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CFD modeling and sensitivity analysis of ex-vessel core melt process

The result's identifiers

  • Result code in IS VaVaI

    <a href="https://www.isvavai.cz/riv?ss=detail&h=RIV%2F68407700%3A21220%2F19%3A00334866" target="_blank" >RIV/68407700:21220/19:00334866 - isvavai.cz</a>

  • Result on the web

  • DOI - Digital Object Identifier

Alternative languages

  • Result language

    angličtina

  • Original language name

    CFD modeling and sensitivity analysis of ex-vessel core melt process

  • Original language description

    The flow and heat transfer behavior of the ex-vessel core melt were investigated using a CFD code ANSYS Fluent along with the experimental data on the spreading of corium available in the literature (VULCANO VE-U7 test). In the numerical simulation of unsteady two-phase flow, the Eulerian Multi-Fluid VOF model was applied for spreading and interfacial surface formation of corium with the surrounding air. The main goal of this work was to provide mesh sensitivity analysis and to explore the impact of the additional heat sink levels towards the concrete substrate on the length of the corium spread. Results show the numerical stability and physical relevancy even for very coarse mesh, which allows for full-scale calculations covering the whole reactor cavity geometry. On the other hand, the effect of the heat flux magnitude transferred to the substrate has relatively low influence.

  • Czech name

  • Czech description

Classification

  • Type

    D - Article in proceedings

  • CEP classification

  • OECD FORD branch

    20303 - Thermodynamics

Result continuities

  • Project

    <a href="/en/project/TH02020666" target="_blank" >TH02020666: Advanced Analytical Tools for Severe Accident Simulations</a><br>

  • Continuities

    P - Projekt vyzkumu a vyvoje financovany z verejnych zdroju (s odkazem do CEP)

Others

  • Publication year

    2019

  • Confidentiality

    S - Úplné a pravdivé údaje o projektu nepodléhají ochraně podle zvláštních právních předpisů

Data specific for result type

  • Article name in the collection

    18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH 2019)

  • ISBN

    978-0-89448-767-5

  • ISSN

  • e-ISSN

  • Number of pages

    12

  • Pages from-to

    2640-2651

  • Publisher name

    American Nuclear Society

  • Place of publication

    New York

  • Event location

    Portland, Oregon

  • Event date

    Aug 18, 2019

  • Type of event by nationality

    WRD - Celosvětová akce

  • UT code for WoS article