The calculation of neutron flux using Monte Carlo method
Küçük Resim Yok
Tarih
2017
Yazarlar
Dergi Başlığı
Dergi ISSN
Cilt Başlığı
Yayıncı
E D P Sciences
Erişim Hakkı
info:eu-repo/semantics/openAccess
Özet
In this study, a hybrid reactor system was designed by using 99-95% Li20Sn80 + 1-5% RG-Pu, 99-95% Li20Sn80 + 1-5% RG-PuF4, and 99-95% Li20Sn80 + 1-5% RG-PuO2 fluids, ENDF/B-VII. 0 evaluated nuclear data library and 9Cr2WVTa structural material. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion-fission hybrid reactor system. The neutron flux was calculated according to the mixture components, radial, energy spectrum in the designed hybrid reactor system for the selected fluids, library and structural material. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.
Açıklama
3rd International Conference on Theoretical and Experimental Studies in Nuclear Applications and Technology (TESNAT) -- MAY 10-12, 2017 -- Adana, TURKEY
Anahtar Kelimeler
Breeder Apex Reactor, Radiation-Damage, Fusion Breeder, Structural Material
Kaynak
3rd International Conference on Theoretical and Experimental Studies in Nuclear Applications and Technology (Tesnat 2017)
WoS Q Değeri
N/A
Scopus Q Değeri
Cilt
154
