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Comparison of CATHARE code using a 3D Reactor Pressure Vessel modelling approach and experimental results on intermediate break LOCAs of ROSA 2 program

Abstract : The CATHARE code is a thermal-hydraulic code employed by several countries in the world as reference tool to study incidental and accidental transients in nuclear power plants. The CEA (Commissariat à l’Energie Atomique et aux Energies renouvelables Alternatives), in collaboration with its French nuclear partners (AREVA NP, EDF and IRSN) in France, develops and validates this code to increase its predictive capabilities. A particular attention is focused today to validate the 3D model for the entire Reactor Pressure Vessel component during a IBLOCA transient. Several 3D phenomena take place in this element and they cannot be realistically predicted using a simple 1D/0D approach. Asymmetrical liquid flow repartitions, and counter current phenomena are some examples. The OECD/ROSA-2 program is here chosen to validate and demonstrate the enhanced prediction capability of CATHARE tool by modelling the entire Reactor Pressure Vessel of the Japanese LSTF facility using one only 3D component. Tests 1, 2 and 7 are selected to this work, corresponding to different intermediate break LOCAs (13% and 17% sizes into hot and cold legs). Comparison with experimental results shows an improvement of global system response by using a 3D approach for the entire Reactor Pressure Vessel instead of a corresponding 1D/0D. In particular the upper plenum and the core voiding are very well predicted and the global pressure trend is better foreseen. The same trend is observed for the three tests. This study shows the important role of the 3D phenomena prediction to the understanding of global system response. A clear improvement using a 3D approach to represent the entire Reactor Pressure Vessel suggests the necessity to follow this way and to carry on with the 3D method to study the integral systems.
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  • HAL Id : cea-02438710, version 1

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S. Carnevali, P. Bazin. Comparison of CATHARE code using a 3D Reactor Pressure Vessel modelling approach and experimental results on intermediate break LOCAs of ROSA 2 program. NUTHOS-11 - The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, Oct 2016, Gyeongju, South Korea. ⟨cea-02438710⟩

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