openmc.mgxs
– Multi-Group Cross Section Generation¶
Energy Groups¶
Module Variables¶
- openmc.mgxs.GROUP_STRUCTURES¶
Dictionary of commonly used energy group structures:
“CASMO-X” (where X is 2, 4, 8, 16, 25, 40 or 70) from the CASMO lattice physics code
“SHEM-361” designed for LWR analysis to eliminate self-shielding calculations of thermal resonances ([HFA2005], [SAN2007], [HEB2008])
“SCALE-X” (where X is 44 which is designed for criticality analysis and 252 is designed for thermal reactors) for the SCALE code suite ([ZAL1999] and [REARDEN2013])
“MPACT-X” (where X is 51 (PWR), 60 (BWR), 69 (Magnox)) from the MPACT reactor physics code ([KIM2019] and [KIM2020])
“ECCO-1968” designed for fine group reactor cell calculations for fast, intermediate and thermal reactor applications ([SAR1990])
activation energy group structures “VITAMIN-J-42”, “VITAMIN-J-175”, “TRIPOLI-315”, “CCFE-709” and “UKAEA-1102”
- SAR1990(1,2)
Sartori, E., OECD/NEA Data Bank: Standard Energy Group Structures of Cross Section Libraries for Reactor Shielding, Reactor Cell and Fusion Neutronics Applications: VITAMIN-J, ECCO-33, ECCO-2000 and XMAS JEF/DOC-315 Revision 3 - DRAFT (December 11, 1990).
- SAN2004
Santamarina, A., Collignon, C., & Garat, C. (2004). French calculation schemes for light water reactor analysis. United States: American Nuclear Society - ANS.
- HFA2005
Hfaiedh, N. & Santamarina, A., “Determination of the Optimized SHEM Mesh for Neutron Transport Calculations,” Proc. Top. Mtg. in Mathematics & Computations, Supercomputing, Reactor Physics and Nuclear and Biological Applications, September 12-15, Avignon, France, 2005.
- SAN2007
Santamarina, A. & Hfaiedh, N. (2007). The SHEM energy mesh for accurate fuel depletion and BUC calculations. Proceedings of the International Conference on Safety Criticality ICNC 2007, St Peterburg (Russia), Vol. I pp. 446-452.
- HEB2008
Hébert, Alain & Santamarina, Alain. (2008). Refinement of the Santamarina-Hfaiedh energy mesh between 22.5 eV and 11.4 keV. International Conference on the Physics of Reactors 2008, PHYSOR 08. 2. 929-938.
- ZAL1999
K. Záleský and L. Marková (1999), Assessment of Nuclear Data Needs for Broad-Group SCALE Library Related to VVER Spent Fuel Applications, IAEA. SCALE44.
- REARDEN2013
B. T. Rearden, M. E. Dunn, D. Wiarda, C. Celik, K. Bekar, M. L. Williams, D. E. Peplow, M. A. Jessee, C. M. Perfetti, I. C. Gauld, W. A. Wieselquist, J. P. Lefebvre, R. A. Lefebvre, W. J. Marshall, A. B. Thompson, F. Havluj, S. E. Skutnik, K. J. Dugan. (2013). Overview of SCALE 6.2. OECD. SCALE252.
- KIM2019
Kim, K.S., Williams, M., Wiarda, D., & Clarno, K. (2019). Development of the multigroup cross section library for the CASL neutronics simulator MPACT: Method and procedure. Annals of Nuclear Energy, 133. pp. 46-58.
- KIM2020
Kim, K.S., Ade, B., & Luciano, N. (2020). Development of the MPACT 69-group Library for Magnox Reactor Analysis using VERA. Proceedings of International Conference on Physics of Reactors PHYSOR2020.
Classes¶
An energy group structure used for multigroup cross-sections. |
Multi-group Cross Sections¶
An abstract multi-group cross section for some energy group structure within some spatial domain. |
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An abstract multi-group cross section for some energy group structure within some spatial domain. |
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An absorption multi-group cross section. |
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A capture multi-group cross section. |
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The fission spectrum. |
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A current multi-group cross section. |
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A diffusion coefficient multi-group cross section. |
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A fission multi-group cross section. |
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An inverse velocity multi-group cross section. |
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A recoverable fission energy production rate multi-group cross section. |
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The scattering multiplicity matrix. |
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A fission production matrix multi-group cross section. |
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A reduced absorption multi-group cross section. |
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A scattering multi-group cross section. |
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A scattering matrix multi-group cross section with the cosine of the change-in-angle represented as one or more Legendre moments or a histogram. |
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The group-to-group scattering probability matrix. |
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A total multi-group cross section. |
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A transport-corrected total multi-group cross section. |
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A multi-group cross section for an arbitrary reaction type. |
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A multi-group matrix cross section for an arbitrary reaction type. |
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An abstract multi-group cross section for some energy group structure on the surfaces of a mesh domain. |
Multi-delayed-group Cross Sections¶
An abstract multi-delayed-group cross section for some energy and delayed group structures within some spatial domain. |
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An abstract multi-delayed-group cross section for some energy group and delayed group structure within some spatial domain. |
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The delayed fission spectrum. |
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A fission delayed neutron production multi-group cross section. |
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A fission delayed neutron production matrix multi-group cross section. |
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The delayed neutron fraction. |
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The decay rate for delayed neutron precursors. |
Multi-group Cross Section Libraries¶
A multi-energy-group and multi-delayed-group cross section library for some energy group structure. |