Neutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport code

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dc.authorwosidDÜZ, Mehtap/AAS-3672-2020
dc.contributor.authorGunay, Mehtap
dc.date.accessioned2024-08-04T20:39:41Z
dc.date.available2024-08-04T20:39:41Z
dc.date.issued2014
dc.departmentİnönü Üniversitesien_US
dc.description.abstractIn this study, the molten salt-heavy metal mixtures 93-85 % Li20Sn80 + 5 % SFG-PuO2 and 2-10% UO2, 93-85% Li20Sn80 + 5 % SFG-PuO2 and 2-10% NpO2, 93-85% Li20Sn80 + 5 % SFG-PuO2 and 2-10% UCO were used as fluids. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library.en_US
dc.identifier.doi10.3139/124.110408
dc.identifier.endpage149en_US
dc.identifier.issn0932-3902
dc.identifier.issue2en_US
dc.identifier.scopus2-s2.0-84900870302en_US
dc.identifier.scopusqualityN/Aen_US
dc.identifier.startpage145en_US
dc.identifier.urihttps://doi.org/10.3139/124.110408
dc.identifier.urihttps://hdl.handle.net/11616/96450
dc.identifier.volume79en_US
dc.identifier.wosWOS:000335542700008en_US
dc.identifier.wosqualityQ4en_US
dc.indekslendigikaynakWeb of Scienceen_US
dc.indekslendigikaynakScopusen_US
dc.language.isoenen_US
dc.publisherCarl Hanser Verlagen_US
dc.relation.ispartofKerntechniken_US
dc.relation.publicationcategoryMakale - Uluslararası Hakemli Dergi - Kurum Öğretim Elemanıen_US
dc.rightsinfo:eu-repo/semantics/closedAccessen_US
dc.subjectStructural Materialen_US
dc.subjectApex Reactoren_US
dc.subjectRadiation-Damageen_US
dc.subjectFusionen_US
dc.subjectWallen_US
dc.titleNeutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport codeen_US
dc.typeArticleen_US

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