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All issues Volume 302 (2024) EPJ Web Conf., 302 (2024) 02003 References
Open Access
Issue
EPJ Web Conf.
Volume 302, 2024
Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo (SNA + MC 2024)
Article Number 02003
Number of page(s) 10
Section Deterministic Transport Codes: Algorithms, HPC & GPU
DOI https://doi.org/10.1051/epjconf/202430202003
Published online 15 October 2024
  1. Nikitin E., Fridman E., Mikityuk K. Solution of the OECD/NEA neutronic SFR benchmark with Serpent-DYN3D and Serpent-PARCS code systems[J]. Annals of Nuclear Energy, 2015, 75:492–497. [Google Scholar]
  2. Lin C.-S., Yang W. S. An assessment of the applicability of multigroup cross sections generated with Monte Carlo method for fast reactor analysis[J]. Nuclear Engineering and Technology, 2020, 52(12):2733–2742. [CrossRef] [Google Scholar]
  3. Tran T. Q., Cherezov A., Du X., Lee D. Verification of a two-step code system MCS/RAST-F to fast reactor core analysis[J]. Nuclear Engineering and Technology, 2021, 54(5):1789–1803. [Google Scholar]
  4. Nguyen T. D. C., Lee H., Lee D. Use of Monte Carlo code MCS for multigroup cross section generation for fast reactor analysis[J]. Nuclear Engineering and Technology, 2021, 53(9):2788–2802. [CrossRef] [Google Scholar]
  5. Martin N., Stewart R., Bays S. A multiphysics model of the versatile test reactor based on the MOOSE framework[J]. Annals of Nuclear Energy, 2022, 172:109066. [Google Scholar]
  6. Guo H., Wu Y., Song Q., Cong T., Gu H. Development of OpenMC/Trivac two-step scheme for fast reactor core neutronics analysis[J]. Annals of Nuclear Energy, 2023, 190. [Google Scholar]
  7. Guo H., Wu Y.-W., Song Q.-F., Shen Y.-Y., Gu H.-Y. Development of multi-group Monte-Carlo transport and depletion coupling calculation method and verification with metal-fueled fast reactor[J]. Nuclear Science and Techniques, 2023, 34(11):163. [CrossRef] [Google Scholar]
  8. Wu Y., Song Q., Feng K., Vidal J.-F., Gu H., Guo H. Multigroup cross-sections generated using Monte-Carlo method with flux-moment homogenization technique for fast reactor analysis[J]. Nuclear Engineering and Technology, 2023. [Google Scholar]
  9. Wu Y., Song Q., Wang R., Xiao Y., Gu H., Guo H. Development and verification of a Monte Carlo two-step method for lead-based fast reactor neutronics analysis[J]. Nuclear Engineering and Technology, 2023. [Google Scholar]
  10. Nikitin E., Fridman E. Extension of the reactor dynamics code DYN3D to SFR applications - Part II: Validation against the Phenix EOL control rod withdrawal tests[J]. Annals of Nuclear Energy, 2018, 119:411–418. [CrossRef] [Google Scholar]
  11. Ponomarev A., Mikityuk K., Zhang L., Nikitin E., Fridman E., Álvarez-Velarde F., Romojaro Otero P., Jiménez-Carrascosa A., García-Herranz N., Lindley B., Baker U., Seubert A., Henry R. Superphénix Benchmark Part I: Results of Static Neutronics[J]. Journal of Nuclear Engineering and Radiation Science, 2022, 8(1):011320. [CrossRef] [Google Scholar]
  12. Nikitin E., Fridman E. Modeling of the FFTF isothermal physics tests with the Serpent and DYN3D codes[J]. Annals of Nuclear Energy, 2019, 132:679–685. [CrossRef] [Google Scholar]
  13. Tran T. Q. Neutronic simulation of the CEFR experiments with the nodal diffusion code system RAST-F[J]. Nuclear Engineering and Technology, 2022: 15. [Google Scholar]
