openmc.mgxs – Multi-Group Cross Section Generation¶
Energy Groups¶
Module Variables¶
-
openmc.mgxs.GROUP_STRUCTURES¶ - “CASMO-X” (where X is 2, 4, 8, 16, 25, 40 or 70) from the CASMO lattice physics code
- “XMAS-172” designed for LWR analysis ([SAR1990], [SAN2004])
- “SHEM-361” designed for LWR analysis to eliminate self-shielding calculations of thermal resonances ([HFA2005], [SAN2007], [HEB2008])
- activation energy group structures “VITAMIN-J-175”, “TRIPOLI-315”, “CCFE-709” and “UKAEA-1102”
[SAR1990] 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. Type: Dictionary of commonly used energy group structures
Classes¶
openmc.mgxs.EnergyGroups |
An energy groups structure used for multi-group cross-sections. |
Multi-group Cross Sections¶
openmc.mgxs.MGXS |
An abstract multi-group cross section for some energy group structure within some spatial domain. |
openmc.mgxs.AbsorptionXS |
An absorption multi-group cross section. |
openmc.mgxs.CaptureXS |
A capture multi-group cross section. |
openmc.mgxs.Chi |
The fission spectrum. |
openmc.mgxs.FissionXS |
A fission multi-group cross section. |
openmc.mgxs.InverseVelocity |
An inverse velocity multi-group cross section. |
openmc.mgxs.KappaFissionXS |
A recoverable fission energy production rate multi-group cross section. |
openmc.mgxs.MultiplicityMatrixXS |
The scattering multiplicity matrix. |
openmc.mgxs.NuFissionMatrixXS |
A fission production matrix multi-group cross section. |
openmc.mgxs.ScatterXS |
A scattering multi-group cross section. |
openmc.mgxs.ScatterMatrixXS |
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. |
openmc.mgxs.ScatterProbabilityMatrix |
The group-to-group scattering probability matrix. |
openmc.mgxs.TotalXS |
A total multi-group cross section. |
openmc.mgxs.TransportXS |
A transport-corrected total multi-group cross section. |
Multi-delayed-group Cross Sections¶
openmc.mgxs.MDGXS |
An abstract multi-delayed-group cross section for some energy and delayed group structures within some spatial domain. |
openmc.mgxs.ChiDelayed |
The delayed fission spectrum. |
openmc.mgxs.DelayedNuFissionXS |
A fission delayed neutron production multi-group cross section. |
openmc.mgxs.DelayedNuFissionMatrixXS |
A fission delayed neutron production matrix multi-group cross section. |
openmc.mgxs.Beta |
The delayed neutron fraction. |
openmc.mgxs.DecayRate |
The decay rate for delayed neutron precursors. |
Multi-group Cross Section Libraries¶
openmc.mgxs.Library |
A multi-energy-group and multi-delayed-group cross section library for some energy group structure. |