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Numerical analysis of core thermal-hydraulic for sodium-cooled fast reactors

Abstract : The paper presents the numerical analysis of core thermal-hydraulic performed by CEA for Sodium-cooled Fast Reactors (SFR). The core thermal-hydraulic analyses are performed at three scales The individual sub-assembly, characterized by pin bundle with helical space wire. In the 1980s, a specific sub-channel scale model for SFR-type subassemblies was developed at CEA. In 2008, this model was re-implemented in the Trio_U CFD code, under the name Trio_U MC (Core Model), for use in design phase. Besides this sub-channel model, refined CFD models of the sub-assembly were also developed. Some examples (sodium or clad temperature inside the bundle) are presented.The whole sub-assemblies.The main objective is to determine the core steady-state hydraulic conditions, for specific objectives, such as obtaining a required mean outlet core temperature while keeping the maximum cladding temperature within a given limit in practice, this is achieved by allocating the fuel sub-assemblies among a number of flow-rate zones. The paper describes the methodology to determine the number and flow zones allocation and the corresponding mass flow rates with Trio_U MC, as well as the associated optimization process. The whole core, with sub-assemblies, inter-wrapper gaps and the hot pool plenum.A model for the interaction of the core sub-assemblies with the adjoining inter-wrapper gaps and with the hot pool plenum has been developed at CEA. Known as Trio_U MC2, it consists in a code coupling of Trio_U MC with a Trio_U CFD domain representing the inter-wrapper gaps and hot pool plenum. This model leads to tube temperatures in the reactor nominal state.
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Submitted on : Monday, February 24, 2020 - 2:49:00 PM
Last modification on : Tuesday, April 28, 2020 - 11:28:17 AM
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  • HAL Id : cea-02489500, version 1

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A. Conti, A. Gerschenfeld, Y. Gorsse, T. Cadiou, R. Lavastre. Numerical analysis of core thermal-hydraulic for sodium-cooled fast reactors. NURETH 2015 - 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Aug 2015, Chicago, United States. ⟨cea-02489500⟩

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