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Title: Energy meshing techniques for processing ENDF/B-VI cross sections using the AMPX code system

Abstract

Modern techniques for the establishment of criticality safety for fissile systems invariably require the use of neutronic transport codes with applicable cross-section data. Accurate cross-section data are essential for solving the Boltzmann Transport Equation for fissile systems. In the absence of applicable critical experimental data, the use of independent calculational methods is crucial for the establishment of subcritical limits. Moreover, there are various independent modern transport codes available to the criticality safety analyst (e.g., KENO V.a., MCNP, and MONK). In contrast, there is currently only one complete software package that processes data from the Version 6 format of the Evaluated Nuclear Data File (ENDF) to a format useable by criticality safety codes. To facilitate independent cross-section processing, Oak Ridge National Laboratory (ORNL) is upgrading the AMPX code system to enable independent processing of Version 6 formats using state-of-the-art procedures. The AMPX code system has been in continuous use at ORNL since the early 1970s and is the premier processor for providing multigroup cross sections for criticality safety analysis codes. Within the AMPX system, the module POLIDENT is used to access the resonance parameters in File 2 of an ENDF/B library, generate point cross-section data, and combine the cross sections withmore » File 3 point data. At the heart of any point cross-section processing code is the generation of a suitable energy mesh for representing the data. The purpose of this work is to facilitate the AMPX upgrade through the development of a new and innovative energy meshing technique for processing point cross-section data.« less

Authors:
; ;  [1]
  1. Oak Ridge National Lab., TN (United States)
Publication Date:
OSTI Identifier:
678137
Report Number(s):
CONF-990605-
Journal ID: TANSAO; ISSN 0003-018X; TRN: 99:009122
Resource Type:
Journal Article
Journal Name:
Transactions of the American Nuclear Society
Additional Journal Information:
Journal Volume: 80; Conference: 1999 annual meeting of the American Nuclear Society (ANS), Boston, MA (United States), 6-10 Jun 1999; Other Information: PBD: 1999
Country of Publication:
United States
Language:
English
Subject:
42 ENGINEERING NOT INCLUDED IN OTHER CATEGORIES; 05 NUCLEAR FUELS; 99 MATHEMATICS, COMPUTERS, INFORMATION SCIENCE, MANAGEMENT, LAW, MISCELLANEOUS; A CODES; NUCLEAR DATA COLLECTIONS; CRITICALITY; FISSILE MATERIALS; MODIFICATIONS; MESH GENERATION; CROSS SECTIONS; SAFETY ANALYSIS

Citation Formats

Dunn, M.E., Greene, N.M., and Leal, L.C. Energy meshing techniques for processing ENDF/B-VI cross sections using the AMPX code system. United States: N. p., 1999. Web.
Dunn, M.E., Greene, N.M., & Leal, L.C. Energy meshing techniques for processing ENDF/B-VI cross sections using the AMPX code system. United States.
Dunn, M.E., Greene, N.M., and Leal, L.C. Wed . "Energy meshing techniques for processing ENDF/B-VI cross sections using the AMPX code system". United States.
@article{osti_678137,
title = {Energy meshing techniques for processing ENDF/B-VI cross sections using the AMPX code system},
author = {Dunn, M.E. and Greene, N.M. and Leal, L.C.},
abstractNote = {Modern techniques for the establishment of criticality safety for fissile systems invariably require the use of neutronic transport codes with applicable cross-section data. Accurate cross-section data are essential for solving the Boltzmann Transport Equation for fissile systems. In the absence of applicable critical experimental data, the use of independent calculational methods is crucial for the establishment of subcritical limits. Moreover, there are various independent modern transport codes available to the criticality safety analyst (e.g., KENO V.a., MCNP, and MONK). In contrast, there is currently only one complete software package that processes data from the Version 6 format of the Evaluated Nuclear Data File (ENDF) to a format useable by criticality safety codes. To facilitate independent cross-section processing, Oak Ridge National Laboratory (ORNL) is upgrading the AMPX code system to enable independent processing of Version 6 formats using state-of-the-art procedures. The AMPX code system has been in continuous use at ORNL since the early 1970s and is the premier processor for providing multigroup cross sections for criticality safety analysis codes. Within the AMPX system, the module POLIDENT is used to access the resonance parameters in File 2 of an ENDF/B library, generate point cross-section data, and combine the cross sections with File 3 point data. At the heart of any point cross-section processing code is the generation of a suitable energy mesh for representing the data. The purpose of this work is to facilitate the AMPX upgrade through the development of a new and innovative energy meshing technique for processing point cross-section data.},
doi = {},
journal = {Transactions of the American Nuclear Society},
number = ,
volume = 80,
place = {United States},
year = {1999},
month = {9}
}