OpenMC is a Monte Carlo particle transport simulation code focused on neutron criticality calculations. It is capable of simulating 3D models based on constructive solid geometry with second-order surfaces. The particle interaction data is based on ACE format cross sections, also used in the MCNP and Serpent Monte Carlo codes. OpenMC was originally developed by members of the Computational Reactor Physics Group at the Massachusetts Institute of Technology starting in 2011. Various universities, laboratories, and other organizations now contribute to the development of OpenMC. For more information on OpenMC, feel free to send a message to the User’s Group mailing list.

References in zbMATH (referenced in 16 articles )

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  1. Cosgrove, P.; Shwageraus, E.: Stability analysis of higher-order neutronics-depletion coupling schemes and Bateman operators (2022)
  2. Dominesey, Kurt A.; Ji, Wei: Reduced-order modeling of neutron transport separated in energy by proper generalized decomposition with applications to nuclear reactor physics (2022)
  3. Liu, Minyun; Ma, Yugao; Guo, Xiaoyu; Liu, Shichang; Liu, Guodong; Huang, Shanfang; Wang, Kan: An improved tracking method for particle transport Monte Carlo simulations (2021)
  4. Jarrett, Michael G.; Kochunas, Brendan M.; Larsen, Edward W.; Downar, Thomas J.: (\textSP_3) limit of the 2D/1D transport equations with varying degrees of angular coupling (2020)
  5. Zhou, Xiafeng; Guo, Zhaoli: Discrete unified gas kinetic scheme for steady multiscale neutron transport (2020)
  6. Bassett, Brody; Kiedrowski, Brian: Meshless local Petrov-Galerkin solution of the neutron transport equation with streamline-upwind Petrov-Galerkin stabilization (2019)
  7. Cai, X.-X.; Kittelmann, T.; Klinkby, E.; Márquez Damián, J. I.: Rejection-based sampling of inelastic neutron scattering (2019)
  8. Sweezy, Jeremy E.: A Monte Carlo volumetric-ray-casting estimator for global fluence tallies on GPUs (2018)
  9. Barcellos, L. F. F. Chaves; Bodmann, B. E. J.; Leite, S. Q. Bogado; Vilhena, M. T.: On a continuous energy Monte Carlo simulator for neutron transport: optimisation with fission, intermediate and thermal distributions (2017)
  10. Josey, C.; Forget, B.; Smith, K.: High order methods for the integration of the Bateman equations and other problems of the form of (y^\prime=F(y,t)y) (2017)
  11. Tramm, John R.; Smith, Kord S.; Forget, Benoit; Siegel, Andrew R.: The random ray method for neutral particle transport (2017)
  12. Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; Johnson, Seth R.; Pandya, Tara M.; Godfrey, Andrew T.: Hot zero power reactor calculations using the Insilico code (2016)
  13. Keady, Kendra P.; Larsen, Edward W.: Stability of Monte Carlo (k)-eigenvalue simulations with CMFD feedback (2016)
  14. Romano, Paul K.: An algorithm for generating random variates from the Madland-Nix fission energy spectrum (2015)
  15. Roberts, Jeremy A.; Forget, Benoit: Multigroup diffusion preconditioners for multiplying fixed-source transport problems (2014)
  16. Romano, Paul K.; Siegel, Andrew R.; Forget, Benoit; Smith, Kord: Data decomposition of Monte Carlo particle transport simulations via tally servers (2013)

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