Zobrazeno 1 - 10
of 12
pro vyhledávání: '"Justin M. Pounders"'
Publikováno v:
Annals of Nuclear Energy. 132:404-412
Currently, most reactor simulation codes use predefined constant or variable time steps for simulation, which is inefficient. Available adaptive time step methods in the area of neutron transport and reactor kinetics are rather based on multistep and
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::4899185c03867e9fd490f0bd1d4b5e37
https://doi.org/10.1016/j.anucene.2019.05.003
https://doi.org/10.1016/j.anucene.2019.05.003
Autor:
J. Boffie, Justin M. Pounders
Publikováno v:
Annals of Nuclear Energy. 116:280-289
The stability and accuracy of an adaptive time step control scheme are analyzed for the transient diffusion equation. This scheme is based on the commonly-implemented backward difference discretization of the diffusion equation and recommends optimal
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::e864ac079a403e35d27a87adecc6c87b
https://doi.org/10.1016/j.anucene.2018.02.044
https://doi.org/10.1016/j.anucene.2018.02.044
Publikováno v:
Journal of Computational and Theoretical Transport. 46:20-45
We improve the convergence properties of cellwise block iteration for discrete-ordinates radiation-transport calculations by adapting it for use as a smoother within a multigrid method. Cellwise bl...
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::7157b84fc56fd16cf2b736b2dfa06fcc
https://doi.org/10.1080/23324309.2016.1161650
https://doi.org/10.1080/23324309.2016.1161650
Publikováno v:
Nuclear Engineering and Design. 363:110620
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::0c980f475901fd380b9634fdc5a7a897
https://doi.org/10.1016/j.nucengdes.2020.110620
https://doi.org/10.1016/j.nucengdes.2020.110620
Publikováno v:
Nuclear Engineering and Design. 293:16-22
A method is detailed for the stochastic generation of multigroup cross sections. This paper describes a solid theoretical framework for development of multigroup (discrete energy) nuclear data from continuous energy evaluated nuclear data files by Mo
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::b86e63ed9394633c654c36170a308872
https://doi.org/10.1016/j.nucengdes.2015.07.027
https://doi.org/10.1016/j.nucengdes.2015.07.027
Autor:
Justin M. Pounders, Farzad Rahnema
Publikováno v:
Nuclear Science and Engineering. 176:273-291
A new solution technique is derived for the time-dependent transport equation. This approach extends the steady-state coarse-mesh transport method that is based on global-local decompositions of la...
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::53cd2d05afe2758d0d0e92e1bb4a0e3a
https://doi.org/10.13182/nse12-108
https://doi.org/10.13182/nse12-108
Publikováno v:
Annals of Nuclear Energy. 38:2024-2078
An 8-group cross section library is provided to augment a previously published 2-group 3D stylized half-core Canadian deuterium uranium (CANDU) reactor benchmark problem. Reference eigenvalues and selected pin and bundle fission rates are also includ
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::2f3dc6deb2b92618dfb134b06b171ab0
https://doi.org/10.1016/j.anucene.2011.03.011
https://doi.org/10.1016/j.anucene.2011.03.011
Publikováno v:
Annals of Nuclear Energy. 38:1172-1185
To assess the accuracy of diffusion or transport methods for reactor calculations, it is desirable to create heterogeneous benchmark problems that are typical of whole core configurations. In this paper we have created two and three dimensional numer
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::fa90737aebdf57b428a362aef3ad6345
https://doi.org/10.1016/j.anucene.2010.11.020
https://doi.org/10.1016/j.anucene.2010.11.020
Publikováno v:
Annals of Nuclear Energy. 38:876-896
A 3D stylized half-core Canadian deuterium uranium (CANDU) reactor benchmark problem is presented. The benchmark problem is comprised of a heterogeneous lattice of 37-element natural uranium fuel bundles, heavy water moderated, heavy water cooled, wi
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::a34507dca86476eaf843e32aebb60102
https://doi.org/10.1016/j.anucene.2010.10.018
https://doi.org/10.1016/j.anucene.2010.10.018
Autor:
Justin M. Pounders, Farzad Rahnema
Publikováno v:
Annals of Nuclear Energy. 37:1595-1600
An efficient response-based solution to the time-dependent neutron transport equation in a semi-infinite slab is derived. The solution is based on polynomial expansions of the source terms and neutron flux in the time domain. The expansion coefficien
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::b27838a0741f3c3983d8f55423a8af11
https://doi.org/10.1016/j.anucene.2010.03.010
https://doi.org/10.1016/j.anucene.2010.03.010