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    Uncertainty Quantification in Support of Severe Accident Analysis Code User Confidence Using MELCOR-DAKOTA

    Source: Journal of Nuclear Engineering and Radiation Science:;2022:;volume( 008 ):;issue: 003::page 31703-1
    Author:
    Boafo
    ,
    Emmanuel K.;Numapau Gyamfi
    ,
    Emmanuel
    DOI: 10.1115/1.4053050
    Publisher: The American Society of Mechanical Engineers (ASME)
    Abstract: Uncertainty and Sensitivity analysis methods are often used in severe accident analysis for validating the complex physical models employed in the system codes that simulate such scenarios. This is necessitated by the large uncertainties associated with the physical models and boundary conditions employed to simulate severe accident scenarios. Uncertainty analysis involves primarily the estimation of the distribution of a key figure-of-merit (FOM) following the propagation of uncertainty in selected input parameters through a system code. The input parameters are sampled within defined ranges based on assigned probability distribution functions (PDFs) for the required number of code runs/realizations using stochastic sampling techniques. Input parameter selection is based on their importance to the key FOM, which is determined by the parameter identification and ranking table (PIRT). Sensitivity analysis investigates the contribution of each uncertain input parameter to the uncertainty of the selected FOM. In this study, the integrated severe accident analysis code MELCOR was coupled with a design analysis kit for optimization and terascale applications (DAKOTA), an optimization and uncertainty quantification tool in order to investigate the effect of input parameter uncertainty on hydrogen generation. The methodology developed was applied to the Fukushima Daiichi unit 1 nuclear power plant (NPP) accident scenario, which was modeled in another study. The results show that there is approximately 22.46% uncertainty in the amount of hydrogen generated as estimated by a single MELCOR run given the uncertainty in selected input parameters. The sensitivity analysis results also reveal that MELCOR input parameters: COR_SC 1141(melt flow rate per unit width at breakthrough candling), COR_ZP (porosity of fuel debris beds), and COR_EDR (characteristic debris size in the core region) contributed most significantly to the uncertainty in hydrogen generation.
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      Uncertainty Quantification in Support of Severe Accident Analysis Code User Confidence Using MELCOR-DAKOTA

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    http://yetl.yabesh.ir/yetl1/handle/yetl/4287364
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    • Journal of Nuclear Engineering and Radiation Science

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    contributor authorBoafo
    contributor authorEmmanuel K.;Numapau Gyamfi
    contributor authorEmmanuel
    date accessioned2022-08-18T13:03:47Z
    date available2022-08-18T13:03:47Z
    date copyright5/26/2022 12:00:00 AM
    date issued2022
    identifier issn2332-8983
    identifier otherners_008_03_031703.pdf
    identifier urihttp://yetl.yabesh.ir/yetl1/handle/yetl/4287364
    description abstractUncertainty and Sensitivity analysis methods are often used in severe accident analysis for validating the complex physical models employed in the system codes that simulate such scenarios. This is necessitated by the large uncertainties associated with the physical models and boundary conditions employed to simulate severe accident scenarios. Uncertainty analysis involves primarily the estimation of the distribution of a key figure-of-merit (FOM) following the propagation of uncertainty in selected input parameters through a system code. The input parameters are sampled within defined ranges based on assigned probability distribution functions (PDFs) for the required number of code runs/realizations using stochastic sampling techniques. Input parameter selection is based on their importance to the key FOM, which is determined by the parameter identification and ranking table (PIRT). Sensitivity analysis investigates the contribution of each uncertain input parameter to the uncertainty of the selected FOM. In this study, the integrated severe accident analysis code MELCOR was coupled with a design analysis kit for optimization and terascale applications (DAKOTA), an optimization and uncertainty quantification tool in order to investigate the effect of input parameter uncertainty on hydrogen generation. The methodology developed was applied to the Fukushima Daiichi unit 1 nuclear power plant (NPP) accident scenario, which was modeled in another study. The results show that there is approximately 22.46% uncertainty in the amount of hydrogen generated as estimated by a single MELCOR run given the uncertainty in selected input parameters. The sensitivity analysis results also reveal that MELCOR input parameters: COR_SC 1141(melt flow rate per unit width at breakthrough candling), COR_ZP (porosity of fuel debris beds), and COR_EDR (characteristic debris size in the core region) contributed most significantly to the uncertainty in hydrogen generation.
    publisherThe American Society of Mechanical Engineers (ASME)
    titleUncertainty Quantification in Support of Severe Accident Analysis Code User Confidence Using MELCOR-DAKOTA
    typeJournal Paper
    journal volume8
    journal issue3
    journal titleJournal of Nuclear Engineering and Radiation Science
    identifier doi10.1115/1.4053050
    journal fristpage31703-1
    journal lastpage31703-12
    page12
    treeJournal of Nuclear Engineering and Radiation Science:;2022:;volume( 008 ):;issue: 003
    contenttypeFulltext
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