3642

3441

كارشناسي ارشد

فيزيك

اصفهان: دانشگاه صنعتي اصفهان ، دانشكده فيزيك

1386

نه ،93،[II]ص.مصور،جدول ،نمودار

ص.ع.به فارسي و انگليسي

فيزيك

ID3441

به فارسي و انگليسي: قابل رويت به صورت نسخه ديجيتال

Abstract In this Thesis transport of neutrons in the Isfahan Miniature Neutron SourceReactor MNSR was simulated by method of Monte Carlo using the MCNPcode At the first stage of this simulation the exact geometry of the reactor corewas introduced to the code via the INP File and the Epithermal Index r wascalculated at the critical state This quantity was compared with the measuredvalues and showed good agreements In the next stage the relative neutron fluxwas calculated at the inner and outer irradiation sites of the reactor In this work the relative neutron flux in the inner and outer irradiation sites ofthe reactor were also measured by the method of Neutron Activation Analysis NAA using some In and Au foils The calculated results were then comparedwith the measured values and showed good agreement in both thermal andepithermal regions

احمد شيراني

ايرج شهابي

محمد حسن علامت ساز