Comparison of the Baseline USL Calculation Methods for Loosely Coupled and Novel Neutronic Systems
Author:
Affiliation:
1. The University of New Mexico, Department of Nuclear Engineering, Albuquerque, New Mexico 87131
Funder
Nuclear Criticality Safety program
Publisher
Informa UK Limited
Subject
Nuclear Energy and Engineering
Link
https://www.tandfonline.com/doi/pdf/10.1080/00295639.2023.2249787
Reference9 articles.
1. ANSI/ANS-8.24-2017, “Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations, American Nuclear Society, La Grange Park, Illinois (2017).
2. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data;CHADWICK M. B.;Nucl. Data Sheets,2011
3. A. SANTAMARINA “Benchmark Phase-V Blind Benchmark on MOX Wet Powders ” Organisation for Economic Co-operation and Development Nuclear Energy Agency Working Party on Nuclear Criticality Safety (2015).
4. Testing the Sampling-Based NUSS-RF Tool for Nuclear Data—Related Global Sensitivity Analysis with Monte Carlo Neutronics Calculations;ZHU T.;Nucl. Sci. Eng.,2016
5. NUSS-RF: stochastic sampling-based tool for nuclear data sensitivity and uncertainty quantification
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