Estimates of the Fast and Termal Flux in Blanket of Critical Reactors by Using Multi-Group Methods
Yazarlar (8)
Prof. Dr. Aybaba HANÇERLIOĞULLARI Kastamonu Üniversitesi, Türkiye
Prof. Dr. Aslı KURNAZ Kastamonu Üniversitesi, Türkiye
Yosef G Ali Madee
Ltfei A Abdalsmd
Salem A A Shufat
Khaled M Elhadad
Hand Hadia Almezogi
Mohamed Ali Mansur
| Makale Türü | Özgün Makale (Uluslararası alan indekslerindeki dergilerde yayınlanan tam makale) | ||
| Dergi Adı | Open Journal of Applied Sciences | ||
| Dergi Tarandığı Indeksler | ISI Web of Knowledge | ||
| Makale Dili | İngilizce | Basım Tarihi | 02-2017 |
| Cilt / Sayı / Sayfa | 7 / 2 / 68–81 | DOI | – |
| UAK Araştırma Alanları |
Nükleer Fizik
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| Özet |
| In 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. |
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