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Computational thermal hydraulic schemes for SFR transient studies

Abstract : Design and safety studies for Sodium cooled Fast Reactors (SFR) include thermal-hydraulic transient calculations. A wide range of transients must be studied, protected or unprotected, symmetrical or dissymmetrical. Calculations include the core and primary circuit with or without secondary circuits and with or without connected circuits, particularly the decay heat removal systems. The modeling and calculation schemes to be implemented are determined based on the physical phenomena, functionalities and the validation field of the various codes used and the parameters of interest in the different parts of the reactor.The codes used by the CEA for these studies are presented CATHARE (system code), TRIO_U (Computational Fluid Dynamics code), TRIO_U MC (core sub-assembly code). This paper focuses on core modeling, circuit modeling and the use of codes separately, chained or coupled.The selected calculation schemes for the following transients are discussed-Loss of supply station power, -Unprotected loss of flow -Unprotected Loss of Heat Sinks -Failure of a pump diagrid connection pipe.Furthermore, the subjects of code validation and uncertainties associated with the input data are touched upon.
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  • HAL Id : cea-02509147, version 1

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M.-S. Chenaud, S. Li, M. Anderhuber, L. Matteo, A. Gerschenfeld. Computational thermal hydraulic schemes for SFR transient studies. NURETH 16 - The 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Aug 2015, Chicago, United States. ⟨cea-02509147⟩

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