Experimental investigation of the pressure loss characteristics of the full-scale MYRRHA fuel bundle in the COMPLOT LBE facility

    Research outputpeer-review


    MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK•CEN, the Belgian Nuclear Research Center. MYRRHA is a pool-type reactor with Lead Bismuth Eutectic (LBE) as primary coolant. Conceived as an accelerator driven system prototype, it is able to operate in sub-critical mode. Operating in critical mode, MYRRHA is identified as the European Technology Pilot Plant for the Lead Cooled Fast Reactor which is one of the Generation IV reactor concepts.
    The MYRRHA fuel assembly (FA) contains a hexagonal bundle of 127 cylindrical fuel pins surrounded by a hexagonal shroud or wrapper. The upper and lower ends of the shroud are connected to the inlet and outlet nozzles guiding the LBE coolant through the FA. Helical wire-spacers wound on the outer surface of each fuel pin, keep the fuel pins separated from one another in the bundle.
    A full-scale mock-up of the MYRRHA FA was constructed and installed in the COMPLOT LBE experimental test facility at SCK-CEN, for the purpose of measuring the axial pressure drop across the assembly. Pressure measurements were taken at various axial positions within a single edge subchannel, using small tappings through the wall of the hexagonal housing, axially spaced in multiples of the wire-wrapper pitch. Additional pressure tappings were placed at the same axial height on different hexagonal walls, to investigate the variation of pressure in one cross-section of the fuel assembly and verify the presence of local maximum and minimum pressures due to the characteristic spiralling pressure and flow field induced by the wire-wrapped bundle assembly.
    The paper will report on the experimental results. These are used to support the design and development of the MYRRHA reactor and are also compared with existing pressure drop correlations as well as numerical results from codes such as CFD.
    Original languageEnglish
    Title of host publication16th International Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16)
    Number of pages13
    ISBN (Electronic)978-0-89448-722-4
    StatePublished - 1 Sep 2015
    Event2015 - NURETH : 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics - ANS, IAEA, University of Pisa, Chicago
    Duration: 30 Aug 20154 Sep 2015


    Conference2015 - NURETH
    Country/TerritoryUnited States

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