V. G. Zimin

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 results of choosing an integration method for modeling the burnup of fuel assemblies with burnable poisons for VVER-1000 reactors. Burnup simulation results from OpenMC were compared with those reported in the OECD benchmark. Eight different numerical integrators can be used to model burnup in the OpenMC code: PI, CE/CM, LE/QI, CE/LI, CF4, EPC-RK4, SI-CE/LI, SI-LE/QI. The test results showed that the SI-CE/LI and SI-LE/QI integrators require significantly more time to calculate one burnup step than the others with the same accuracy, so they were excluded from further consideration. The PI integrator showed low integration accuracy at the same burnup steps with other integrators. However, PI has a high performance compared to other integrators, and as the integration step decreases, it converges to one solution, which can be chosen as a reference for assessing the quality of other integrators. On the basis of the results obtained using the fine step PI integrator, it was decided to use the CE/LI integrator for further work. The results obtained with CE/LI were compared with those obtained with the VVER-1000 LEU and MOX benchmark for the codes MCU, TVS-M, WIMS8A, HELIOS, and MULTICELL and showed good agreement. Thus, we can conclude the applicability of the CE/LI integrator as part of OpenMC for modeling the burnup of fuel assemblies containing burnable poisons. During the work, the resources of the high-performance computer center of the National Research Nuclear University MEPhI were used.

The neutron kinetics equations belong to the class of stiff equations for numerical time integration schemes. In this work, the accuracy and speed of algorithms based on backward differentiation...