  14. Nguyen T. D. C., Tran T. Q., Lee D. Coupled neutronics/thermal-hydraulic analysis of ANTS-100e using MCS/RAST-F two-step code system[J]. Nuclear Engineering and Technology, 2023. [Google Scholar]
  15. Guo H., Feng K., Wu Y., Jin X., Huo X., Gu H. Preliminary verification of multi-group cross-sections generation and locally heterogeneous transport calculation using OpenMC with CEFR start-up tests benchmark[J]. Progress in Nuclear Energy, 2022, 154(104484). [Google Scholar]
  16. Boyd W., Shaner S., Li L., Forget B., Smith K. The OpenMOC method of characteristics neutral particle transport code[J]. Annals of Nuclear Energy, 2014, 68:43–52. [Google Scholar]
  17. Gunow G., Forget B., Smith K. Full core 3D simulation of the BEAVRS benchmark with OpenMOC[J]. Annals of Nuclear Energy, 2019, 134:299–304. [CrossRef] [Google Scholar]
  18. Leppänen J., Pusa M., Fridman E. Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code[J]. Annals of Nuclear Energy, 2016, 96:126–136. [CrossRef] [Google Scholar]
  19. Goorley T., James M., Booth T., Brown F., Bull J., Cox L. J., Durkee J., Elson J., Fensin M., Forster R. A., Hendricks J., Hughes H. G., Johns R., Kiedrowski B., Martz R., Mashnik S., McKinney G., Pelowitz D., Prael R., Sweezy J., Waters L., Wilcox T., Zukaitis T. Taylor & Francis, 2012. Initial MCNP6 Release Overview[J]. Nuclear Technology, 2012, 180(3):298–315. [Google Scholar]
  20. Park H. J., Shim H., Joo H., Kim C. Qualification test of few group constants generated from an MC method by the two-step neutronics analysis system McCARD/MASTER[J]. undefined, 2011. [Google Scholar]
  21. Wang K., Li Z., She D., Liang J., Xu Q., Qiu Y., Yu J., Sun J., Fan X., Yu G. RMC - A Monte Carlo code for reactor core analysis[J]. Annals of Nuclear Energy, 2015, 82:121–129. [Google Scholar]
  22. Boyd W., Nelson A., Romano P. K., Shaner S., Forget B., Smith K. Multigroup CrossSection Generation with the OpenMC Monte Carlo Particle Transport Code[J]. Nuclear Technology, 2019, 205(7):928–944. [CrossRef] [Google Scholar]
  23. Qin S., Li Y., He Q., Cao L., Wang Y., Wu Y., Wu H. Homogenized cross-section generation for pebble-bed type high-temperature gas-cooled reactor using NECP- MCX[J]. Nuclear Engineering and Technology, 2023, 55(9):3450–3463. [CrossRef] [Google Scholar]
  24. OECD/NEA. OECD/NEA, 2016. Benchmark for neutronic analysis of sodium- cooled fast reactor cores with various fuel types and core sizes[R]. NEA/NSC/R(2015)9, 2016. [Google Scholar]
  25. Vidal J.-F., Archier P., Calloo A., Jacquet P., Tommasi J. An improved energycollapsing method for core-reflector modelization in sfr core calculations using the paris platform[C]//PHYSOR 2012., 2012 Knoxville, Tennessee, USA: 15. [Google Scholar]
  26. Vidal J.-F., Archier P., Faure B., Jouault V., Palau J.-M., Pascal V., Rimpault G., Auffret F., Graziano L., Masiello E., Santandrea S. APOLLO3® homogenization techniques for transport core calculations - application to the ASTRID CFV core[J]. Nuclear Engineering and Technology, 2017, 49:1379–1387. [CrossRef] [Google Scholar]
  27. Qiu R., Ma X., Xu Q., Liu J., Chen Y. American Society of Mechanical Engineers, 2017. Development and Verification of Multi-Group Cross Section Process Code TXMAT for Fast Reactor RBEC-M Analysis[C]//Volume 3: Nuclear Fuel and Material, Reactor Physics and Transport Theory; Innovative Nuclear Power Plant Design and New Technology Application., 2017 Shanghai, China: V003T02A015. [Google Scholar]

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