Neutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport code
Küçük Resim Yok
Tarih
2014
Yazarlar
Dergi Başlığı
Dergi ISSN
Cilt Başlığı
Yayıncı
Carl Hanser Verlag
Erişim Hakkı
info:eu-repo/semantics/closedAccess
Özet
In this study, the molten salt-heavy metal mixtures 93-85 % Li20Sn80 + 5 % SFG-PuO2 and 2-10% UO2, 93-85% Li20Sn80 + 5 % SFG-PuO2 and 2-10% NpO2, 93-85% Li20Sn80 + 5 % SFG-PuO2 and 2-10% UCO were used as fluids. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library.
Açıklama
Anahtar Kelimeler
Structural Material, Apex Reactor, Radiation-Damage, Fusion, Wall
Kaynak
Kerntechnik
WoS Q Değeri
Q4
Scopus Q Değeri
N/A
Cilt
79
Sayı
2
