Publication:
Determination of plutonium and uranium content and burnup using six group delayed neutrons

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dc.contributor.authorAKYÜREK, TAYFUN
dc.contributor.authorsAkyurek, T.; Usman, S.
dc.date.accessioned2022-03-14T09:15:17Z
dc.date.accessioned2026-01-11T16:19:16Z
dc.date.available2022-03-14T09:15:17Z
dc.date.issued2019-07
dc.description.abstractIn this study, investigation of spent fuel was performed using six group delayed neutron parameters. Three used fuels (F1, F2, and F11) which are burnt over the years in the core of Missouri University of Science and Technology Reactor (MSTR), were investigated. F16 fresh fuel was used as plutonium free fuel element and compared with irradiated used fuels to develop burnup and Pu discrimination method. The fast fission factor of the MSTR was calculated to be 1.071 which was used for burnup calculations. Burnup values of F2 and F11 fuel elements were estimated to be 1.98 g and 2.7 g, respectively. Pu-239 conversion was calculated to be 0.36 g and 0.50 g for F2 and F11 elements, respectively. (C) 2019 Korean Nuclear Society, Published by Elsevier Korea LLC.
dc.identifier.doi10.1016/j.net.2019.01.005
dc.identifier.issn1738-5733
dc.identifier.urihttps://hdl.handle.net/11424/242857
dc.identifier.wosWOS:000468474600002
dc.language.isoeng
dc.publisherKOREAN NUCLEAR SOC
dc.relation.ispartofNUCLEAR ENGINEERING AND TECHNOLOGY
dc.rightsinfo:eu-repo/semantics/openAccess
dc.subjectFuel elements
dc.subjectBurnup
dc.subjectDelay neutrons
dc.subjectSix group parameters
dc.subjectMISSOURI UNIVERSITY
dc.subjectMOX FUEL
dc.subjectSPECTRUM
dc.subjectFISSION
dc.titleDetermination of plutonium and uranium content and burnup using six group delayed neutrons
dc.typearticle
dspace.entity.typePublication
oaire.citation.endPage948
oaire.citation.issue4
oaire.citation.startPage943
oaire.citation.titleNUCLEAR ENGINEERING AND TECHNOLOGY
oaire.citation.volume51

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