A Serpent/OpenFOAM coupling approach for the study of the fuel burnup

In nuclear reactors, fuel burnup calculations assess the time evolution of both the fuel composition and the reactivity during the in-core reactor irradiation. Furthermore, this analysis is also relevant for the fuel management and for determining the amount of long-lived radionuclides in spent nuclear fuel. The development of accurate modelling approaches is recommended to investigate the coupling between the thermal-hydraulics and the fuel burnup. In this paper, a multi-physics approach that solves the neutron transport problem and the time evolution of the nuclide concentration, and concurrently describes the heat transfer between the fuel and the coolant is presented. The goal is achieved by coupling the Monte Carlo code Serpent for neutronics/fuel burnup modelling with the OpenFOAM toolkit for thermalhydraulics description thanks to an iterative procedure. The adopted approach was tested on a case study of a simplified fuel cell, composed of an UO2 pin surrounded by water.