HAL will be down for maintenance from Friday, June 10 at 4pm through Monday, June 13 at 9am. More information
Skip to Main content Skip to Navigation
Conference papers

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.
Complete list of metadata

Cited literature [5 references]  Display  Hide  Download

Contributor : Amplexor Amplexor Connect in order to contact the contributor
Submitted on : Monday, February 24, 2020 - 2:49:00 PM
Last modification on : Tuesday, April 28, 2020 - 11:28:17 AM
Long-term archiving on: : Monday, May 25, 2020 - 6:25:41 PM


Files produced by the author(s)


  • HAL Id : cea-02489500, version 1




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⟩



Record views


Files downloads