Izvestiya vuzov. Yadernaya Energetika

The peer-reviewed scientific and technology journal. ISSN: 0204-3327

Cross-Verification of the OpenMC Code Based on Benchmark Problems for High- and Low-Enriched Core with IRT-3M Type Fuel Assemblies

9/17/2026 2026 - #03 Modelling processes at nuclear facilities

Umarov Sh.B. Tikhomirov G.V.

DOI: https://doi.org/10.26583/npe.2026.3.14

UDC: 621.039.52

A cross-verification of the OpenMC code was performed to assess its applicability for neutronic analysis and computational support of the WWR-SM research reactor undergoing conversion to high-density uranium–molybdenum fuel. Due to the absence of experimental data for advanced fuel types currently under development, the verification was carried out using specially designed test problems representing simplified yet representative core configurations with IRT-3M fuel assemblies used in the WWR-SM reactor. OpenMC calculations employing the ENDF/B-VIII.0 and ENDF/B-VII.1 evaluated nuclear data libraries were compared against MCNP/ENDF/B-VII results taken as reference. The study presents comparative evaluations of the effective neutron multiplication factor, integral reaction rates, and fuel depletion parameters, along with explanations of the primary sources of discrepancies between the codes. Based on the conducted cross-verification, the applicability of OpenMC for computational support of the WWR-SM research reactor during its conversion to high-density uranium–molybdenum fuel has been substantiated.

References

  1. Reduced Enrichment for Research and Test Reactors (RERTR) Program. U.S. Department of Energy, National Nuclear Security Administration, Argonne National Laboratory. URL: https://rertr.nse.anl.gov/ (accessed Nov. 23, 2025).
  2. Newton T.H. Jr., Wilson E.H., Bergeron A., Horelik N., Stevens J.G. Neutronic analyses for HEU to LEU fuel conversion of the Massachusetts Institute of Technology MITR reactor. Argonne National Laboratory, MIT Nuclear Reactor Laboratory Publ., 2010, 33 p.
  3. Jamison L., Kim Y., Hofman G., Shehee T., Karay N. Research and Test Reactor Fuels. U.S. Department of Energy, National Nuclear Security Administration, Office of Conversion, Material Management and Minimization Program Publ., 2022, 92 p.
  4. Fernandes A.C., Santos J.P., Marques J.G., Kling A., Ramos A.R., Barradas N.P. Validation of the Monte Carlo model supporting core conversion of the Portuguese Research Reactor for neutron fluence rate determinations. Annals of Nuclear Energy. 2010;37(9):1139–1145. DOI: https://doi.org/10.1016/j.anucene.2010.05.004
  5. Holmström S., Baumeister B., Wight J. Qualification of advanced LEU fuels for high-power research reactor conversion designs. EPJ Nuclear Sciences & Technologies. 2025;11:18. DOI: https://doi.org/10.1051/epjn/2025012
  6. Kim Y.S., Hofman G.L. Improved performance of U–Mo dispersion fuel by Si addition in Al matrix. Argonne National Laboratory Publ., 2011, 104 p.
  7. Cherepnin Yu.S., Pulinets A.A., Kravtsova O.A., Simonidi A.R., Chizhova E.S., Larionov I.A., Logvenchev S.I. Development of FA with High-Density Fuel for Foreign Research Nuclear Reactors of Russian Design. Atomic Energy. 2021;129:234–236. DOI: https://doi.org/10.1007/s10512-021-00740-9
  8. Daydas S., Tiftikci A. Neutronic examination of the U–Mo accident tolerant fuel for VVER-1200 reactors. Nuclear Engineering and Technology. 2024;56(7):2625–2632. DOI: https://doi.org/10.1016/j.net.2024.02.021
  9. Savva P., Varvayanni M., Fernandes A.C., Marques J.G., Catsaros N. Comparing neutronics codes performance in analyzing a fresh-fuelled research reactor core. Annals of Nuclear Energy. 2014;63:731–741. DOI: https://doi.org/10.1016/j.anucene.2013.08.036
  10. Ivakhin S.V., Radaev A.I., Tikhomirov G.V., Shchurovskaya M.V. Fuel burnup calculations of an IRT research reactor for conversion to LEU. Izvestia vuzov. Yadernaya Energetika. 2012;3:49–59. URL: https://nuclear-power-engineering.ru/article/2012/03/11/ (accessed Aug. 26, 2026) (in Russian).
  11. Research Reactor Benchmarking Database: Facility Specification and Experimental Data. IAEA Technical Reports Series No. 480. Vienna, International Atomic Energy Agency Publ., 2022, 489 p., ISBN 978-92-0-134822-4.
  12. Research reactor core conversion from the use of highly enriched uranium to the use of low enriched uranium fuels. Vienna, International Atomic Energy Agency Publ., 1980, 710 p.
  13. Alferov V.P., Radaev A.I., Shchurovskaya M.V., Tikhomirov G.V., Hanan N.A., van Heerden F.A. Comparative validation of Monte Carlo codes for the conversion of a research reactor. Annals of Nuclear Energy. 2015;77:273–280. DOI: https://doi.org/10.1016/j.anucene.2014.11.032
  14. Zizin M.N. Test neutron-physical calculations of nuclear reactors. Moscow, Atomizdat Publ., 1980, 88 p. (in Russian).
  15. Ivakhin S.V., Radaev A.I., Tikhomirov G.V., Shchurovskaya M.V. Test problems for an IRT-type reactor with LEU U–Mo fuel. Preprint. Moscow, MEPhI Publ., 2013, 91 p. (in Russian).
  16. Ivakhin S.V., Tikhomirov G.V., Shchurovskaya M.V., Radaev A.I. Development of benchmark problems for IRT research reactor core neutronic calculations. Collection of Scientific Papers of MEPhI. Moscow, MEPhI Publ., 2012, pp. 102–108 (in Russian).
  17. OpenMC User’s Guide. URL: https://docs.openmc.org/en/stable/usersguide/index.html (accessed 10.10.2025).
  18. Guo H., Huo X., Feng K., Gu H. Verification of OpenMC for fast reactor physics analysis with China experimental fast reactor start-up tests. Nuclear Engineering and Technology. 2022;54(10):3897–3908. DOI: https://doi.org/10.1016/j.net.2022.05.021
  19. Plotnikov D.A., Romanenko V.I., Solovyev D.A., Zimin V.G., Molev I.A., Kudratov V.N., Ivakhin S.V., Djarum B. Application of OpenMC to benchmark a VVER-1000 with mixed U/MOX core. Global Nuclear Safety. 2025;15(3):26–35. DOI: https://doi.org/10.26583/gns-2025-03-03 (in Russian).
  20. Albugami A.O., Alomari A.S., Almarshad A.I. Modeling and simulation of VERA core physics benchmark using OpenMC code. Nuclear Engineering and Technology. 2023;55(12):3388–3400. DOI: https://doi.org/10.1016/j.net.2023.05.036

IRT-3M OpenMC cross-verification neutron multiplication coefficient fuel burnout research reactor WWR-SM

Link for citing the article: Umarov Sh.B., Tikhomirov G.V. Cross-Verification of the OpenMC Code Based on Benchmark Problems for High- and Low-Enriched Core with IRT-3M Type Fuel Assemblies. Izvestiya vuzov. Yadernaya Energetika. 2026, no. 3, pp. 196-206; DOI: https://doi.org/10.26583/npe.2026.3.14 (in Russian).

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