Investigation of neutronic potential of a moderated (D-T) fusion driven hybrid reactor fueled with thorium to breed fissile fuel for LWRs

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dc.authorid0000-0002-5756-4216
dc.contributor.authorYapici, H
dc.contributor.authorSahin, N
dc.contributor.authorBayrak, M
dc.date.accessioned2019-08-01T13:38:39Z
dc.date.available2019-08-01T13:38:39Z
dc.date.issued2000
dc.departmentNiğde ÖHÜ
dc.description.abstractThe potential of a moderated hybrid reactor fueled with ThC2 or ThF4 is investigated by a PF (plant factor) 75% under a first wall fusion neutron current load of 5 MW/m(2) LWR (Light Water Reactor) fuel rods containing ThC2 or ThF4 are replaced in the fissile fuel zone of the hybrid reactor. It is considered that gas (He or CO2), or flibe (Li2BeF4), or natural lithium is the coolant. The behaviour of the neutronic potential is observed for four years. At the end of the operation period, the Cumulative Fissile Fuel Enrichment (CFFE) values varied between 3.55 and 7% depending on the fuel and coolant type. Calculations show that the best neutronic performance is obtained with Flibe, followed by air and natural lithium coolants. After 48 months, the maximum CFFE value is 7% in the ThF4 fuel and flibe coolant mode, and the lowest CFFE value 3.55% is in the ThC2 fuel and natural lithium coolant mode. Consequently, these enrichments would be sufficient for LWRs. The Tritium Breeding Ratio (TBR) values are greater than 1.05 for all investigated natural lithium coolant modes, and the investigated hybrid reactor is self-sufficient in the tritium required for the D,T fusion driver in these modes during the operation period. The blanket energy multiplication factor M, varies between 2.45 and 3.68 depending on the type of fuel and coolant at the end of the operation period. At the same time, the peak-to-average fission power density ratio decreases by similar to 25%. The lowest radial neutron leakage out of the blanket is in the blanket with the flibe coolant modes. (C) 1999 Elsevier Science Ltd. All rights reserved.
dc.identifier.doi10.1016/S0196-8904(99)00118-1
dc.identifier.endpage447
dc.identifier.issn0196-8904
dc.identifier.issue5
dc.identifier.scopus2-s2.0-0033890157
dc.identifier.scopusqualityQ1
dc.identifier.startpage435
dc.identifier.urihttps://dx.doi.org/10.1016/S0196-8904(99)00118-1
dc.identifier.urihttps://hdl.handle.net/11480/5779
dc.identifier.volume41
dc.identifier.wosWOS:000083727800002
dc.identifier.wosqualityQ3
dc.indekslendigikaynakWeb of Science
dc.indekslendigikaynakScopus
dc.institutionauthor[0-Belirlenecek]
dc.language.isoen
dc.publisherPERGAMON-ELSEVIER SCIENCE LTD
dc.relation.ispartofENERGY CONVERSION AND MANAGEMENT
dc.relation.publicationcategoryMakale - Uluslararası Hakemli Dergi - Kurum Öğretim Elemanı
dc.rightsinfo:eu-repo/semantics/closedAccess
dc.subjectfusion
dc.subjecthybrid reactor
dc.subjectthorium
dc.subjectneutronic potential
dc.titleInvestigation of neutronic potential of a moderated (D-T) fusion driven hybrid reactor fueled with thorium to breed fissile fuel for LWRs
dc.typeArticle

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