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-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

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 within some spatial domain.

openmc.mgxs.MatrixMGXS

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.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.