Neutronic investigation of the application of certain plutonium-mixed fluids in a fusion-fission hybrid reactor

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dc.authorwosidDÜZ, Mehtap/AAS-3672-2020
dc.contributor.authorGunay, Mehtap
dc.contributor.authorKasap, Hizir
dc.date.accessioned2024-08-04T20:37:51Z
dc.date.available2024-08-04T20:37:51Z
dc.date.issued2014
dc.departmentİnönü Üniversitesien_US
dc.description.abstractIn this study, the fluids that were investigated were contained increased mole fractions of the mixtures molten salt 99-95% Li20Sn80 and, the heavy metals 1-5% SFG-Pu, SFG-PuF4, SFG-PuO2. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for the neutron multiplicity between liquid first wall and blanket. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. The nuclear parameters of a fusion-fission hybrid reactor such as neutron flux, heating, fission reaction rate were investigated according to the mixture components, radial energy spectrum in the designed system. In this study, the effect of spent fuel-grade Pu content in the designed system on these nuclear parameters were calculated by using the three-dimensional Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0. Crown Copyright (C) 2013 Published by Elsevier Ltd. All rights reserved.en_US
dc.identifier.doi10.1016/j.anucene.2013.08.024
dc.identifier.endpage436en_US
dc.identifier.issn0306-4549
dc.identifier.scopus2-s2.0-84883723379en_US
dc.identifier.scopusqualityQ1en_US
dc.identifier.startpage432en_US
dc.identifier.urihttps://doi.org/10.1016/j.anucene.2013.08.024
dc.identifier.urihttps://hdl.handle.net/11616/96199
dc.identifier.volume63en_US
dc.identifier.wosWOS:000327829400049en_US
dc.identifier.wosqualityQ3en_US
dc.indekslendigikaynakWeb of Scienceen_US
dc.indekslendigikaynakScopusen_US
dc.language.isoenen_US
dc.publisherPergamon-Elsevier Science Ltden_US
dc.relation.ispartofAnnals of Nuclear Energyen_US
dc.relation.publicationcategoryMakale - Uluslararası Hakemli Dergi - Kurum Öğretim Elemanıen_US
dc.rightsinfo:eu-repo/semantics/closedAccessen_US
dc.subjectHybrid reactoren_US
dc.subjectNeutron fluxen_US
dc.subjectHeatingen_US
dc.subjectMCNPX-2.7.0en_US
dc.titleNeutronic investigation of the application of certain plutonium-mixed fluids in a fusion-fission hybrid reactoren_US
dc.typeArticleen_US

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