Implementation of hybrid simulation schemes in COBAYA3/SUBCHANFLOW coupled codes for the efficient direct prediction of local safety parameters

The precise prediction of power generation, heat transfer and flow distribution within a reactor core is of great importance to asses the safety features of any reactor design. The necessity to better describe the most important safety related physical phenomena prevailing in LWRs drive the extensions of current neutronic (N)/thermal-hydraulic (TH) coupled methodologies. Nowadays, several computer codes that solve the time dependent neutron diffusion or transport equations are coupled with TH codes at nodal level. This coarse spatial discretization of both N and TH does not allow direct prediction of local phenomena at pin or subchannel levels. Moreover, pin by pin simulations are currently performed using different strategies and methodologies. The main drawback of these approaches is the considerable computational time needed when addressing whole core solutions. Consequently, new fast running and accurate approaches are needed to simulate reactor cores using multi physics and multi scale methodologies. This type of analysis includes for instance, the use of mixed nodal based solutions with pin level solutions for both N and TH. This paper discusses a methodology implemented to achieve coupled N/TH simulations based on hybrid schemes. First, an overview of the state of the art involving non-conform geometry is presented, followed with the description of the codes used for this purpose and their extensions to perform hybrid simulations. Results for the coupled N/TH scheme are presented for a full size PWR core in steady state.