The 3D Core Thermohydraulics and Neutronics Solution in the TRAB-SMABRE Accident and Transient Code

Jaakko Miettinen, Hanna Räty, Antti Daavittila

    Research output: Chapter in Book/Report/Conference proceedingConference article in proceedingsScientific

    Abstract

    The TRAB-SMABRE code is a result of code development efforts carried out at VTT Processes in Finland for calculating transient and accident behaviour in Finnish LWR plants.
    The operating plants are two 770 MWe BWR units in Olkiluoto and two 500 MWe PWR units of VVER-440 type in Loviisa. In addition a new PWR plant of the 1600 MWe EPR type will be built into Olkiluoto. The TRAB core model, two group neutronics solution with nodal expansion method, has been initially developed as a transient 3D transient code for the BWR plant transients, and HEXTRAN code for the 3D VVER transients having the hexagonal fuel geometry.
    The SMABRE model thermohydraulic model is a drift flux based LOCA model. HEXTRAN and SMABRE were coupled in parallel for making the ATWS analyses in VVER plants possible and TRAB and SMABRE were coupled in parallel for calculating the transients in the PWR plants with the squared core array. As the new step the TRAB core model was coupled internally with the SMABRE for making possible the BWR analyses with the flow reversal possible. and as an optional tool for the PWR plant analyses with the squared core array. The model predicts well the 3D core thermohydraulics with the encapsulated fuel, but not for the open PWR core.
    To overcome this deficiency a thermohydraulics simulation model for the core was introduced based on the 3D porous media thermohydraulics solution PORFLO. In the paper the basic equations of the 3D neutronics, SMABRE thermohydraulics and PORFLO thermohydraulics are described. For the BWR plant the calculation results using the parallel coupling of neutronics and thermohydraulics and internal coupling will be compared for the two transients, MSIV closure in the steam line and partial load reduction, both compared against the real plant data. The calculation result proves that the internal coupling gives the most extensive possibilities for the core simulation and is recommended for the further BWR analyses For the PWR plant the control rod withdrawal results are compared for three different thermohydrualic solution of the PWR core, parallel coupling without cross-flow, internal coupling without cross flow and internal coupling with cross flow.
    The comparison proves that the internal coupling is the best solution even for the PWR transient and accident analyses.
    If the power differences between neighbouring fuel bundles are large, the core thermohydraulics need to be calculated by considering the 3D cross-flow as well.
    Original languageEnglish
    Title of host publicationProceedings
    Subtitle of host publication15th International Conference on Nuclear Engineering, ICONE15
    PublisherJapan Society of Mechanical Engineers
    Number of pages7
    Publication statusPublished - 2007
    MoE publication typeB3 Non-refereed article in conference proceedings
    Event15th International Conference on Nuclear Engineering, ICONE15 - Nagoya, Aichi, Japan
    Duration: 22 Apr 200726 Apr 2007

    Conference

    Conference15th International Conference on Nuclear Engineering, ICONE15
    Abbreviated titleICONE15
    Country/TerritoryJapan
    CityNagoya, Aichi
    Period22/04/0726/04/07

    Keywords

    • LWR reactor dynamics
    • thermohydraulic-neutronics coupling

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