Nuclear data sensitivity and uncertainty analysis of effective neutron multiplication factor in various MYRRHA core configurations

Sensitivity and uncertainty analysis of βeff for MYRRHA using a Monte Carlo technique

“…From the viewpoint of ensuring safety margins and reducing uncertainty in the MYRRHA design parameters, uncertainty quantification of reactor physics parameters is one of the most important tasks. To this end, nuclear data sensitivity and uncertainty (S/U) analyses have been extensively conducted for various MYRRHA core configurations using different calculation tools, geometric models, and nuclear data libraries [3][4][5]; these works have focused on the effective neutron multiplication factor k eff as the primary neutronic safety parameter. The effective delayed neutron fraction b eff can be ranked second in the list of neutronic safety parameters, because, besides of reactor kinetics, it is used to determine other design and safety parameters such as control rod worth and Doppler coefficient.…”

Correlationν̅p − σ − χin the fast neutron range via integral information

“…They are not contradictory and should be used together for a final evaluation, for instance using a set of differential and integral weights (as presented in reference [31]); -for a complete evaluation using the integral information, a variety of quantities needs to be involved, covering both a large energy region and divers integral quantities (not only k eff , but also reaction rates, spectra indexes, activation, or emitted neutron spectra from pulsed spheres. Additionally, the correlations between benchmarks need to be taken into account, for instance using a generalized chi-2 instead of equation (1). In this case, the experimental covariance matrices need to be assessed with care since their impact will be of importance; -as mentioned, integral benchmarks can also bring correlations between isotopes (e.g.…”

“…Many simulation tools are capable of using such matrices to propagate nuclear data uncertainties on final quantities, with either perturbation theories [1][2][3][4][5][6], or Monte Carlo sampling [3,[7][8][9][10][11][12][13]. These results can for instance be used for the review procedure of new facilities, or during the safety assessment of new reactor core designs [14].…”

Nuclear data analyses for improving the safety of advanced lead-cooled reactors

“…In this paper, a target accuracy assessment of the effective neutron multiplication factor, keff, for MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) [4] lead-bismuth cooled fast reactor is performed with the JEFF-3.3 [5] and ENDF/B-VIII.0 [6] state-of-the-art nuclear data libraries and the SUMMON system [7]. Uncertainty quantification (UQ) analyses have already been carried out for previous MYRRHA designs [8][9][10] or with older nuclear data evaluations [11]; nevertheless, since JEFF-3.3 and ENDF/B.VIII.0 were released, an updated analysis was required in order to take into account the changes in the new nuclear data libraries and their impact in the criticality safety coefficients.…”