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

• “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])

• “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-33” intended for fast reactor criticality benchmarks. It’s derived as a subset of VITAMIN‑J

• “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”, “LLNL-616”, “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

openmc.mgxs.EnergyGroups

An energy group structure used for multigroup cross-sections.

Multi-group Cross Sections

openmc.mgxs.MGXS	An abstract multi-group cross section for some energy group structure
openmc.mgxs.MatrixMGXS	An abstract multi-group cross section for some energy group structure
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.Current	A current multi-group cross section.
openmc.mgxs.DiffusionCoefficient	A diffusion coefficient multi-group cross section.
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.ReducedAbsorptionXS	A reduced absorption 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.
openmc.mgxs.ArbitraryXS	A multi-group cross section for an arbitrary reaction type.
openmc.mgxs.ArbitraryMatrixXS	A multi-group matrix cross section for an arbitrary reaction type.
openmc.mgxs.MeshSurfaceMGXS	An abstract multi-group cross section for some energy group structure on the surfaces of a mesh domain.

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.MatrixMDGXS	An abstract multi-delayed-group cross section for some energy group and delayed group structure 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.