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Title: Impact of the 235U Covariance Data in Benchmark Calculations

Abstract

The error estimation for calculated quantities relies on nuclear data uncertainty information available in the basic nuclear data libraries such as the U.S. Evaluated Nuclear Data File (ENDF/B). The uncertainty files (covariance matrices) in the ENDF/B library are generally obtained from analysis of experimental data. In the resonance region, the computer code SAMMY is used for analyses of experimental data and generation of resonance parameters. In addition to resonance parameters evaluation, SAMMY also generates resonance parameter covariance matrices (RPCM). SAMMY uses the generalized least-squares formalism (Bayes method) together with the resonance formalism (R-matrix theory) for analysis of experimental data. Two approaches are available for creation of resonance-parameter covariance data. (1) During the data-evaluation process, SAMMY generates both a set of resonance parameters that fit the experimental data and the associated resonance-parameter covariance matrix. (2) For existing resonance-parameter evaluations for which no resonance-parameter covariance data are available, SAMMY can retroactively create a resonance-parameter covariance matrix. The retroactive method was used to generate covariance data for 235U. The resulting 235U covariance matrix was then used as input to the PUFF-IV code, which processed the covariance data into multigroup form, and to the TSUNAMI code, which calculated the uncertainty in the multiplication factormore » due to uncertainty in the experimental cross sections. The objective of this work is to demonstrate the use of the 235U covariance data in calculations of critical benchmark systems.« less

Authors:
 [1];  [1];  [1];  [1];  [1]
  1. ORNL
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
USDOE National Nuclear Security Administration (NNSA)
OSTI Identifier:
962609
DOE Contract Number:
DE-AC05-00OR22725
Resource Type:
Conference
Resource Relation:
Conference: International Conference on the Physics of Reactors - Nuclear Power: A Sustainable Resource Cassino, Interlaken, Switzerland, 20080914, 20080919
Country of Publication:
United States
Language:
English
Subject:
29 ENERGY PLANNING, POLICY AND ECONOMY; BENCHMARKS; COMPUTER CODES; CROSS SECTIONS; EVALUATION; MATRICES; MULTIPLICATION FACTORS; NUCLEAR DATA COLLECTIONS; NUCLEAR POWER; PHYSICS; R MATRIX; RESONANCE; covariance; SAMMY; ENDF/B; nuclear data; TSUNAMI

Citation Formats

Leal, Luiz C, Mueller, Don, Arbanas, Goran, Wiarda, Dorothea, and Derrien, Herve. Impact of the 235U Covariance Data in Benchmark Calculations. United States: N. p., 2008. Web.
Leal, Luiz C, Mueller, Don, Arbanas, Goran, Wiarda, Dorothea, & Derrien, Herve. Impact of the 235U Covariance Data in Benchmark Calculations. United States.
Leal, Luiz C, Mueller, Don, Arbanas, Goran, Wiarda, Dorothea, and Derrien, Herve. 2008. "Impact of the 235U Covariance Data in Benchmark Calculations". United States. doi:.
@article{osti_962609,
title = {Impact of the 235U Covariance Data in Benchmark Calculations},
author = {Leal, Luiz C and Mueller, Don and Arbanas, Goran and Wiarda, Dorothea and Derrien, Herve},
abstractNote = {The error estimation for calculated quantities relies on nuclear data uncertainty information available in the basic nuclear data libraries such as the U.S. Evaluated Nuclear Data File (ENDF/B). The uncertainty files (covariance matrices) in the ENDF/B library are generally obtained from analysis of experimental data. In the resonance region, the computer code SAMMY is used for analyses of experimental data and generation of resonance parameters. In addition to resonance parameters evaluation, SAMMY also generates resonance parameter covariance matrices (RPCM). SAMMY uses the generalized least-squares formalism (Bayes method) together with the resonance formalism (R-matrix theory) for analysis of experimental data. Two approaches are available for creation of resonance-parameter covariance data. (1) During the data-evaluation process, SAMMY generates both a set of resonance parameters that fit the experimental data and the associated resonance-parameter covariance matrix. (2) For existing resonance-parameter evaluations for which no resonance-parameter covariance data are available, SAMMY can retroactively create a resonance-parameter covariance matrix. The retroactive method was used to generate covariance data for 235U. The resulting 235U covariance matrix was then used as input to the PUFF-IV code, which processed the covariance data into multigroup form, and to the TSUNAMI code, which calculated the uncertainty in the multiplication factor due to uncertainty in the experimental cross sections. The objective of this work is to demonstrate the use of the 235U covariance data in calculations of critical benchmark systems.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2008,
month = 1
}

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  • The ENDF/B-VI (Evaluated Nuclear Data File) data evaluations reflect several major changes relative to version V data that will be of interest in thermal reactor analysis. Some of the important changes in the {sup 235}U, {sup 239}Pu, and {sup 238}U data are summarized. It can be seen that significant modifications have been made in the {sup 235}U and {sup 239}Pu thermal data. Also, completely new resonance evaluations based on the Reich-Moore formalism have been performed for {sup 235}U, {sup 239}Pu, and {sup 238}U, and the number of resolved resonances included in the evaluations has increased dramatically. To assess the meritmore » of these modifications, it is first necessary to review the performance of ENDF/B-V in thermal reactor analysis and then to project the estimated impact of ENDF/B-VI on the version V results.« less
  • A series of tests investigating dynamic pulse buckling of a cylindrical shell under axial impact is compared to several 2D and 3D finite element simulations of the event. The purpose of the work is to investigate the performance of various analysis codes and element types on a problem which is applicable to radioactive material transport packages, and ultimately to develop a benchmark problem to qualify finite element analysis codes for the transport package design industry. During the pulse buckling tests, a buckle formed at each end of the cylinder, and one of the two buckles became unstable and collapsed. Numericalmore » simulations of the test were performed using PRONTO, a Sandia developed transient dynamics analysis code, and ABAQUS/Explicit with both shell and continuum elements. The calculations are compared to the tests with respect to deformed shape and impact load history.« less
  • A series of tests investigating dynamic pulse buckling of a cylindrical shell under axial impact is compared to several finite element simulations of the event. The purpose of the study is to compare the performance of the various analysis codes and element types with respect to a problem which is applicable to radioactive material transport packages, and ultimately to develop a benchmark problem to qualify finite element analysis codes for the transport package design industry.
  • The generation of Multi-group cross sections together with relevant uncertainties is fundamental to assess the quality of neutronic simulations. Key information that are needed to propagate the microscopic experimental uncertainties up to macroscopic reactor calculations are (1) the experimental covariance matrices, (2) the correlations between the parameters of the model and (3) the multi-group covariance matrices. This work proposes to apply a Monte-Carlo approach to provide explicit data covariances between the resonance parameters. (authors)