Systematic and quantitative uncertainty analysis for rod ejection accident of pressurized water reactor

A systematic uncertainty analysis strategy based on stochastic sampling method is developed for best estimate analysis of the rod ejection accident in pressurized water reactor (PWR). Self-developed SAMP module and DAKOTA code are used to randomly sample the multigroup nuclear data, important neutronics and thermal-hydraulic (T-H) input parameters respectively. The 3D coupled reactor core model for Almaraz PWR plant is established using RELAP5/PARCS code system and the uncertainties of the core power, fuel enthalpy and local power distribution are obtained. The results indicate that the uncertainties of nuclear data and input parameters can result in large uncertainty on the results which should be carefully considered.