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Estimates of the Fast and Termal Flux in Blanket of Critical Reactors by Using Multi-Group Methods

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dc.contributor.authorHançerlioğullari, Aybaba
dc.contributor.authorKurnaz, Aslı
dc.contributor.authorMadee, Yosef G. Ali
dc.contributor.authorAbdalsmd, Ltfei A.
dc.contributor.authorShufat, Salem A. A.
dc.contributor.authorElhadad, Khaled M.
dc.contributor.authorAlmezogi, Hand Hadia
dc.contributor.authorMansur, Mansur Mohamed Ali
dc.date.accessioned2026-01-02T23:56:25Z
dc.date.issued2017-01-01
dc.description.abstractIn this study, based differential equations methods are used to solve equations because these methods are dependent on boundary value data more than other mathematical equations. We have calculated neutron flux, criticality and geometrical eigenvalue by using multi-group method and solving the neutron diffusion equation for finite and infinite cylindrical and spherical reactors in this study. For the calculation of the total neutron flux cross sections, we need the neutron diffusion equation. Thus, we have established the relationship between neuron flow and cross-section of neuron depending on neutron energy. Critical calculations have been made by comparing the results with MNCP (montecarlo n-partical) simulation methods. For necessary computer calculations, the programme, Wolfram-Matematica-7 has been used.
dc.description.urihttps://doi.org/10.4236/ojapps.2017.72007
dc.description.urihttp://www.scirp.org/journal/PaperDownload.aspx?paperID=74518
dc.description.urihttps://dx.doi.org/10.4236/ojapps.2017.72007
dc.identifier.doi10.4236/ojapps.2017.72007
dc.identifier.eissn2165-3925
dc.identifier.endpage81
dc.identifier.issn2165-3917
dc.identifier.openairedoi_dedup___::5b8b1a17f3f9ec7d56a4eb6c30f8e98d
dc.identifier.startpage68
dc.identifier.urihttps://hdl.handle.net/20.500.12597/36369
dc.identifier.volume07
dc.publisherScientific Research Publishing, Inc.
dc.relation.ispartofOpen Journal of Applied Sciences
dc.rightsOPEN
dc.subject.sdg7. Clean energy
dc.titleEstimates of the Fast and Termal Flux in Blanket of Critical Reactors by Using Multi-Group Methods
dc.typeArticle
dspace.entity.typePublication
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