Coupled Neutronics/Thermal Hydraulics Assessment of Graphite Moderated Molten Salt Reactors

A. Nalbandyan, E.B. Klinkby, B. Lauritzen, J. Groth-Jensen, R. Steyn

    Research output: Chapter in Book/Report/Conference proceedingArticle in proceedingsResearchpeer-review

    Abstract

    To model a Molten Salt Reactor (MSR) core, we apply a Multiphysics coupling scheme between the finite volume Computational Fluid Dynamics (CFD) code OpenFOAM and the Monte Carlo based neutronics code Serpent. The scheme employs the Serpent Multiphysics interface, which allows for high fidelity coupling to OpenFOAM by directly passing variable fields between the two codes. We simulate a graphite-moderated channel type MSR and compare the simulation results to data available on the Molten Salt Reactor Experiment (MSRE). Specifically, fuel and graphite temperature profiles and fuel velocity fields are derived for steady state operation and compared to the results of model calculations performed at the Oak Ridge National Laboratory (ORNL). A simple transient scenario of a step reactivity insertion is also modeled and the feedback of the system is evaluated and compared to experimental results.
    Original languageEnglish
    Title of host publicationProceedings of the 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics
    PublisherAmerican Nuclear Society
    Publication date2019
    Pages5342-5355
    ISBN (Electronic)9781510893450
    Publication statusPublished - 2019
    Event18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics - Portland, United States
    Duration: 18 Aug 201923 Aug 2019
    Conference number: 18

    Conference

    Conference18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics
    Number18
    CountryUnited States
    CityPortland
    Period18/08/201923/08/2019

    Cite this

    Nalbandyan, A., Klinkby, E. B., Lauritzen, B., Groth-Jensen, J., & Steyn, R. (2019). Coupled Neutronics/Thermal Hydraulics Assessment of Graphite Moderated Molten Salt Reactors. In Proceedings of the 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (pp. 5342-5355). American Nuclear Society.