2023
DOI: 10.2172/1909545
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MCNP<sup>®</sup> Code V.6.3.0 Release Notes

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Cited by 8 publications
(3 citation statements)
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“…The output results are evaluated through nuclear data libraries, taking account of low-energy interaction probabilities [76]. The last version, called MCNP6, is clearly described by Rising et al [77]. The code considers new neutron-induced fission systems and isotopes like 240,242 Pu and 252 Cf.…”
Section: Mcpnx Codementioning
confidence: 99%
“…The output results are evaluated through nuclear data libraries, taking account of low-energy interaction probabilities [76]. The last version, called MCNP6, is clearly described by Rising et al [77]. The code considers new neutron-induced fission systems and isotopes like 240,242 Pu and 252 Cf.…”
Section: Mcpnx Codementioning
confidence: 99%
“…Monte Carlo (MC) simulation is a powerful method for modeling the behavior of particles in various physical systems, including electron-solid interactions [21][22][23]25,26]. Some software packages have been specifically developed for this purpose, such as Geant4 [27] and Monte Carlo N-Particle Transport Codes (MCNP) [28]. These packages provide libraries and tools for defining complex material geometries, specifying interaction physics, and performing detailed particle transport and interaction simulations.…”
Section: Monte Carlo Simulationsmentioning
confidence: 99%
“…Many traditionally developed neutron-transport codes are export-controlled (e.g. MCNP (Rising et al, 2023), Shift (Hamilton & Evans, 2019), and MCATK (Adams et al, 2015)) and some are known to be difficult to install, use, and develop in. MC/DC is open-source, and thus, similar to other open-source Monte Carlo neutron-transport codes (e.g., OpenMC (Romano et al, 2015)), it promotes knowledge sharing, collaboration, and inclusive, community-driven development.…”
Section: Statement Of Needmentioning
confidence: 99%