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 section processing and libraries provide excellent results for advanced reactor analysis.
| Original language | English |
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| Title of host publication | International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019 |
| Publisher | American Nuclear Society |
| Pages | 2846-2855 |
| Number of pages | 10 |
| ISBN (Electronic) | 9780894487699 |
| State | Published - 2019 |
| Event | 2019 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019 - Portland, United States Duration: Aug 25 2019 → Aug 29 2019 |
Publication series
| Name | International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019 |
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Conference
| Conference | 2019 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2019 |
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| Country/Territory | United States |
| City | Portland |
| Period | 08/25/19 → 08/29/19 |
Funding
This research was supported by the US Nuclear Regulatory Commission Office of Research. This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the US Department of Energy (DOE). The US government retains and the publisher, by accepting the article for publication, acknowledges that the US government retains a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for US government purposes. DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-plan).
Keywords
- 1597-group
- AMPX
- Advanced reactors
- Cross section
- SCALE