A full-core coupled neutronic/thermal-hydraulic code for the modeling of lead-cooled nuclear fast reactors

• First neutronic/thermal-hydraulic full-core modeling of lead-cooled fast reactor with hexagonal fuel elements.
• Separation of time scales allows fast 3D problem solution by 1D + 2D models.
• Space and time numerical convergence demonstrated.
• Accuracy demonstrated by benchmark against analytical solutions.

A new multi-physics simulation tool – FRENETIC (Fast REactor NEutronics/Thermal-hydraulICs) – is presented for the quasi-3D analysis of a lead-cooled fast reactor core with the hexagonal fuel element configuration, as currently proposed within the framework of the European project LEADER. The tool implements coupled neutronic (NE) and thermal-hydraulic (TH) models. In the NE module, a 2D + 1D full-core multi-group diffusion solver has been developed based on a coarse-mesh nodal scheme and adapted to cope with the hexagonal geometry. In the TH module, the hexagonal elements, described by 1D (axial) transient advection and conduction in the coolant coupled to conduction in the fuel pins, are thermally coupled to each other in the transverse directions to obtain the full-core evolution of the distribution of the TH variables. The NE and TH modules are coupled at each TH time step by transferring to the TH module the distribution of the power source computed by the NE module, which is the driver of the TH evolution; alternately, the temperature distribution computed by the TH module is input to the NE module, in order to update the cross sections. The code is benchmarked against pure TH and pure NE analytical solutions and the results of a coupled NE/TH pseudo-transient (criticality search) are also presented. The convergence of the numerical solution is demonstrated both in space and time by computational simulations.