Development and Verification of Sampling Code NECP-SOUL for Evaluated Nuclear Data Files in ENDF-6 Format

Abstract

A code called NECP-SOUL is developed to generate samples of evaluated nuclear data files in the ENDF-6 format. The motivation for the development is to obtain more accurate uncertainty quantification results based on the covariance data given in evaluated nuclear data files than the sampling method based on multigroup covariances. NECP-SOUL can handle all covariance data in ENDF-6 files, including average fission neutron multiplicities, resonance parameters, cross sections, angular distributions, energy distributions, and cross sections for the production of radioactive nuclides.

In this work, samples of ENDF-6 files are input to the nuclear data processing code to generate A Compact ENDF (ACE)-formatted libraries, and these ACE-formatted libraries are used by the Monte Carlo code. Through statistical analysis, the uncertainty of the neutronics results can be obtained. NECP-SOUL is verified against another similar code, SANDY.

In addition, the advantage of sampling ENDF-6 files is analyzed by comparing the uncertainty quantification results with those obtained by sampling multigroup covariance matrices. Moreover, the covariances of cross sections and angular distributions newly provided in ENDF/B-VIII.0 are used to analyze their effect on the uncertainty qualification results.

Keywords:

• Uncertainty quantification
• statistical sampling method
• covariance
• ENDF-6

Disclosure Statement

No potential conflict of interest was reported by the authors.

Additional information

Funding

This work was supported by the National Key R&D Program of China [no. 2022YFB1902600] and the National Natural Science Foundation of China [no. 12075183 and 12135019].