COBRA-SFS predictions of single assembly spent fuel heat transfer data

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

doi:10.2172/5513095

Cited by 4 publications

(8 citation statements)

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“…The surface emissivity of polished 316SS increases from 0.28 at 24°C to 0.57 at 232°C. * The surface emissivity values were a major source of uncertainty in simulations in previous work, with low confidence in these values and small changes in the fuel tube emittance representing a large change in the radiative heat transfer to (and from) the fuel tube (Lombardo et al, 1986).…”

Section: Uncertainties In Materials Thermal Propertiesmentioning

confidence: 89%

“…The average basket temperature was used to predict thermal expansion in McKinnon et al (1989), and the gap conductance between fuel pellet and cladding is assumed constant (Rector and Michener, 1989). (Creer et al, 1987;McKinnon et al, 1986McKinnon et al, , 1989McKinnon et al, and 1992Rector et al, 1986 (Creer et al, 1987;McKinnon et al, 1986;Rector et al, 1986a;Rector et al, 1986b (Rector et al, 1986a) Cask stainless steel inner liner 0.6 (Lombardo et al, 1986) Cask surface (stainless steel) 0.3 (Rector et al, 1986b) In addition to uncertainty in standard/reference values for thermal properties, there also exists uncertainty in the temperature dependence. Several of the thermal properties are also a function of temperature and therefore change over time as the cask (and contents) cools and is subject to changes in atmospheric temperature.…”

Section: Uncertainties In Materials Thermal Propertiesmentioning

confidence: 99%

“…Similarly, the surface emissivity depends on temperature, fabrication and surface finish, including polishing and subsequent oxide buildup (Lombardo et al, 1986). The surface emissivity of polished 316SS increases from 0.28 at 24°C to 0.57 at 232°C.…”

Section: Uncertainties In Materials Thermal Propertiesmentioning

confidence: 99%

“…o using separated laboratory components to determine contact resistance (McKinnon et al, 1986) o large uncertainties exist in the contact resistance between the heater rod and the flange, as well as the contact heat transfer area (Lombardo et al, 1986)…”

Section: Uncertainties In the Manufacturing And Assembling Processesmentioning

confidence: 99%

“…2. Composition of Available Data and Models .............................................................................. 11 2.1 (Cuta and Adkins, 2014) and Cask Array (Easton, 2012) (Creer et al, 1987;McKinnon et al, 1986McKinnon et al, , 1989McKinnon et al, and 1992Rector et al, 1986) Table 2-3 Density, thermal conductivity and specific heat capacity of gases at 200°C (Incropera and DeWitt, 1990)........................................................................................................ 16 Table A-1 General description of maturity levels in PCMM elements (Oberkampf et al 2007).............................................................................................................................................…”

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confidence: 99%

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LLNL Input to SNL Report on the Composition of Available Data for Used Nuclear Fuel Storage and Transportation Analysis

Sutton¹,

Wen²

2014

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Section: Uncertainties In Materials Thermal Propertiesmentioning

confidence: 89%

Section: Uncertainties In Materials Thermal Propertiesmentioning

confidence: 99%

Section: Uncertainties In Materials Thermal Propertiesmentioning

confidence: 99%

Section: Uncertainties In the Manufacturing And Assembling Processesmentioning

confidence: 99%

mentioning

confidence: 99%

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LLNL Input to SNL Report on the Composition of Available Data for Used Nuclear Fuel Storage and Transportation Analysis

Sutton¹,

Wen²

2014

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COBRA-SFS: A thermal-hydraulic analysis code for spent fuel storage and transportation casks

Michener¹,

Rector²,

Cuta³

et al. 1995

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Printod inThis document presents a complete description of Cycle 2 of COBRA-SFS, and consists of three main parts. Part I describes the conservation equations, constitutive models, and solution methods used in the code. Part 11 presents the User Manual, with guidance on code applications, and complete input instructions. This part also includes a detailed description of the auxiliary code RADGEN, used to generate grey body view factors required as input for radiative heat transfer modeling in the code. Part I11 describes the code structure, platform dependent coding, and program hierarchy. Installation instructions are also given for the various platform versions of the code that are available.iii

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Testing and analyses of the TN-24P PWR spent-fuel dry storage cask loaded with consolidated fuel

McKinnon¹,

Michener²,

Jensen³

et al. 1989

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SUBJECTS Light water reactor fuel I High-level radioactive waste management TOPICS Spent-fuel storage Thermal hydraulics models AUDIENCE Fuels engineers I R&D scientists Heat transfer Shielding Testing and Analyses of the TN-24P PWR Spent-Fuel Dry Storage Cask Loaded With Consolidated Fuel Full-scale testing has confirmed that the TN-24P storage cask offers a technically sound and practical method for storing consolidated spent fuel. COBRA-SFS code predictions of cask performance at conditions near its design limits agreed very well with actual test data. BACKGROUND As at-reactor storage basins attain maximum capacity, many utilities are expected to implement dry spent-fuel storage systems. To demonstrate the storage of dry spent fuel in large metal casks, EPRI and DOE have sponsored tests of metal casks loaded with unconsolidated fuel at the Idaho National Engineering Laboratory (INEL). This most recent study was initiated to investigate a TN-24P cask containing consolidated fuel. OBJECTIVES To demonstrate the thermal, shielding, and operational performance of the TN-24P cask loaded with consolidated spent nuclear fuel; to assess the ability of the COBRA-SFS heat transfer code (developed by Pacific Northwest Laboratory) to model the cask system and predict thermal performance. APPROACH Prior to the testing, the TN-24P cask contained 24 unconsolidated PWR assemblies from Virginia Power's Surry nuclear power station. The project team replaced the assemblies with 24 canisters of spent fuel, consolidated at a ratio of two assemblies per canister. INEL's rod consolidation project provided the filled test canisters. Researchers used the COBRA-SFS computer code to predict cask thermal performance. The team then instrumented and tested the cask in horizontal and vertical positions with three internal storage environments (nitrogen, helium, and vacuum). They compared the COBRA-SFS predictions with actual test data, refined the code to reflect test results, and performed posttest predictions. Transnuclear, Inc., the cask manufacturer, sponsored an additional test to simulate the insulating influence of impact limiters. RESULTS The TN-24P cask is well suited to store consolidated spent fuel. Its heat transfer performance was exceptionally good, as peak cladding temperatures for a cask heat load of 23.3 kW were well under 300°C with helium, ORDERING INFORMATION Requests for copies of this report should be directed to Research Reports Center (ARC), Box 50490, Palo Alto, CA 94303, (415) 965-4081. There is no charge for reports requested by EPRI member utilities and affiliates, U.S. utility associations, U.S. government agencies (federal, state, and local), media, and foreign organizations with which EPRI has an information exchange agreement. On request, ARC will send a catalog of EPRI reports.

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COBRA-SFS predictions of single assembly spent fuel heat transfer data

Cited by 4 publications

References 1 publication

LLNL Input to SNL Report on the Composition of Available Data for Used Nuclear Fuel Storage and Transportation Analysis

LLNL Input to SNL Report on the Composition of Available Data for Used Nuclear Fuel Storage and Transportation Analysis

COBRA-SFS: A thermal-hydraulic analysis code for spent fuel storage and transportation casks

Testing and analyses of the TN-24P PWR spent-fuel dry storage cask loaded with consolidated fuel

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