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We measured the neutron capture-to-fission cross-section ratio (α ratio) and the capture cross section of 235 U between 0.2 and 200 eV at the n_TOF facility at CERN. The simultaneous measurement of neutron-induced capture and fission rates was performed by means of the n_TOF BaF 2 Total Absorption Calorimeter (TAC), used for detection of γ rays, in combination with a set of micromegas detectors used as fission tagging detectors. The energy dependence of the capture cross section was obtained with help of the 6 Li(n, t) standard reaction determining the n_TOF neutron fluence; the well-known integral of the 235 U(n, f) cross section between 7.8 and 11 eV was then used for its absolute normalization. The α ratio, obtained with slightly higher statistical fluctuations, was determined directly, without need for any reference cross section. To perform the analysis of this measurement we developed a new methodology to correct the experimentally observed effect that the probabilities of detecting a fission reaction in the TAC and the micromegas detectors are not independent. The results of this work have been used in a new evaluation of 235 U performed within the scope of the Collaborative International Evaluated Library Organisation (CIELO) Project, and are consistent with the ENDF/B-VIII.0 and JEFF-3.3 capture cross sections below 4 eV and above 100 eV. However, the measured capture cross section is on average 10% larger between 4 and 100 eV.

IFMIF-DONES (International Fusion Materials Irradiation Facility- DEMO Oriented NEutron Source) is an installation aimed to irradiate with a high neutron flux materials relevant for the construction of the DEMOnstration fusion power plant (DEMO), in order to study the damage due to irradiation. Neutrons are generated using a 40 MeV and 125 mA deuteron beam impinging on a thick liquid lithium target. With these characteristics, damage due to irradiation comparable to those in the first wall of a fusion power reactor are achieved. In this paper we investigate the differences in the neutronic calculations of the IFMIF-DONES design when using different nuclear data libraries. We have first studied the differences in the neutron production due to Li(d,xn) reactions between different models and evaluations, comparing the different results with experimental data. Additionally, we have tested the performance of the MCNP6.2 and Geant4 Monte Carlo codes when using deuteron incident data libraries. Then, we have performed neutronic calculations of the IFMIF-DONES design using the most reliable Li(d,xn) neutron production models available, which are the FZK-2005 and the JENDL/DEU-2020 evaluations according to the results obtained in the first part of the study. Thus, the differences in these evaluations are propagated to different neutronic calculation results: neutron flux, primary displacement damage, gas production and heating in the materials to be irradiated. Finally, we have also carried out these same neutronic calculations, but using different nuclear data libraries for the neutron transport.