openmc.mgxs
– MultiGroup Cross Section Generation¶
Energy Groups¶
Module Variables¶
 openmc.mgxs.GROUP_STRUCTURES¶
Dictionary of commonly used energy group structures:
“CASMOX” (where X is 2, 4, 8, 16, 25, 40 or 70) from the CASMO lattice physics code
“SHEM361” designed for LWR analysis to eliminate selfshielding calculations of thermal resonances ([HFA2005], [SAN2007], [HEB2008])
“SCALEX” (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])
“MPACTX” (where X is 51 (PWR), 60 (BWR), 69 (Magnox)) from the MPACT reactor physics code ([KIM2019] and [KIM2020])
“ECCO1968” designed for fine group reactor cell calculations for fast, intermediate and thermal reactor applications ([SAR1990])
activation energy group structures “VITAMINJ42”, “VITAMINJ175”, “TRIPOLI315”, “CCFE709” and “UKAEA1102”
 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: VITAMINJ, ECCO33, ECCO2000 and XMAS JEF/DOC315 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 1215, 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. 446452.
 HEB2008
Hébert, Alain & Santamarina, Alain. (2008). Refinement of the SantamarinaHfaiedh energy mesh between 22.5 eV and 11.4 keV. International Conference on the Physics of Reactors 2008, PHYSOR 08. 2. 929938.
 ZAL1999
K. Záleský and L. Marková (1999), Assessment of Nuclear Data Needs for BroadGroup 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. 4658.
 KIM2020
Kim, K.S., Ade, B., & Luciano, N. (2020). Development of the MPACT 69group Library for Magnox Reactor Analysis using VERA. Proceedings of International Conference on Physics of Reactors PHYSOR2020.
Classes¶
An energy group structure used for multigroup crosssections. 
Multigroup Cross Sections¶
An abstract multigroup cross section for some energy group structure within some spatial domain. 

An abstract multigroup cross section for some energy group structure within some spatial domain. 

An absorption multigroup cross section. 

A capture multigroup cross section. 

The fission spectrum. 

A current multigroup cross section. 

A diffusion coefficient multigroup cross section. 

A fission multigroup cross section. 

An inverse velocity multigroup cross section. 

A recoverable fission energy production rate multigroup cross section. 

The scattering multiplicity matrix. 

A fission production matrix multigroup cross section. 

A reduced absorption multigroup cross section. 

A scattering multigroup cross section. 

A scattering matrix multigroup cross section with the cosine of the changeinangle represented as one or more Legendre moments or a histogram. 

The grouptogroup scattering probability matrix. 

A total multigroup cross section. 

A transportcorrected total multigroup cross section. 

A multigroup cross section for an arbitrary reaction type. 

A multigroup matrix cross section for an arbitrary reaction type. 

An abstract multigroup cross section for some energy group structure on the surfaces of a mesh domain. 
Multidelayedgroup Cross Sections¶
An abstract multidelayedgroup cross section for some energy and delayed group structures within some spatial domain. 

An abstract multidelayedgroup cross section for some energy group and delayed group structure within some spatial domain. 

The delayed fission spectrum. 

A fission delayed neutron production multigroup cross section. 

A fission delayed neutron production matrix multigroup cross section. 

The delayed neutron fraction. 

The decay rate for delayed neutron precursors. 
Multigroup Cross Section Libraries¶
A multienergygroup and multidelayedgroup cross section library for some energy group structure. 