SENSMG: FirstOrder Sensitivities of Neutron Reaction Rates, ReactionRate Ratios, Leakage, k_{eff} , and α Using PARTISN
Abstract
SENSMG is a tool for calculating the firstorder sensitivities of reactionrate ratios, k_{eff}, and α in critical problems and reactionrate ratios, reaction rates, and leakage in fixedsource problems to multigroup cross sections, isotope densities, material mass densities, and interface locations using the PARTISN multigroup discreteordinates code by implementing Generalized Perturbation Theory. SENSMG can be used for onedimensional spherical and slab (r) and twodimensional cylindrical (rz) geometries. For fixedsource (leakage) problems, SENSMG relies on the MISC and/or SOURCES4C codes to compute neutron source rate densities from spontaneous fission and (α,n) sources. SENSMG is a combination of Python and Fortran and was developed under Linux. This computer code abstract describes all user inputs, the input file, and output files. Furthermore, this computer code abstract describes how SENSMG can be modified to support different computer platforms, PARTISN versions, or crosssection availability. Several verification problems are presented in which SENSMG results are compared with MCNP6, SCALE6.2, and direct perturbations (central differences).
 Authors:

 Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
 Publication Date:
 Research Org.:
 Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
 Sponsoring Org.:
 USDOE National Nuclear Security Administration (NNSA)
 OSTI Identifier:
 1467217
 Report Number(s):
 LAUR1822532
Journal ID: ISSN 00295639
 Grant/Contract Number:
 AC5206NA25396
 Resource Type:
 Accepted Manuscript
 Journal Name:
 Nuclear Science and Engineering
 Additional Journal Information:
 Journal Volume: 192; Journal Issue: 1; Journal ID: ISSN 00295639
 Publisher:
 American Nuclear Society  Taylor & Francis
 Country of Publication:
 United States
 Language:
 English
 Subject:
 73 NUCLEAR PHYSICS AND RADIATION PHYSICS; multigroup; discrete ordinates; adjoint; firstorder sensitivity analysis
Citation Formats
Favorite, Jeffrey A. SENSMG: FirstOrder Sensitivities of Neutron Reaction Rates, ReactionRate Ratios, Leakage, keff , and α Using PARTISN. United States: N. p., 2018.
Web. doi:10.1080/00295639.2018.1471296.
Favorite, Jeffrey A. SENSMG: FirstOrder Sensitivities of Neutron Reaction Rates, ReactionRate Ratios, Leakage, keff , and α Using PARTISN. United States. doi:10.1080/00295639.2018.1471296.
Favorite, Jeffrey A. Mon .
"SENSMG: FirstOrder Sensitivities of Neutron Reaction Rates, ReactionRate Ratios, Leakage, keff , and α Using PARTISN". United States. doi:10.1080/00295639.2018.1471296. https://www.osti.gov/servlets/purl/1467217.
@article{osti_1467217,
title = {SENSMG: FirstOrder Sensitivities of Neutron Reaction Rates, ReactionRate Ratios, Leakage, keff , and α Using PARTISN},
author = {Favorite, Jeffrey A.},
abstractNote = {SENSMG is a tool for calculating the firstorder sensitivities of reactionrate ratios, keff, and α in critical problems and reactionrate ratios, reaction rates, and leakage in fixedsource problems to multigroup cross sections, isotope densities, material mass densities, and interface locations using the PARTISN multigroup discreteordinates code by implementing Generalized Perturbation Theory. SENSMG can be used for onedimensional spherical and slab (r) and twodimensional cylindrical (rz) geometries. For fixedsource (leakage) problems, SENSMG relies on the MISC and/or SOURCES4C codes to compute neutron source rate densities from spontaneous fission and (α,n) sources. SENSMG is a combination of Python and Fortran and was developed under Linux. This computer code abstract describes all user inputs, the input file, and output files. Furthermore, this computer code abstract describes how SENSMG can be modified to support different computer platforms, PARTISN versions, or crosssection availability. Several verification problems are presented in which SENSMG results are compared with MCNP6, SCALE6.2, and direct perturbations (central differences).},
doi = {10.1080/00295639.2018.1471296},
journal = {Nuclear Science and Engineering},
number = 1,
volume = 192,
place = {United States},
year = {2018},
month = {7}
}
Web of Science
Works referenced in this record:
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 Favorite, Jeffrey A.
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 Evans, R. T.; Cacuci, D. G.
 Nuclear Science and Engineering, Vol. 172, Issue 2
Using the LevenbergMarquardt Method for Solutions of Inverse Transport Problems in One and TwoDimensional Geometries
journal, October 2011
 Bledsoe, Keith C.; Favorite, Jeffrey A.; Aldemir, Tunc
 Nuclear Technology, Vol. 176, Issue 1
Computational Evaluation of Neutron Multiplicity Measurements of PolyethyleneReflected Plutonium Metal
journal, February 2014
 Miller, E. C.; Mattingly, J. K.; Clarke, S. D.
 Nuclear Science and Engineering, Vol. 176, Issue 2
Works referencing / citing this record:
Application of Neutron Multiplicity Counting Experiments to Optimal CrossSection Adjustments
journal, February 2020
 Clark, Alexander R.; Mattingly, John; Favorite, Jeffrey A.
 Nuclear Science and Engineering, Vol. 194, Issue 4