| Article ID | Journal | Published Year | Pages | File Type |
|---|---|---|---|---|
| 297013 | Nuclear Engineering and Design | 2012 | 9 Pages |
The accuracy and the degree of spatial resolution of safety studies, required for new reactor concepts, imply the use of coupled 3D neutronic and 3D thermal hydraulic codes. Tools to perform the coupling between neutronic codes both deterministic and stochastic with plant or sub-channel codes are being developed worldwide. With the increase of computational resources, Monte Carlo codes like MCNPX are acquiring much more relevance. They are able to obtain results without major approximations in the geometry and with point-wise cross section representation. This paper describes the development of a coupled neutronics/thermal-hydraulics code system based on Monte Carlo code MCNPX and the sub-channel code COBRA-IV. In the current work the temperature dependence of nuclear data is handled with the pseudo material approach and based on JEFF 3.1 data libraries compiled with NJOY. The code has been applied to a sodium fast reactor (SFR) concept at both fuel assembly and full core scale. This is the first step toward a more comprehensive tool that takes into account more phenomena and feedback effects.
► Neutronics and thermal-hydraulics coupling with MCNPX and COBRA-IV. ► We study the pseudo-material approach in fast reactors. ► We carry out a coupled calculation in a SFR fuel assembly at pin-by-pin level. ► SFR full core analysis with fuel assemblies grouped by radial rings. ► The keff difference is 200 pcm between the MCNPX stand alone calculation and the coupled solution.
