Nuclear Data Assessment for Advanced Reactors (NUREG/CR-7289, ORNL/TM-2021/2002)

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Publication Information

Manuscript Completed: August 2021
Date Published: March 2022

Prepared by:
F. Bostelmann
G. Ilas
C. Celik
A. M. Holcomb
W. A. Wieselquist

Oak Ridge National Laboratory
Oak Ridge, TN 37831-6283

Timothy Drzewiecki, NRC Project Manager

Prepared for:
Division of Advanced Reactors and Non-power Production and Utilization Facilities
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555-0001

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Abstract

Advanced reactor concepts being developed throughout the industry are significantly different from light-water reactor (LWR) designs with respect to geometry, materials, and operating conditions, and consequently, with respect to their reactor physics behavior. Given the limited operating experience with non-LWRs, the accurate simulation of reactor physics and the quantification of associated uncertainties are important for ensuring that the nuclear design for advanced reactor concepts include appropriate margins. Nuclear data are a major source of input uncertainties in reactor physics analysis. As part of a project sponsored by the US Nuclear Regulatory Commission at Oak Ridge National Laboratory (ORNL), key nuclear data relevant to reactor safety analysis in selected advanced reactor technologies¹ were identified, and their impacts on important key figures of merit were assessed based on (1) a review of available advanced reactor specifications, (2) analysis of previous studies performed at ORNL and other research institutions, and (3) sensitivity and uncertainty analyses performed for six selected benchmarks—three experimental and three computational—to quantify the impacts of the identified key nuclear data on several key metrics.

This report summarizes the key nuclear data—nominal data and nuclear data uncertainties—considering the most important nuclear reactions in the fuel and in various materials for the moderator, coolant, and structure of the considered advanced reactors.

The major nominal missing data that were identified consist of thermal scattering data and ^135mXe cross section data for molten salt reactor (MSR) analysis. The identified major gaps with respect to nuclear data uncertainties are (1) the missing uncertainties in the thermal scattering data for high-temperature gas-cooled reactors and moderated MSR systems, and (2) the incomplete uncertainties on angular distributions, particularly for fast spectrum systems such as sodium-cooled fast reactors, fast molten salt reactors, and heat pipe reactors.

Large uncertainties of reactions that are not commonly considered to be relevant in LWR studies were found to be significant for several advanced reactor systems. The large uncertainty of ²³⁸U inelastic scattering in the fast energy range contributes significantly to large output uncertainties in all fast spectrum systems. The large uncertainty of ²³⁵U (n,ɣ) in the fast energy range causes significant reactivity uncertainties in fast neutron spectrum systems that use ²³⁵U-enriched fuel. A large uncertainty of ⁷Li (n,ɣ) causes a large fraction of uncertainty in the output quantities investigated for MSR systems in which lithium is part of the salt.

Special attention should be paid to differences in cross section and uncertainties of different evaluated nuclear data library releases. Significant differences were found in nuclear data that can lead to major differences in reactivity calculations, even for well-known nuclides. In particular, differences in ²³⁵U, ²³⁸U, and ²³⁹Pu nominal and uncertainty data between the Evaluated Nuclear Data File (ENDF)/B-VII.1 and ENDF/B-VIII.0 nuclear data releases are the major causes of differences in calculations when using these libraries.

For MSR systems containing a large amount of FLiBe salt, the update from the ENDF/B-VII.0 to the ENDF/B-VII.1 release of the tritium production cross section for ⁶Li is significant. Similarly, the update of the ¹²C capture cross section from ENDF/B-VII-0 to ENDF/B-VII.1 can have a significant impact on reactivity calculations of graphite-moderated systems, and the update of the ³⁵Cl (n,p) cross section between these libraries have an impact on fast spectrum molten chloride MSRs.

Given the limited amount of experimental measurement data, no conclusion regarding the better performance of either investigated ENDF/B library is drawn. The presented sensitivity analyses inform about nuclear data for which a change could cause a significant change in the calculated metric of interest. The uncertainty analyses, in particular the ranking of contribution to the output uncertainties, can be used to guide future measurement and evaluation efforts to reduce the significant nuclear data uncertainties and thereby significantly reduce the overall observed uncertainties of key figures of merit.

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¹ Selected advanced reactor technologies: graphite moderated high temperature gas-cooled reactor, fluoride salt-cooled high temperature reactor, graphite-moderated molten salt reactor, molten chloride fast spectrum reactor, fast-spectrum heat pipe reactor, sodium-cooled fast reactor