Neutronic analysis of a tokamak hybrid blanket cooled by thorium-molten salt fuel mixture

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dc.authoridKarakoc, Alper/0000-0002-3692-5895
dc.authoridTUNC, GUVEN/0000-0001-7038-8168
dc.contributor.authorSahin, Haci Mehmet
dc.contributor.authorTunc, Guven
dc.contributor.authorKarakoc, Alper
dc.date.accessioned2024-09-29T16:00:33Z
dc.date.available2024-09-29T16:00:33Z
dc.date.issued2024
dc.departmentKarabük Üniversitesien_US
dc.description.abstractIn this study, a novel approach has been investigated by using a mixture of thorium and molten salts as a dualpurpose coolant and medium for the production of fissile fuel in a Fusion Fission Hybrid Reactor (FFHR) for the reference geometry of ITER. The study highlighted the broader benefits of thorium fuel cycling, safety features, and reduced radioactive minor actinides generation. The use of a thorium-melted salt coolant for fissile fuel production in a fusion-fission hybrid reactor represented a promising path towards efficient and sustainable nuclear energy, with potential benefits in terms of safety features and reduced generation of radioactive minor actinides. In this study, SS 316 LN-IG was selected as the first wall material for the reactor, and a molten salt fuel mixture of LiF-ThF4 was used as the coolant, taking into account the eutectic points of the material, the nominal fusion power in the FFHR for the Tokamak design concept is considered to be 500 MW. The nuclear code MCNP6 was used with the nuclear data libraries ENDF/B-VIII and CLAW-IV for the neutron calculations. The time evolution of the isotopes in the reactor was calculated with the interface code MCNPAS. The study results are evaluated in terms of tritium breeding ratio, energy multiplication factor, radiation damage, fissile fuel production and fuel burn-up value.The 4-year operation history of total TBR value is calculated and always above 1.05 and increases with time.Th initially decreased from 631.3 tonnes to 587.2 tonnes, while 233U production during this period was 9.1 tonnes. According to these results, the first wall replacement period was calculated as 3.9 years.en_US
dc.identifier.doi10.1016/j.pnucene.2024.105306
dc.identifier.issn0149-1970
dc.identifier.issn1878-4224
dc.identifier.scopus2-s2.0-85195368149en_US
dc.identifier.scopusqualityQ2en_US
dc.identifier.urihttps://doi.org/10.1016/j.pnucene.2024.105306
dc.identifier.urihttps://hdl.handle.net/20.500.14619/5217
dc.identifier.volume174en_US
dc.identifier.wosWOS:001263843900001en_US
dc.identifier.wosqualityN/Aen_US
dc.indekslendigikaynakWeb of Scienceen_US
dc.indekslendigikaynakScopusen_US
dc.language.isoenen_US
dc.publisherPergamon-Elsevier Science Ltden_US
dc.relation.ispartofProgress in Nuclear Energyen_US
dc.relation.publicationcategoryMakale - Uluslararası Hakemli Dergi - Kurum Öğretim Elemanıen_US
dc.rightsinfo:eu-repo/semantics/closedAccessen_US
dc.subjectFFHRen_US
dc.subjectITERen_US
dc.subjectTokamaken_US
dc.subjectThoriumen_US
dc.subjectMolten salt mixtureen_US
dc.titleNeutronic analysis of a tokamak hybrid blanket cooled by thorium-molten salt fuel mixtureen_US
dc.typeArticleen_US

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