其他摘要：This work carries out a calculation of the pressure drop of the coolant flow past spacer grids in a preliminary design of a nuclear fuel assembly of a modular, integral PWR. In this small modular, integral reactor the coolant flows along the core driven by natural circulation. The analysis will focus on considering a cross section of 1/12 of the entire fuel element despite a single asymmetry and an axial segment. A 3D CFD simulation is performed to estimate the pressure drop during steady state flow rate of single-phase light water at constant temperature. Bundle cross-flows are disregarded as a first approximation. Appropriate boundary conditions are applied at fuel pin walls and symmetry planes, namely outlet absolute pressure and mass flow rate at inlet that are kept constant. Results presented in non-dimensional, normalized way show the expected behavior. However, due to modelling hypothesis based on a limited knowledge of spring geometrical details, the results cannot be considered useful for design optimization purposes.
