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  1. No Access

    Article

    Selecting Burnup Algorithms in OpenMC Using the Calculated Benchmark of LEU Assembly and MOX Fuel

    OpenMC is a state-of-the-art Monte Carlo neutron transport simulation code that uses the Python programming language as an API. OpenMC supports eight burnup simulation algorithms. This study presents the resul...

    H. A. Tanash, D. A. Solovyov, V. G. Zimin, A. L. Lobarev… in Physics of Atomic Nuclei (2023)

  2. No Access

    Article

    Method for Determining the Time to Attaining the Minimum Controlled Power Level for VVER

    Reactor startup to the minimum controlled level (MCL) of power is one of the most hazardous nuclear operations in the service life of the reactor. Currently, codes for neutronic calculations such as reactor si...

    A. I. Al-Shamayleh, D. A. Solovyov, A. A. Semyonov… in Physics of Atomic Nuclei (2022)