Uncertainty comparison between ENDF/B-VIII.0 and ENDF/B-VII.1 for fast reactor BN-600 using high-precision sampling method

A method of using small samples to achieve high-precision related variable sequence sampling was proposed. This method mainly uses two techniques: Latin hypercube sampling method for normal distribution and transform the covariance between variables of standard normal distribution sampling samples as unit. Applying this method, the fast reactor uncertainty analysis and sensitivity analysis program, named SUFR1.0, was developed, and conducted verification analysis. Using the NJOY2016 program, 33-group covariance libraries based on ENDF/B-VIII.0 and ENDF/B-VII.1 were developed for uncertainty and sensitivity analysis for fast reactors BN-600. Using the above 33-group covariance libraries, the uncertainty of the effective multiplication factor k(eff) and the uncertainty of the power distribution of core were analyzed. It was found that the covariance matrix values between ENDF/B-VIII.0 and ENDF/B-VII.1, for some nuclides, such as U-238, (PU)-P-239 and Fe-56 within some neutron energy range, are quite different. At the same time, these relatively different cross section types will also cause a large difference in the relative uncertainty of k(eff). For the relative uncertainty of power, the calculation result of ENDF/B-VIII.0 is significantly smaller than that of ENDF/B-VII.1. (C) 2021 Elsevier Ltd. All rights reserved.

Automated summary

A method of using small samples to achieve high-precision related variable sequence sampling was proposed, and a fast reactor uncertainty analysis and sensitivity analysis program was developed for verification. The analysis showed significant differences in covariance matrix values between different libraries for certain nuclides, leading to varying levels of uncertainty for key parameters.