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Title: Summary: Verification of the ENDF/B-VII.1 and VIII.0 AMPX 1597-group Libraries for Advanced Reactor Analysis

Abstract

The SCALE/AMPX multigroup (MG) cross section processing procedure has been updated to minimize reactivity differences for various advanced thermal and fast reactor designs, as observed for the current MG libraries, resulting in excellent agreement between the calculations with the new MG libraries and the continuous-energy reference calculations. The SCALE MG calculations are widely applied to thermal spectrum light-water reactor systems, as well as fast spectrum metallic systems. Due to growing interest from industry and regulators in applying SCALE for the design of fast spectrum reactors—both sodium and molten salt—it was desirable to review and improve the SCALE/AMPX procedure for unresolved resonance self-shielded data and high-energy neutron spectra. The data were improved by generating MG unresolved resonance data based on the analytic probability table method with the narrow resonance approximation and by using very fine and intermediate group structures that are typical for fast system analysis. This study focused on verifying the improved SCALE/AMPX MG cross section processing procedure and the new AMPX 1597-group library with the ENDF/B-VII.1 and VIII.0 evaluated nuclear data. The verification was made by performing reaction rate analysis and benchmark calculations for various thermal and fast reactor systems. Results indicate that the improved SCALE/AMPX MG cross sectionmore » processing and libraries provide excellent results for advanced reactor analysis.« less

Authors:
ORCiD logo [1]; ORCiD logo [1]; ORCiD logo [1]; ORCiD logo [1]; ORCiD logo [1]
  1. ORNL
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1561666
DOE Contract Number:  
AC05-00OR22725
Resource Type:
Conference
Resource Relation:
Conference: International Conference on Mathematics and Computational Methods applied to Nuclear Science and Engineering (M&C 2019) - Portland, Oregon, United States of America - 8/25/2019 8:00:00 AM-8/29/2019 8:00:00 AM
Country of Publication:
United States
Language:
English

Citation Formats

Kim, Kang Seog, Bostelmann, Friederike, Holcomb, Andrew M., Ilas, Germina, and Wieselquist, William. Summary: Verification of the ENDF/B-VII.1 and VIII.0 AMPX 1597-group Libraries for Advanced Reactor Analysis. United States: N. p., 2019. Web.
Kim, Kang Seog, Bostelmann, Friederike, Holcomb, Andrew M., Ilas, Germina, & Wieselquist, William. Summary: Verification of the ENDF/B-VII.1 and VIII.0 AMPX 1597-group Libraries for Advanced Reactor Analysis. United States.
Kim, Kang Seog, Bostelmann, Friederike, Holcomb, Andrew M., Ilas, Germina, and Wieselquist, William. Sun . "Summary: Verification of the ENDF/B-VII.1 and VIII.0 AMPX 1597-group Libraries for Advanced Reactor Analysis". United States. https://www.osti.gov/servlets/purl/1561666.
@article{osti_1561666,
title = {Summary: Verification of the ENDF/B-VII.1 and VIII.0 AMPX 1597-group Libraries for Advanced Reactor Analysis},
author = {Kim, Kang Seog and Bostelmann, Friederike and Holcomb, Andrew M. and Ilas, Germina and Wieselquist, William},
abstractNote = {The SCALE/AMPX multigroup (MG) cross section processing procedure has been updated to minimize reactivity differences for various advanced thermal and fast reactor designs, as observed for the current MG libraries, resulting in excellent agreement between the calculations with the new MG libraries and the continuous-energy reference calculations. The SCALE MG calculations are widely applied to thermal spectrum light-water reactor systems, as well as fast spectrum metallic systems. Due to growing interest from industry and regulators in applying SCALE for the design of fast spectrum reactors—both sodium and molten salt—it was desirable to review and improve the SCALE/AMPX procedure for unresolved resonance self-shielded data and high-energy neutron spectra. The data were improved by generating MG unresolved resonance data based on the analytic probability table method with the narrow resonance approximation and by using very fine and intermediate group structures that are typical for fast system analysis. This study focused on verifying the improved SCALE/AMPX MG cross section processing procedure and the new AMPX 1597-group library with the ENDF/B-VII.1 and VIII.0 evaluated nuclear data. The verification was made by performing reaction rate analysis and benchmark calculations for various thermal and fast reactor systems. Results indicate that the improved SCALE/AMPX MG cross section processing and libraries provide excellent results for advanced reactor analysis.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {2019},
month = {9}
}

Conference:
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