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Gas Ingress during Side Vessel Break: Experiments and Numerical Simulations with a Simplified Pool-type Scaled Water Model

    Research outputpeer-review

    Abstract

    In the safety analysis of a pool-type liquid metal cooled reactor, the scenario of a large break of the reactor vessel is evaluated. Among the different possibilities the fall of the full bottom of the reactor vessel is analyzed as a bounding case for possible ingress of gas. It is analysed whether gas from the surroundings might enter the system and, more specifically, enter the core. The presence of gas would interrupt the cooling of the core region. A water model has been designed and studied both experimentally and numerically in several conditions, to evaluate the amount of gas entering from the side break of the vessel, taking into account different falling heights of the bottom of the vessel and to validate the ability of CFD to correctly reproduce the phenomena observed. The experimental data obtained with water with the simplified configuration were compared with a predictions by a CFD model with a 2D-axisymmetric domain. This confirmed the ability of CFD model to correctly predict the phenomena.

    Original languageEnglish
    Title of host publicationProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
    PublisherAmerican Nuclear Society
    Pages2534-2548
    Number of pages15
    ISBN (Electronic)9780894487934
    DOIs
    StatePublished - 2023
    Event2023 - NURETH: 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics - Washington Hilton, Washington D.C.
    Duration: 20 Aug 202325 Aug 2023
    https://www.ans.org/meetings/nureth20/

    Publication series

    NameProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023

    Conference

    Conference2023 - NURETH
    Abbreviated titleNURETH-20
    Country/TerritoryUnited States
    CityWashington D.C.
    Period2023-08-202023-08-25
    Internet address

    ASJC Scopus subject areas

    • Nuclear Energy and Engineering
    • Instrumentation

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