Uncertainty quantificaition on feedback and safety paramerters of the lead-cooled fast reactor

Safety parameters determined using modern codes have a direct impact from nuclear data uncertainties. Evaluation of these uncertainties will lead to a better understanding of their impact on reactor core design and identification of the design safety limits. In this study, “Best Estimate Plus Uncertainty” approach is applied to LFR to propagate nuclear data uncertainties through multiple scales of core modelling. Nuclear Data library ENDF/B-VII.0 with variance-covariance library COMMARA-2.0 is used in this work. For steady state, uncertainties are propagated to selected neutron feedback coefficients using perturbation theory code PERSENT. The evaluated response parameters include Doppler coefficient, core radial expansion coefficient, and fuel/coolant/structure density worth coefficients. Main contributors of uncertainty were traced back to a number of common nuclide reaction pairs including U-238 inelastic, U-238 gamma, Pu-239 gamma, etc. The nuclear data uncertainty was then propagated through transient safety analysis for evaluating core safety performance. This is accomplished by modelling the reactor system in Mini SAS coupled with DAKOTA for uncertainty quantification based on stochastic sampling approach.