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

The paper presents the numerical analysis of core thermal hydraulics performed by CEA for Sodiumcooled Fast Reactors (SFR). This core thermal-hydraulic analysis is performed at three scales: Individual sub-assemblies, characterized by its triangular-lattice pin bundle with helical spacer wires: In the 1980s, a specific subchannel 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 studies. Besides this subchannel model, refined CFD models of the subassembly were also developed. Some examples (sodium or clad temperature inside the bundle) are presented. Complete core: The nominal thermal-hydraulic behavior of the complete core must be optimized 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 subassemblies 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. This whole-core subchannel model can also be used to determine the peak cladding temperature reached during transients. 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 can be used to determine hex-can temperatures in the nominal state; it addition, it can also be used to predict the effect of inter-wrapper flows on decay heat removal during loss-offlow accidental transients.