Abstract. In nuclear safety research, the quality of the results of simulation codes is widely determined by the reactor design and safe operation, and the description of neutron transport in the reactor core is a feature of particular importance. Moreover, for the long effort that is made, there remain uncertainties in simulation results due to the neutronic data and input specification that need a huge effort to be eliminated. A realistic estimation of these uncertainties is required for finding out the reliability of the results. This explains the increasing demand in recent years for calculations in the nuclear fields with best-estimate codes that proved confidence bounds of simulation results. All this has lead to the Benchmark for Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of LWRs of the NEA. The UAM-Benchmark coupling multi-physics and multiscale analysis using as a basis complete sets of input specifications of boiling water reactors (BWR) and pressurized water reactors (PWR). In this study, the results of the transport calculations carried out using the SCALE-6.2 program (TRITON/NEWT and TRITON/KENO modules) as well as Monte Carlo SERPENT code, are presented. Additionally, they have been made uncertainties calculation for a PWR 15 Â 15 and a BWR 7 Â 7 fuel elements, in two different configurations (with and without control rod), and two different states, Hot Full Power (HFP) and Hot Zero Power (HZP), using the TSUNAMI module, which uses the Generalized Perturbation Theory (GPT), and SAMPLER, which uses stochastic sampling techniques for cross-sections perturbations. The results obtained and validated are compared with references results and similar studies presented in the exercise I-2 (Lattice Physics) of UAM-Benchmark.

One of the challenges of study the neutronics of reactor is to generate reliable parameterized libraries, which contains information to simulate the core in all possible operational and transient conditions. These libraries must include tables of cross-sections and other neutronics and kinetics parameters and are obtained by simulating all the assemblies in a transport code. At lattice level, one can use branch calculations to change "instantaneously" the feedback parameters as a function of burnup. When using random sampling for the lattice calculations, one can obtain statistical information of the output parameters and use it in a core simulation to characterize the accuracy of the data estimating uncertainties when simulating a heterogeneous system at different scales of detail. This work presents the methodology to generate NEMTAB libraries from data obtained in SCALE code system to be used in PARCS simulations. The code TXT2NTAB is used to reorder the cross-sections tables in NEMTAB format and generate another NEMTAB of standard deviation. With these libraries, the authors perform a steady state calculation for an LWR to propagate several uncertainties at the core level. The methodology allows to obtain statistical information of the most important output parameters: multiplication factor (keff), axial power peak (Pz) and axial peak node (Nz).