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Title: Initial verification and validation of ENDF/B-VII.0 libraries of MC{sup 2}-2 against fast critical systems.

Abstract

No abstract prepared.

Authors:
; ; ;
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
NE
OSTI Identifier:
920220
Report Number(s):
ANL/NE/CP-59598
TRN: US0805648
DOE Contract Number:
DE-AC02-06CH11357
Resource Type:
Conference
Resource Relation:
Journal Name: Trans. Am. Noc. Soc. Vol. 97; Journal Issue: 2007; Conference: ANS/ENS International Winter Meeting and Nuclear Technology Expo; Nov 11-15, 2007; Washington, DC
Country of Publication:
United States
Language:
ENGLISH
Subject:
73 NUCLEAR PHYSICS AND RADIATION PHYSICS; M CODES; VALIDATION; VERIFICATION; NUCLEAR DATA COLLECTIONS; NEUTRON SPECTRA; FAST NEUTRONS

Citation Formats

Lee, C. H., Yang, W. S., Hill, R. N., and Nuclear Engineering Division. Initial verification and validation of ENDF/B-VII.0 libraries of MC{sup 2}-2 against fast critical systems.. United States: N. p., 2007. Web.
Lee, C. H., Yang, W. S., Hill, R. N., & Nuclear Engineering Division. Initial verification and validation of ENDF/B-VII.0 libraries of MC{sup 2}-2 against fast critical systems.. United States.
Lee, C. H., Yang, W. S., Hill, R. N., and Nuclear Engineering Division. Mon . "Initial verification and validation of ENDF/B-VII.0 libraries of MC{sup 2}-2 against fast critical systems.". United States. doi:.
@article{osti_920220,
title = {Initial verification and validation of ENDF/B-VII.0 libraries of MC{sup 2}-2 against fast critical systems.},
author = {Lee, C. H. and Yang, W. S. and Hill, R. N. and Nuclear Engineering Division},
abstractNote = {No abstract prepared.},
doi = {},
journal = {Trans. Am. Noc. Soc. Vol. 97},
number = 2007,
volume = ,
place = {United States},
year = {Mon Jan 01 00:00:00 EST 2007},
month = {Mon Jan 01 00:00:00 EST 2007}
}

Conference:
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  • ANSI/ANS-8.1-1998;2007, Nuclear Criticality Safety in Operations with Fissionable Material Outside Reactors, and ANSI/ANS-8.24-2007, Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations, require validation of a computer code and the associated data through benchmark evaluations based on physical experiments. The performance of the code and data are validated by comparing the calculated and the benchmark results. A SCALE procedure has been established to generate a Verified, Archived Library of Inputs and Data (VALID). This procedure provides a framework for preparing, peer reviewing, and controlling models and data sets derived from benchmark definitions so that the models and data canmore » be used with confidence. The procedure ensures that the models and data were correctly generated using appropriate references with documented checks and reviews. Configuration management is implemented to prevent inadvertent modification of the models and data or inclusion of models that have not been subjected to the rigorous review process. VALID entries for criticality safety are based on critical experiments documented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). The findings of a criticality safety validation of SCALE 6.1 utilizing the benchmark models vetted in the VALID library at Oak Ridge National Laboratory are summarized here.« less
  • No abstract prepared.
  • An extensive and systematic verification, validation and data testing effort using diverse Monte Carlo and deterministic methods with ENDF/B data (Versions V.2 and VI.5) is in progress. Methods verification is obtained by comparing independent methods, including continuous-energy Monte Carlo methods using the VIM and MCNP codes and 1D and 2D transport calculations using multigroup cross section sets generated with the ETOE-II/MC-2 system combined with the TWODANT code. Each of these code systems relies on independently processed data libraries. Inter-comparison of these results is used for verification. Further verification is achieved by comparing and plotting the point-wise cross section libraries ofmore » VIM and MCNP. Where possible, benchmark models are taken from both the ICSBEP Handbook and the CSEWG Benchmark Specifications. Inter-comparison of these results quantifies the generally small differences between these references. Calculations are repeated with both ENDF/B-V.2 and ENDF/B-VI.5 data. Inter-comparison of these results quantifies their data dependence. Analyses of criticality and reaction rate ratios are used for the validation and data testing. Calculated neutron balances and neutron energy spectra are also obtained and analyzed to explain the observed differences.« less