium, nitrogen, and vacuum. Except for the backfill medium and the radial power distribution, nothing was changed in the COBRA-SFS input between cases. This combination of load patterns and backfill gases provided a total of 51 runs in the test matrix.iv

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Section: Analytical Resultsmentioning

confidence: 99%

Section: Cask Model Used For Thermal Analysismentioning

confidence: 99%

Evaluation of Effect of Fuel Assembly Loading Patterns on Thermal and Shielding Performance of a Spent Fuel Storage/Transportation Cask

Cuta¹,

Jenquin²,

McKinnon³

2001

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“…The detailed model of the canister and internals (including the fuel assemblies) shown in Figure 2-2 has the typical mesh resolution generally used for the basket structure in COBRA-SFS models of spent fuel storage systems. Finer mesh resolution can be specified, if needed, but comparison with temperature measurements from single assembly and multiassembly experiments, including testing of storage systems with spent fuel loaded in the basket (Lombardo et al 1986;Rector et al 1986;Creer et al 1987) has shown that this meshing is sufficient for resolution of temperature gradients typical of spent fuel storage systems. The mesh includes the basket plates, poison plates, and basket support structures, including the shims on these structures that are used to ensure firm contact between the basket frame and canister inner shell.…”

Section: Cobra-sfs Model Descriptionmentioning

confidence: 99%

“…It is applicable to both steady-state and transient conditions in single-phase gas-cooled spent fuel packages with radiation, convection, and conduction heat transfer. The code has been validated in blind pretest calculations using test data from spent fuel packages loaded with actual spent fuel assemblies as well as electrically heated single-assembly tests (Creer et al 1987, Lombardo et al 1986).…”

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confidence: 99%

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

Cuta¹,

Adkins²

2014

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This report fulfills the M4 milestone M4FT-14PN0810037. Thermal analysis is being undertaken at Pacific Northwest National Laboratory (PNNL) in support of inspections of selected storage modules at various locations around the United States, as part of the Used Fuel Disposition Campaign of the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) Fuel Cycle Research and Development. This report documents pre-inspection predictions of temperatures for two modules at the Diablo Canyon Power Plant ISFSI identified as candidates for inspection. These are HI-STORM 100 modules of a site-specific design for storing PWR 17x17 fuel in MPC-32 canisters. The temperature predictions reported in this document were obtained with detailed COBRA-SFS models of these storage systems, with the following boundary conditions and assumptions.

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“…The data indicatesthat four or more rods may have developed leaks before TC lance insertionbefore testing, three or more rods during testing, and another 5 in the six months after testing. McKinnon 1986McKinnon , 1987aMcKinnon , 1987bMcKinnon , 1989McKinnon , 1992 or in the code manuals (Rector 1986a(Rector , 1986bLombardo 1986;McCann 1987aMcCann , 1987bMcCann , 1987c HYDRA has been designed to provide a user-orientedinput interface, which eliminatesthe need for internal code changes. Any applicationfor which the code is an appropriatechoice can be completelydescribed through the constructionof an input file.…”

Section: 18mentioning

confidence: 99%

Spent nuclear fuel storage -- Performance tests and demonstrations

McKinnon¹,

DeLoach²

1993

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This report summarizes the results of heat transfer and shielding performance tests and demonstrations conducted from 1983 through 1992 by or in cooperation with the U. S. Department of Energy (DOE), Office of Commercial Radioactive Waste Management (OCRWM). The performance tests cor,sisted of 6 to tions was estimated to be extremely small, and has had no adverse effect on testing efforts. iii ACRONYMS AND INITIALISMS BWR boil ing water reactor CCP

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COBRA-SFS (Spent Fuel Storage): A thermal-hydraulic analysis computer code: Volume 3, Validation assessments

Cited by 8 publications

References 0 publications

Evaluation of Effect of Fuel Assembly Loading Patterns on Thermal and Shielding Performance of a Spent Fuel Storage/Transportation Cask

Evaluation of Effect of Fuel Assembly Loading Patterns on Thermal and Shielding Performance of a Spent Fuel Storage/Transportation Cask

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

Spent nuclear fuel storage -- Performance tests and demonstrations

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