SCALE Continuous-Energy Eigenvalue Sensitivity Coefficient Calculations

Perfetti, Christopher M.; Rearden, Bradley T.; Martin, William R.

doi:10.13182/NSE15-12

Title: SCALE Continuous-Energy Eigenvalue Sensitivity Coefficient Calculations

Abstract

Sensitivity coefficients describe the fractional change in a system response that is induced by changes to system parameters and nuclear data. The Tools for Sensitivity and UNcertainty Analysis Methodology Implementation (TSUNAMI) code within the SCALE code system makes use of eigenvalue sensitivity coefficients for an extensive number of criticality safety applications, including quantifying the data-induced uncertainty in the eigenvalue of critical systems, assessing the neutronic similarity between different critical systems, and guiding nuclear data adjustment studies. The need to model geometrically complex systems with improved fidelity and the desire to extend TSUNAMI analysis to advanced applications has motivated the development of a methodology for calculating sensitivity coefficients in continuous-energy (CE) Monte Carlo applications. The Contributon-Linked eigenvalue sensitivity/Uncertainty estimation via Tracklength importance CHaracterization (CLUTCH) and Iterated Fission Probability (IFP) eigenvalue sensitivity methods were recently implemented in the CE-KENO framework of the SCALE code system to enable TSUNAMI-3D to perform eigenvalue sensitivity calculations using continuous-energy Monte Carlo methods. This work provides a detailed description of the theory behind the CLUTCH method and describes in detail its implementation. This work explores the improvements in eigenvalue sensitivity coefficient accuracy that can be gained through the use of continuous-energy sensitivity methods and also compares severalmore »« less

Authors:

Perfetti, Christopher M. ^[1]; Rearden, Bradley T. ^[1]; Martin, William R. ^[2]

Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Univ. of Michigan, Ann Arbor, MI (United States)

Publication Date:: Thu Feb 25 00:00:00 EST 2016

Research Org.:: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Sponsoring Org.:: USDOE National Nuclear Security Administration (NNSA)

OSTI Identifier:: 1246770

Grant/Contract Number:: AC05-00OR22725

Resource Type:: Accepted Manuscript

Journal Name:: Nuclear Science and Engineering

Additional Journal Information:: Journal Volume: 182; Journal Issue: 3; Journal ID: ISSN 0029-5639

Publisher:: American Nuclear Society

Country of Publication:: United States

Language:: English

Subject:: 46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND TECHNOLOGY; TSUNAMI; eigenvalue sensitivity coefficients; Monte Carlo

Citation Formats


                    Perfetti, Christopher M., Rearden, Bradley T., and Martin, William R. SCALE Continuous-Energy Eigenvalue Sensitivity Coefficient Calculations.  United States: N. p., 2016. 
Web.  doi:10.13182/NSE15-12.

Copy to clipboard


                    Perfetti, Christopher M., Rearden, Bradley T., & Martin, William R. SCALE Continuous-Energy Eigenvalue Sensitivity Coefficient Calculations.  United States.  https://doi.org/10.13182/NSE15-12

Copy to clipboard


                    Perfetti, Christopher M., Rearden, Bradley T., and Martin, William R. Thu .  
"SCALE Continuous-Energy Eigenvalue Sensitivity Coefficient Calculations".  United States.  https://doi.org/10.13182/NSE15-12.  https://www.osti.gov/servlets/purl/1246770.

Copy to clipboard


                    
@article{osti_1246770,

  title        = {SCALE Continuous-Energy Eigenvalue Sensitivity Coefficient Calculations},

  author       = {Perfetti, Christopher M. and Rearden, Bradley T. and Martin, William R.},

