OPENMC: ARCHITECTURAL EVOLUTION AND NEW HORIZONS IN STOCHASTIC RADIATION TRANSPORT MODELING

This paper provides a comprehensive analysis of the OpenMC software suite – a next-generation tool for solving the neutral particle transport equation using the Monte Carlo method with continuous-energy data. The transition from traditional imperative input formats to object-oriented experiment programming via the Python API is examined. Particular attention is paid to the mathematical formalism of the solved equations (Boltzmann transport equation, eigenvalue problem, depletion kinetics), the implementation of hybrid parallelism (MPI/OpenMP), and the efficiency of handling nuclear data in HDF5 format. The role of the code's open architecture in addressing scientific verification challenges, CAD system integration, and multiphysics modeling of nuclear power plants is discussed.