Abstract
An assessment of nuclear data performance for non-light-water reactor (non-LWR) reactivity calculations was performed at Oak Ridge National Laboratory that involved a thorough literature review to collect related observations made across different research institutions, an interrogation of the latest ENDF/B evaluated nuclear data libraries, and propagation of nuclear data uncertainties to key figures of merit associated with reactor safety for six non-LWR benchmarks. The outcome of this comprehensive study was published in a technical report issued by the US Nuclear Regulatory Commission. This paper provides a summary of the study’s key observations and conclusions and demonstrates with two examples how the various methods available in the SCALE code system were used to identify key cross section uncertainties for non-LWR reactivity analyses.
| Original language | English |
|---|---|
| Article number | 1159478 |
| Journal | Frontiers in Energy Research |
| Volume | 11 |
| DOIs | |
| State | Published - 2023 |
Funding
This work was supported by the US Nuclear Regulatory Commission. Although many studies assessing the impact of nuclear data uncertainties are available in the public literature, a comprehensive overview of the impact of nuclear data uncertainties for reactivity in the most relevant non-LWRs designs (in terms of reactor concepts for which license applications are expected in the near future in the United States) based on the same set of evaluated nuclear data libraries and using the same simulation approaches did not exist until recently. A recently concluded project at Oak Ridge National Laboratory (ORNL) sponsored by the US Nuclear Regulatory Commission (NRC) addressed this need by performing a thorough literature study to collect the observations made across different research institutions and by using SCALE [Wieselquist, W. A., Lefebvre, R. A., and Jessee, M. A. (Eds.), 2020] to systematically propagate nuclear data uncertainties to key figures of merit associated with reactor safety for five non-LWR types: high-temperature gas-cooled reactor (HTGR), molten salt reactor (MSR), fluoride salt–cooled high-temperature reactor (FHR), heat pipe reactor (HPR), and sodium-cooled fast reactors (SFRs). As part of this study, missing nominal nuclear data and nuclear data uncertainties were identified for reactivity analyses as well as for further fuel depletion analysis. This paper provides a summary of key observations and conclusions obtained during this study, while providing just two examples to demonstrate how the computational analyses were performed. Detailed analysis results are available in a comprehensive technical report () issued by the NRC. It is noted that the study focused on systems with U enriched or mixed uranium/plutonium fuel based on spent LWR fuel; U-fueled systems were not considered here. 235 233 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 ).
| Funders | Funder number |
|---|---|
| U.S. Department of Energy | |
| U.S. Nuclear Regulatory Commission | |
| Oak Ridge National Laboratory | |
| UT-Battelle | DE-AC05-00OR22725 |
| National Research Council Canada |
Keywords
- non-LWR
- nuclear data
- reactivity
- scale
- sensitivity analysis
- uncertainty analysis