  abstractNote = {Sensitivity coefficients describe the fractional change in a system response that is induced by changes to system parameters and nuclear data. The Tools for Sensitivity and UNcertainty Analysis Methodology Implementation (TSUNAMI) code within the SCALE code system makes use of eigenvalue sensitivity coefficients for an extensive number of criticality safety applications, including quantifying the data-induced uncertainty in the eigenvalue of critical systems, assessing the neutronic similarity between different critical systems, and guiding nuclear data adjustment studies. The need to model geometrically complex systems with improved fidelity and the desire to extend TSUNAMI analysis to advanced applications has motivated the development of a methodology for calculating sensitivity coefficients in continuous-energy (CE) Monte Carlo applications. The Contributon-Linked eigenvalue sensitivity/Uncertainty estimation via Tracklength importance CHaracterization (CLUTCH) and Iterated Fission Probability (IFP) eigenvalue sensitivity methods were recently implemented in the CE-KENO framework of the SCALE code system to enable TSUNAMI-3D to perform eigenvalue sensitivity calculations using continuous-energy Monte Carlo methods. This work provides a detailed description of the theory behind the CLUTCH method and describes in detail its implementation. This work explores the improvements in eigenvalue sensitivity coefficient accuracy that can be gained through the use of continuous-energy sensitivity methods and also compares several sensitivity methods in terms of computational efficiency and memory requirements.},

  doi          = {10.13182/NSE15-12},

  journal      = {Nuclear Science and Engineering},

  number       = 3,

  volume       = 182,

  place        = {United States},

  year         = {Thu Feb 25 00:00:00 EST 2016},

  month        = {Thu Feb 25 00:00:00 EST 2016}

}

Copy to clipboard

Journal Article:

Free Publicly Available Full Text

Accepted Manuscript (DOE)

Publisher's Version of Record

https://doi.org/10.13182/NSE15-12

Other availability

Search WorldCat to find libraries that may hold this journal

Citation Metrics:

Cited by: 37 works

Citation information provided by
Web of Science

Save / Share:

Export Metadata

Save to My Library

Works referenced in this record:

The Relations Between Various Contributon Variables Used in Spatial Channel Theory
journal, June 1977

Williams, M. L.
Nuclear Science and Engineering, Vol. 63, Issue 2
DOI: 10.13182/nse77-a27033

Works referencing / citing this record:

Adjoint-Based Sensitivity and Uncertainty Analysis for Density and Composition: A User’s Guide
journal, March 2017

Favorite, Jeffrey A.; Perkó, Zoltán; Kiedrowski, Brian C.
Nuclear Science and Engineering, Vol. 185, Issue 3
DOI: 10.1080/00295639.2016.1272990

Nuclear data-induced uncertainty quantification of prompt neutron decay constant based on perturbation theory for ADS experiments at KUCA
journal, August 2019

Endo, Tomohiro; Watanabe, Kenichi; Chiba, Go
Journal of Nuclear Science and Technology, Vol. 57, Issue 2
DOI: 10.1080/00223131.2019.1647893

Calculating the k -Eigenvalue Sensitivity to Typical Geometric Perturbations with the Adjoint-Weighted Method in the Continuous-Energy Reactor Monte Carlo Code RMC
journal, June 2019

Li, Hao; Yu, Ganglin; Huang, Shanfang
Nuclear Science and Engineering, Vol. 193, Issue 11
DOI: 10.1080/00295639.2019.1614800

Experimental analysis and uncertainty quantification using random sampling technique for ADS experiments at KUCA
journal, November 2017

Endo, Tomohiro; Chiba, Go; van Rooijen, Willem Frederik Geert
Journal of Nuclear Science and Technology, Vol. 55, Issue 4
DOI: 10.1080/00223131.2017.1403387

Further development of methodology to model TRISO fuel and BISO absorber particles and related uncertainty quantification using SCALE 6
journal, May 2019

Sihlangu, S. F.; Naicker, V. V.; Hou, J.
Journal of Nuclear Science and Technology, Vol. 56, Issue 8
DOI: 10.1080/00223131.2019.1617204

Estimating Code Biases for Criticality Safety Applications with Few Relevant Benchmarks
journal, May 2019

Perfetti, Christopher M.; Rearden, Bradley T.
Nuclear Science and Engineering, Vol. 193, Issue 10
DOI: 10.1080/00295639.2019.1604048

