This article presents the first results of the neutronics analysis of RBMK-1000 at the 3D assembly level using the OpenMC code as an early stage of the long-term large academic research project. The goal of the large academic research project is to develop a Multi-group Monte Carlo computational module around the OpenMC code using Python programming language. The results of this paper include the criticality calculation and depletion analysis of RBMK-1000. The effective multiplication factor value (keff) has been determined. In depletion calculation, the fission reaction rates of U-235, U-236, U-238, production of fissionable materials, and two fission product poisons Xe-135 and Sm-149 have been investigated. There is ongoing work for benchmarking the first results from OpenMC against Serpent 2 Monte Carlo code. The parametric design data of RBMK-1000 have been taken from the OECD - Nuclear Energy Agency (NEA) SFCOMPO 2.0 (Spent Fuel Isotopic Composition) database that includes measured isotopic concentrations of spent nuclear fuel, with operational histories and design data. The case study was the Leningrad-1 NPP. The ENDF/B-VII.1 data library has been used in this analysis. Moreover, this is the first time to date that the OpenMC code has been used in neutronic studies of the RBMK reactors.