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    Experimental and Numerical Investigation of Solidification of Gallium in an Initially Emptied Horizontal Pipe Flow

    Source: Journal of Nuclear Engineering and Radiation Science:;2021:;volume( 007 ):;issue: 003::page 031602-1
    Author:
    Somers-Neal, Shawn
    ,
    Pegarkov, Alex
    ,
    Matida, Edgar
    ,
    Tang, Vinh
    ,
    Kaya, Tarik
    DOI: 10.1115/1.4049096
    Publisher: The American Society of Mechanical Engineers (ASME)
    Abstract: In a reactor core meltdown under postulated severe accidents, the molten material (corium) could be ejected or relocated through existing vessel penetrations (cooling pipe connections), thus potentially contaminating other locations in the power plant. There exists, however, a potential for plugging of melt flow due to its complete solidification, providing the availability of an adequate heat sink. Therefore, a numerical model was created to simulate the flow of molten metal through an initially empty horizontal pipe. The numerical model was verified using a previously developed analytical model and validated against experimental tests with gallium (low melting temperature) as a substitute for corium. The numerical model was able to predict the penetration length (length of distance traveled by the molten metal) after a complete blockage occurred with an average percent error range of 9%. Since the numerical model has been verified and validated, the model was updated to predict the penetration length in the cooling pipe in case of a severe accident. The model was used to predict the penetration length for different Reynolds numbers and pipe diameters, which resulted in the range of penetration length from about 0.33 m to 0.93 m.
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      Experimental and Numerical Investigation of Solidification of Gallium in an Initially Emptied Horizontal Pipe Flow

    URI
    http://yetl.yabesh.ir/yetl1/handle/yetl/4276543
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    • Journal of Nuclear Engineering and Radiation Science

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    contributor authorSomers-Neal, Shawn
    contributor authorPegarkov, Alex
    contributor authorMatida, Edgar
    contributor authorTang, Vinh
    contributor authorKaya, Tarik
    date accessioned2022-02-05T21:54:07Z
    date available2022-02-05T21:54:07Z
    date copyright3/16/2021 12:00:00 AM
    date issued2021
    identifier issn2332-8983
    identifier otherners_007_03_031602.pdf
    identifier urihttp://yetl.yabesh.ir/yetl1/handle/yetl/4276543
    description abstractIn a reactor core meltdown under postulated severe accidents, the molten material (corium) could be ejected or relocated through existing vessel penetrations (cooling pipe connections), thus potentially contaminating other locations in the power plant. There exists, however, a potential for plugging of melt flow due to its complete solidification, providing the availability of an adequate heat sink. Therefore, a numerical model was created to simulate the flow of molten metal through an initially empty horizontal pipe. The numerical model was verified using a previously developed analytical model and validated against experimental tests with gallium (low melting temperature) as a substitute for corium. The numerical model was able to predict the penetration length (length of distance traveled by the molten metal) after a complete blockage occurred with an average percent error range of 9%. Since the numerical model has been verified and validated, the model was updated to predict the penetration length in the cooling pipe in case of a severe accident. The model was used to predict the penetration length for different Reynolds numbers and pipe diameters, which resulted in the range of penetration length from about 0.33 m to 0.93 m.
    publisherThe American Society of Mechanical Engineers (ASME)
    titleExperimental and Numerical Investigation of Solidification of Gallium in an Initially Emptied Horizontal Pipe Flow
    typeJournal Paper
    journal volume7
    journal issue3
    journal titleJournal of Nuclear Engineering and Radiation Science
    identifier doi10.1115/1.4049096
    journal fristpage031602-1
    journal lastpage031602-10
    page10
    treeJournal of Nuclear Engineering and Radiation Science:;2021:;volume( 007 ):;issue: 003
    contenttypeFulltext
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