Publication
This paper presents nuclear reactor transient measurements with thermal feedback and corresponding coupled neutronic-thermohydraulic simulations performed for the natural circulation cooled Training Reactor of Budapest University of Technology and Economics. During the experiments, the outlet temperatures of several fuel assemblies were measured with thermocouples, while the reactor power was simultaneously logged by an ex-core neutron detector. The numerical simulations were carried out with the internally coupled TRACE/Point Kinetics and TRACE/PARCS code systems. The parameterised group constant library of the PARCS neutronic model, as well as the reactivity coefficients of the Point Kinetic model, were generated by the Serpent 2 Monte Carlo code. In the former case, special attention was given to the applied diffusion coefficients. The 3D/1D TRACE calculation model was prepared with the SNAP graphical interface. It was shown that after applying the necessary corrections, related to the peculiar natural circulation flow and reactor parameters, the measurement and both simulation results showed good agreement in terms of peak nominal power and average assembly outlet temperature. Although several anomalies are presented and modelling details that need improvement are highlighted, this study suggests that the used code systems could be appropriate tools to reproduce and analyse transients performed in small-core, high-leakage, natural circulation cooled pool-type reactors.