Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS
Abstract
Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surface temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.
- Authors:
-
- North Carolina State Univ., Raleigh, NC (United States). Dept. of Nuclear Engineering
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Publication Date:
- Research Org.:
- Idaho National Laboratory (INL), Idaho Falls, ID (United States)
- Sponsoring Org.:
- USDOE Office of Nuclear Energy (NE)
- OSTI Identifier:
- 1357750
- Alternate Identifier(s):
- OSTI ID: 1326416
- Report Number(s):
- INL/JOU-16-38314
Journal ID: ISSN 0306-4549; PII: S0306454916302894
- Grant/Contract Number:
- AC07-05ID14517
- Resource Type:
- Accepted Manuscript
- Journal Name:
- Annals of Nuclear Energy (Oxford)
- Additional Journal Information:
- Journal Name: Annals of Nuclear Energy (Oxford); Journal Volume: 95; Journal Issue: C; Journal ID: ISSN 0306-4549
- Publisher:
- Elsevier
- Country of Publication:
- United States
- Language:
- English
- Subject:
- 22 GENERAL STUDIES OF NUCLEAR REACTORS; CASL; VERA-CS; Uncertainty quantification; Sensitivity analysis
Citation Formats
Brown, C. S., and Zhang, Hongbin. Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS. United States: N. p., 2016.
Web. doi:10.1016/j.anucene.2016.05.016.
Brown, C. S., & Zhang, Hongbin. Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS. United States. https://doi.org/10.1016/j.anucene.2016.05.016
Brown, C. S., and Zhang, Hongbin. Tue .
"Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS". United States. https://doi.org/10.1016/j.anucene.2016.05.016. https://www.osti.gov/servlets/purl/1357750.
@article{osti_1357750,
title = {Uncertainty quantification and sensitivity analysis with CASL Core Simulator VERA-CS},
author = {Brown, C. S. and Zhang, Hongbin},
abstractNote = {Uncertainty quantification and sensitivity analysis are important for nuclear reactor safety design and analysis. A 2x2 fuel assembly core design was developed and simulated by the Virtual Environment for Reactor Applications, Core Simulator (VERA-CS) coupled neutronics and thermal-hydraulics code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). An approach to uncertainty quantification and sensitivity analysis with VERA-CS was developed and a new toolkit was created to perform uncertainty quantification and sensitivity analysis with fourteen uncertain input parameters. Furthermore, the minimum departure from nucleate boiling ratio (MDNBR), maximum fuel center-line temperature, and maximum outer clad surface temperature were chosen as the selected figures of merit. Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis and coolant inlet temperature was consistently the most influential parameter. We used parameters as inputs to the critical heat flux calculation with the W-3 correlation were shown to be the most influential on the MDNBR, maximum fuel center-line temperature, and maximum outer clad surface temperature.},
doi = {10.1016/j.anucene.2016.05.016},
journal = {Annals of Nuclear Energy (Oxford)},
number = C,
volume = 95,
place = {United States},
year = {Tue May 24 00:00:00 EDT 2016},
month = {Tue May 24 00:00:00 EDT 2016}
}
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