GPT-Free Sensitivity Analysis for Monte Carlo Models
journal, January 2019

Wu, Zeyun; Liang, Jingang; Peng, Xingjie
Nuclear Technology, Vol. 205, Issue 7
DOI: 10.1080/00295450.2018.1556062

Nuclear data-induced uncertainty quantification of prompt neutron decay constant based on perturbation theory for ADS experiments at KUCA
text, January 2019

Endo, Tomohiro; Watanabe, Kenichi; Chiba, Go
Taylor & Francis
DOI: 10.6084/m9.figshare.9616073.v1

Nuclear data-induced uncertainty quantification of prompt neutron decay constant based on perturbation theory for ADS experiments at KUCA
text, January 2019

Endo, Tomohiro; Watanabe, Kenichi; Chiba, Go
Taylor & Francis
DOI: 10.6084/m9.figshare.9616073

Similar Records in DOE PAGES and OSTI.GOV collections:

Impact of Increased Latent Generations on Sensitivity Calculations with SCALE

Conference Greene, Travis M. ; Marshall, William B.J. ; Clarity, Justin B. - ANS Proceedings of Topical Meetings

Analyses of cross section sensitivity data from systems with fissile material allow analysts to associate an importance for each material, nuclide, reaction, and neutron energy by simulating real world criticality scenarios. Although criticality safety validation efforts can be guided by the cross-section sensitivity and uncertainty data generated for a particular system, these calculations can often be computationally expensive and sometimes cumbersome without proper guidance. The TSUNAMI suite within the SCALE code package has several methods for generating sensitivity data, including multigroup and continuous energy (CE) capabilities. The release of SCALE 6.3 has three different CE methods for generating cross sectionmore »« less
https://doi.org/10.13182/T126-37484

Full Text Available
Use of SCALE Continuous-Energy Monte Carlo Tools for Eigenvalue Sensitivity Coefficient Calculations

Conference Perfetti, Christopher M. ; Rearden, Bradley T.

The TSUNAMI code within the SCALE code system makes use of eigenvalue sensitivity coefficients for an extensive number of criticality safety applications, such as quantifying the data-induced uncertainty in the eigenvalue of critical systems, assessing the neutronic similarity between different critical systems, and guiding nuclear data adjustment studies. The need to model geometrically complex systems with improved fidelity and the desire to extend TSUNAMI analysis to advanced applications has motivated the development of a methodology for calculating sensitivity coefficients in continuous-energy (CE) Monte Carlo applications. The CLUTCH and Iterated Fission Probability (IFP) eigenvalue sensitivity methods were recently implemented in themore »« less
SCALE 6.2 Continuous-Energy TSUNAMI-3D Capabilities

Conference Perfetti, Christopher M ; Rearden, Bradley T

The TSUNAMI (Tools for Sensitivity and UNcertainty Analysis Methodology Implementation) capabilities within the SCALE code system make use of sensitivity coefficients for an extensive number of criticality safety applications, such as quantifying the data-induced uncertainty in the eigenvalue of critical systems, assessing the neutronic similarity between different systems, quantifying computational biases, and guiding nuclear data adjustment studies. The need to model geometrically complex systems with improved ease of use and fidelity and the desire to extend TSUNAMI analysis to advanced applications have motivated the development of a SCALE 6.2 module for calculating sensitivity coefficients using three-dimensional (3D) continuous-energy (CE) Montemore »« less
Full Text Available
Analysis of SCALE Criticality and Sensitivity Calculations for Reflected HEU Cylinders

Conference Shaw, Alex M. ; Marshall, William B. J. - ANS Proceedings of Topical Meetings

