Accurate modelling of nuclear reactors is fundamental for benchmarking and code development. This research presents developments regarding the modelling of the AKR-2 zero-power reactor located at the Dresden University of Technology (TU Dresden) in Germany. The reactor model was written using the Monte Carlo (MC) codes Serpent and MCNP. The code-to-code verification was performed using stochastic volume simulations, graphical comparison of slice plots and criticality calculations. The validation of the model is done through comparison with results from a series of experiments: critical control rod configurations, the radial distribution of the thermal neutron flux in the central experimental channel, the critical experiment, reactivity curves of the control rods, and the fast spectrum neutron flux at the largest experimental channel. The model reproduces qualitatively and mostly also quantitatively all of the experimental results, representing first step towards establishing an experimental platform for nuclear data measurements and code validation. However, the discrepancies for the central radial neutron flux and control rod curves highlight the need to investigate deeply the geometrical and material data of the reactor. Finally, the overview of the current research ongoing at the AKR-2 and the future steps in the modelling of the AKR-2 are outlined.