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Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code

creativeworkseries.issn2452-3038
dc.contributor.authorHossain, M. I.
dc.contributor.authorMollah, A. S.
dc.contributor.authorAkter, Y.
dc.contributor.authorFardin, M. Z.
dc.date.accessioned2024-09-18T07:43:56Z
dc.date.available2024-09-18T07:43:56Z
dc.date.issued2023
dc.description.abstractThe goal of this study is to perform neutronic calculations of the VVER-1000 MOX core computational benchmarks with an OpenMC code along with ENDF/B-VII.1 nuclear data library. The results of neutronic analysis using the OpenMC Monte Carlo code for the VVER-1000 MOX core, containing 30% mixed oxide fuel with low enriched uranium fuel, are presented in this study. As per the benchmark report, all six states are considered in the present study. The k<jats:sub>eff</jats:sub> values, assembly average fission reaction rates, and pin-by-pin fission rates were calculated as per benchmark criteria. In addition, 2D thermal and fast neutron-flux distribution were also generated. The reactivity results and neutron flux distribution were compared with other results in which benchmark analysis was performed using the same core geometry and it showed great similarity with slight deviation. This shows that the modeling of the VVER-1000 MOX core was done successfully using OpenMC. Because OpenMC was successfully used for neutronics calculation of the VVER-1000 whole core, it may be mentioned here that OpenMC code can also be utilized for neutronics and other reactor core physics analyses of the VVER-1200 reactor which is to be commissioned in Bangladesh in the upcoming year.
dc.identifier.citationHossain MI, Mollah AS, Akter Y, Fardin MZ (2023) Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code. Nuclear Energy and Technology 9(4): 215-225. https://doi.org/10.3897/nucet.9.91090
dc.identifier.doi10.3897/nucet.9.91090
dc.identifier.issn2452-3038
dc.identifier.urihttps://openrepository.mephi.ru/handle/123456789/14896
dc.identifier.urihttps://nucet.pensoft.net/article/91090/list/9/
dc.relation.ispartofNuclear Energy and Technology
dc.subjectNeutron Flux
dc.subjectMixed Oxide (MOX) fuel Assembly
dc.subjectLow Enriched Uranium (LEU) Assembly
dc.subjectOpenMC
dc.subjectVVER-1000 MOX core
dc.subjectBurnup
dc.titleNeutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code
dc.typejournal-article
dspace.entity.typePublication
oaire.citation.issue4
oaire.citation.volume9
relation.isJournalIssueOfPublication2b3df35a-b8ac-4215-8327-c6220c814cc7
relation.isJournalIssueOfPublication.latestForDiscovery2b3df35a-b8ac-4215-8327-c6220c814cc7
relation.isJournalOfPublicatione11ecc6c-1851-485a-af73-18119b4fce7f
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