Specifications for a coupled neutronics thermal-hydraulics SFR test case
Coupling neutronics/thermal-hydraulics calculations for the design of nuclear reactors is a growing trend in the scientific community. This approach allows to properly represent the mutual feedbacks between the neutronic distribution and the thermal-hydraulics properties of the materials composing the reactor, details which are often lost when separate analysis are performed. In this work, a test case for a generation IV sodium-cooled fast reactor (SFR), based on the ASTRID concept developed by CEA, is proposed.