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Title: SENSMG: First-Order Sensitivities of Neutron Reaction Rates, Reaction-Rate Ratios, Leakage, keff , and α Using PARTISN

Abstract

SENSMG is a tool for calculating the first-order sensitivities of reaction-rate ratios, keff, and α in critical problems and reaction-rate ratios, reaction rates, and leakage in fixed-source problems to multigroup cross sections, isotope densities, material mass densities, and interface locations using the PARTISN multigroup discrete-ordinates code by implementing Generalized Perturbation Theory. SENSMG can be used for one-dimensional spherical and slab (r) and two-dimensional cylindrical (r-z) geometries. For fixed-source (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 cross-section availability. Several verification problems are presented in which SENSMG results are compared with MCNP6, SCALE6.2, and direct perturbations (central differences).

Authors:
 [1]
  1. 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):
LA-UR-18-22532
Journal ID: ISSN 0029-5639
Grant/Contract Number:  
AC52-06NA25396
Resource Type:
Accepted Manuscript
Journal Name:
Nuclear Science and Engineering
Additional Journal Information:
Journal Volume: 192; Journal Issue: 1; Journal ID: ISSN 0029-5639
Publisher:
American Nuclear Society - Taylor & Francis
Country of Publication:
United States
Language:
English
Subject:
73 NUCLEAR PHYSICS AND RADIATION PHYSICS; multigroup; discrete ordinates; adjoint; first-order sensitivity analysis

Citation Formats

Favorite, Jeffrey A. SENSMG: First-Order Sensitivities of Neutron Reaction Rates, Reaction-Rate Ratios, Leakage, keff , and α Using PARTISN. United States: N. p., 2018. Web. doi:10.1080/00295639.2018.1471296.
Favorite, Jeffrey A. SENSMG: First-Order Sensitivities of Neutron Reaction Rates, Reaction-Rate Ratios, Leakage, keff , and α Using PARTISN. United States. doi:10.1080/00295639.2018.1471296.
Favorite, Jeffrey A. Mon . "SENSMG: First-Order Sensitivities of Neutron Reaction Rates, Reaction-Rate 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: First-Order Sensitivities of Neutron Reaction Rates, Reaction-Rate Ratios, Leakage, keff , and α Using PARTISN},
author = {Favorite, Jeffrey A.},
abstractNote = {SENSMG is a tool for calculating the first-order sensitivities of reaction-rate ratios, keff, and α in critical problems and reaction-rate ratios, reaction rates, and leakage in fixed-source problems to multigroup cross sections, isotope densities, material mass densities, and interface locations using the PARTISN multigroup discrete-ordinates code by implementing Generalized Perturbation Theory. SENSMG can be used for one-dimensional spherical and slab (r) and two-dimensional cylindrical (r-z) geometries. For fixed-source (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 cross-section 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}
}

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Works referenced in this record:

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journal, March 2016

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Using the Levenberg-Marquardt Method for Solutions of Inverse Transport Problems in One- and Two-Dimensional Geometries
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Computational Evaluation of Neutron Multiplicity Measurements of Polyethylene-Reflected Plutonium Metal
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    Works referencing / citing this record:

    Application of Neutron Multiplicity Counting Experiments to Optimal Cross-Section Adjustments
    journal, February 2020

    • Clark, Alexander R.; Mattingly, John; Favorite, Jeffrey A.
    • Nuclear Science and Engineering, Vol. 194, Issue 4
    • DOI: 10.1080/00295639.2019.1698267