The SCALE code package offers multiple nuclear data libraries and sensitivity and uncertainty (S/U) methods supporting and derived from Monte Carlo (MC) transport. The CSAS and TSUNAMI-3D sequences use KENO MC, utilizing either continuous-energy (CE) cross sections or multigroup (MG) cross section libraries. TSUNAMI-3D has two CE calculational methods: the iterated fission probability (IFP) method, and the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Tracklength importance CHaracterization (CLUTCH) method. Previous work has shown poor agreement between CLUTCH and confirmatory direct perturbation calculations in specific applications (e.g., fissionable and polyethylene reflectors). The HEU-MET-FAST-084 (HMF-084) International Criticality Safety Benchmark Evaluation Project evaluation consists ofmore »« less
https://doi.org/10.13182/T126-37513
SCALE 6.2 Continuous-energy Tsunami- 3D capabilities - 14566

Conference Perfetti, Christopher M. ; Rearden, Bradley T.

The TSUNAMI (Tools for Sensitivity and Uncertainty Analysis Methodology Implementation) capabilities within the SCALE code system make use of sensitivity coefficients for an extensive number of criticality safety applications, such as quantifying the data-induced uncertainty in the eigenvalue of critical systems, assessing the neutronic similarity between different systems, quantifying computational biases, and guiding nuclear data adjustment studies. The need to model geometrically complex systems with improved ease of use and fidelity and the desire to extend TSUNAMI analysis to advanced applications have motivated the development of a SCALE 6.2 module for calculating sensitivity coefficients using three-dimensional (3D) continuous-energy (CE) Montemore »« less

Similar Records

Title: SCALE Continuous-Energy Eigenvalue Sensitivity Coefficient Calculations

Abstract

Citation Formats

The Relations Between Various Contributon Variables Used in Spatial Channel Theory journal, June 1977

Adjoint-Based Sensitivity and Uncertainty Analysis for Density and Composition: A User’s Guide journal, March 2017

Nuclear data-induced uncertainty quantification of prompt neutron decay constant based on perturbation theory for ADS experiments at KUCA journal, August 2019

Calculating the k -Eigenvalue Sensitivity to Typical Geometric Perturbations with the Adjoint-Weighted Method in the Continuous-Energy Reactor Monte Carlo Code RMC journal, June 2019

Experimental analysis and uncertainty quantification using random sampling technique for ADS experiments at KUCA journal, November 2017

Further development of methodology to model TRISO fuel and BISO absorber particles and related uncertainty quantification using SCALE 6 journal, May 2019

Estimating Code Biases for Criticality Safety Applications with Few Relevant Benchmarks journal, May 2019

GPT-Free Sensitivity Analysis for Monte Carlo Models journal, January 2019

Nuclear data-induced uncertainty quantification of prompt neutron decay constant based on perturbation theory for ADS experiments at KUCA text, January 2019

Nuclear data-induced uncertainty quantification of prompt neutron decay constant based on perturbation theory for ADS experiments at KUCA text, January 2019

The Relations Between Various Contributon Variables Used in Spatial Channel Theory
journal, June 1977

Adjoint-Based Sensitivity and Uncertainty Analysis for Density and Composition: A User’s Guide
journal, March 2017

Nuclear data-induced uncertainty quantification of prompt neutron decay constant based on perturbation theory for ADS experiments at KUCA
journal, August 2019

Calculating the k -Eigenvalue Sensitivity to Typical Geometric Perturbations with the Adjoint-Weighted Method in the Continuous-Energy Reactor Monte Carlo Code RMC
journal, June 2019

Experimental analysis and uncertainty quantification using random sampling technique for ADS experiments at KUCA
journal, November 2017

Further development of methodology to model TRISO fuel and BISO absorber particles and related uncertainty quantification using SCALE 6
journal, May 2019

Estimating Code Biases for Criticality Safety Applications with Few Relevant Benchmarks
journal, May 2019

GPT-Free Sensitivity Analysis for Monte Carlo Models
journal, January 2019

Nuclear data-induced uncertainty quantification of prompt neutron decay constant based on perturbation theory for ADS experiments at KUCA
text, January 2019

Nuclear data-induced uncertainty quantification of prompt neutron decay constant based on perturbation theory for ADS experiments at KUCA
text, January 2019