Analysis of Small Modular SMART Reactor Core Fuel Burn up using MCNPX code | ||
| Arab Journal of Nuclear Sciences and Applications | ||
| Volume 57, Issue 3, July 2024, Pages 35-43 PDF (700.03 K) | ||
| Document Type: Original Article | ||
| DOI: 10.21608/ajnsa.2024.290367.1819 | ||
| Author | ||
| Moustafa Abdelwahab Aziz* | ||
| Nuclear Safety Center, Egyptian Atomic Energy Authority, Cairo, Egypt | ||
| Abstract | ||
| SMART Nuclear reactor, which is conceptually developed by KAERI (Korea Atomic Energy Research Institute), is a Small modular reactor of 330 Mwth. MCNPX computer code Package which based on Monte Carlo method is used to model the reactor core and evaluate neutronic characteristics of the core. Reactor multiplication factor is evaluated with time. Fuel burnup and depletion of fissile 235U and breeding of plutonium are calculated with fuel burnup along the life cycle of the reactor core. Radial and axial Power and flux mapping distributions are calculated along all fuel assemblies of the core. Delayed neutron fraction, prompt neutron life time as well as fuel , moderator and void temperature coefficient of reactivity are evaluated and analysed. The results indicated that average core burnup extends to 27 GWd/T after 3 operation years which corresponds to 235U burnup ration of 71 %. Power distribution is compared to previously published results and satisfactory agreements were found | ||
| Keywords | ||
| SMART Reactor; MCNPX Code; Neutronic Analysis; Fuel Burnup | ||
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