Thermal-hydraulic and neutronic codes coupling for the analysis of a Lead Fast Reactor

Abstract

In this work the thermal-hydraulics and neutronics behavior of a Lead Fast Reactor (LFR) core is investigated evaluating the power generation distribution taking into account the local temperature field. The temperature field is evaluated using the CFD finite element code FEMuS and exchanged with the multiscale neutron code DONJON-DRAGON, which interpolates the macroscopic cross-sections according to the local temperature field and local lead density distribution. As a result, the neutron flux changes and defines a new power density distribution which is used to update the temperature field into the CFD code. The coupling between neutron and CFD codes is achieved through their inclusion into the numerical platform SALOME. The numerical libraries MED, included into the SALOME platform, are used to exchange data run-time between FEMuS and DONJON.