964 research outputs found

    Nuclear data libraries for IFMIF-DONES neutronic calculations

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    International Fusion Materials Irradiation Facility-DEMO Oriented NEutron Source (IFMIF-DONES) 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 that in the first wall of a fusion power reactor is achieved. In this paper we investigate the differences in the neutronic calculations of the IFMIF-DONES design when using different nuclear data libraries. We first studied the differences in neutron production due to Li(d, xn) reactions between different models and evaluations, comparing the different results with experimental data. Additionally, we tested the performance of the MCNP6.2 and Geant4 Monte Carlo codes when using deuteron incident data libraries. Then, we 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 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 also carried out these same neutronic calculations while using different nuclear data libraries for the neutron transport

    Measurement of the neutron capture cross section of the fissile isotope 235U with the CERN n_TOF Total Absorption Calorimeter and a fission tagging based on micromegas detectors

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    Actual and future nuclear technologies require more accurate nuclear data on the (n, gamma) cross sections and -ratios of fissile isotopes. Their measurement presents several difficulties, mainly related to the strong fission gamma-ray background competing with the weaker gamma-ray cascades used as the experimental signature of the (n,gamma) process. A specific setup has been used at the CERN n_TOF facility in 2012 for the measurement of the (n,gamma) cross section and alpha-ratios of fissile isotopes and used for the case of the 235U isotope. The setup consists in a set of micromegas fission detectors surrounding 235U samples and placed inside the segmented BaF2 Total Absorption Calorimeter.Postprint (published version

    New measurement of the 242Pu(n,Îł) cross section at n_TOF

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    WONDER-2015 – 4th International Workshop On Nuclear Data Evaluation for Reactor applicationsThe use of MOX fuel (mixed-oxide fuel made of UO2 and PuO2) in nuclear reactors allows substituting a large fraction of the enriched Uranium by Plutonium reprocessed from spent fuel. With the use of such new fuel composition rich in Pu, a better knowledge of the capture and fission cross sections of the Pu isotopes becomes very important. In particular, a new series of cross section evaluations have been recently carried out jointly by the European (JEFF) and United States (ENDF) nuclear data agencies. For the case of 242Pu, the two only neutron capture time-of-flight measurements available, from 1973 and 1976, are not consistent with each other, which calls for a new time-of flight capture cross section measurement. In order to contribute to a new evaluation, we have perfomed a neutron capture cross section measurement at the n_TOF-EAR1 facility at CERN using four C6D6 detectors, using a high purity target of 95 mg. The preliminary results assessing the quality and limitations (background, statistics and γ-flash effects) of this new experimental data are presented and discussed, taking into account that the aimed accuracy of the measurement ranges between 7% and 12% depending on the neutron energy regionMinisterio de Economía y Competitividad FPA2013-45083-PMinisterio de Economía y Competitividad FPA2014-53290-C2-2-

    Measurement of the neutron capture cross section of the fissile isotope 235U with the CERN n-TOF total absorption calorimeter and a fission tagging based on micromegas detectors

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    The accuracy on neutron capture cross section of fissile isotopes must be improved for the design of future nuclear systems such as Gen-IV reactors and Accelerator Driven Systems. The High Priority Request List of the Nuclear Energy Agency, which lists the most important nuclear data requirements, includes also the neutron capture cross sections of fissile isotopes such as 233,235U and 239,241Pu. A specific experimental setup has been used at the CERN n TOF facility for the measurement of the neutron capture cross section of 235U by a set of micromegas fission detectors placed inside a segmented BaF2 Total Absorption Calorimeter.Plan Nacional de I+D+I FĂ­sica de particulas FPA2014-53290-C2-
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