1996
DOI: 10.13182/nse124-369
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A Hybrid Multigroup/Continuous-Energy Monte Carlo Method for Solving the Boltzmann-Fokker-Planck Equation

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Cited by 19 publications
(7 citation statements)
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“…๐œ‡ s = photon scattering coefficient [cm -1 ] ๐‘… pf = Photon gain rate term from photofission reactions [cm -3 s -1 eV -1 ] ๐‘„ p = photon source distribution (includes bremsstrahlung and neutron fission gammas) [cm -3 s -1 sr -1 eV -1 ] ๐œ“ n = neutron angular flux [cm -2 s -1 sr -1 eV -1 ] ๐›ด t = total macroscopic neutron cross section [cm -1 ] ๐›ด s = scattering macroscopic neutron cross section [cm -1 ] ๐œ’ = Neutron fission energy spectrum ๐œˆ n = average number of neutrons produced per fission ๐›ด f = fission macroscopic neutron cross section [cm -1 ] ๐œ™ n = Neutron scalar flux [cm -2 s -1 eV -1 ] ๐‘„ n = neutron source distribution (includes photoneutrons) [cm -3 s -1 sr -1 eV -1 ]. (Morel 1996, Wang 2018, Duderstadt and Hamilton 1976 The photofission term was extrapolated to be similar to a neutron fission term. The boundary conditions for the multiparticle (electron-photon-neutron) transport problem solved by MCNP are vacuum ๐œ“ e ๐‘Ÿ,๐ธ,ฮฉ = ๐ฟ ๐‘Ÿ,ฮฉ = ๐œ“ n ๐‘Ÿ,๐ธ,ฮฉ = 0 for ฮฉ โ€ข e s < 0 for all points r s on boundary surface S B .…”
Section: Mcnp Modelmentioning
confidence: 99%
“…๐œ‡ s = photon scattering coefficient [cm -1 ] ๐‘… pf = Photon gain rate term from photofission reactions [cm -3 s -1 eV -1 ] ๐‘„ p = photon source distribution (includes bremsstrahlung and neutron fission gammas) [cm -3 s -1 sr -1 eV -1 ] ๐œ“ n = neutron angular flux [cm -2 s -1 sr -1 eV -1 ] ๐›ด t = total macroscopic neutron cross section [cm -1 ] ๐›ด s = scattering macroscopic neutron cross section [cm -1 ] ๐œ’ = Neutron fission energy spectrum ๐œˆ n = average number of neutrons produced per fission ๐›ด f = fission macroscopic neutron cross section [cm -1 ] ๐œ™ n = Neutron scalar flux [cm -2 s -1 eV -1 ] ๐‘„ n = neutron source distribution (includes photoneutrons) [cm -3 s -1 sr -1 eV -1 ]. (Morel 1996, Wang 2018, Duderstadt and Hamilton 1976 The photofission term was extrapolated to be similar to a neutron fission term. The boundary conditions for the multiparticle (electron-photon-neutron) transport problem solved by MCNP are vacuum ๐œ“ e ๐‘Ÿ,๐ธ,ฮฉ = ๐ฟ ๐‘Ÿ,ฮฉ = ๐œ“ n ๐‘Ÿ,๐ธ,ฮฉ = 0 for ฮฉ โ€ข e s < 0 for all points r s on boundary surface S B .…”
Section: Mcnp Modelmentioning
confidence: 99%
“…To generate cross-section tables for electron/photon transport problems that will use the multigroup Boltzmann-Fokker-Planck algorithm, 44 the CEPXS 47 code developed by Sandia National Laboratory and available from RSICC can be used. The CEPXS manuals describe the algorithms and physics database upon which the code is based; the physics package is essentially the same as ITS version 2.1.…”
Section: F Multigroup Tablesmentioning
confidence: 99%
“…That work makes use of recursion relationships described by Morel (Morel, 1979) to generate modified moments of the weight function defined by the selected cross section model. These modified moments are then used as inputs for the algorithm developed by Sloan (1983), and extended in work by Morel et al (1996), for generating discrete cross section values.…”
Section: Generation Of Discrete Differential Cross Sectionsmentioning
confidence: 99%