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.

Monte Carlo method, nuclear reactor physics, OpenMC, neutron transport, HDF5, parallel computing, Python API, fuel depletion, CRAM, DAGMC.

Romano P.K., Forget B. "The OpenMC Monte Carlo particle transport code". Annals of Nuclear Energy. 2013. 51: 274–281.

Romano P.K., Horelik N.E., Herman B.R., et al. "OpenMC: A state-of-the-art Monte Carlo code for research and development in nuclear engineering". Annals of Nuclear Energy. 2015. 82: 90–97.

Pusa M. "Rational approximations to the matrix exponential in burnup calculations". Nuclear Science and Engineering. 2011. 169(2): 155–167.

OpenMC Documentation https://docs.openmc.org.