Assessment of the neutronic performance of some alternative fluids in a fusion-fission hybrid reactor by using Monte Carlo method

Basit Öğe Kaydı
dc.authorwosidDÜZ, Mehtap/AAS-3672-2020
dc.contributor.authorGunay, Mehtap
dc.date.accessioned2024-08-04T20:37:36Z
dc.date.available2024-08-04T20:37:36Z
dc.date.issued2013
dc.departmentİnönü Üniversitesien_US
dc.description.abstractIn this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVFa Ferritic steel structural material and the molten salt-heavy metal mixtures 99-95% Li20Sn80-1-5% SFG-Pu, 99-95% Li20Sn80-1-5% SFG-PuF4, or 99-95% Li20Sn80-1-5% SFG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as neutron flux, tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fissile fuel breeding in liquid first wall, blanket and shield zones and investigates effects of spent fuel grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.O. Published by Elsevier Ltd.en_US
dc.identifier.doi10.1016/j.anucene.2013.04.036
dc.identifier.endpage97en_US
dc.identifier.issn0306-4549
dc.identifier.scopus2-s2.0-84878325909en_US
dc.identifier.scopusqualityQ1en_US
dc.identifier.startpage93en_US
dc.identifier.urihttps://doi.org/10.1016/j.anucene.2013.04.036
dc.identifier.urihttps://hdl.handle.net/11616/96072
dc.identifier.volume60en_US
dc.identifier.wosWOS:000322854700012en_US
dc.identifier.wosqualityQ2en_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.subjectTBRen_US
dc.subjectMen_US
dc.subjectFissile fuelen_US
dc.subjectMCNPX-2.7.0en_US
dc.titleAssessment of the neutronic performance of some alternative fluids in a fusion-fission hybrid reactor by using Monte Carlo methoden_US
dc.typeArticleen_US

Dosyalar

Koleksiyon

WoS İndeksli Yayınlar Koleksiyonu
Scopus İndeksli Yayınlar Koleksiyonu