Assessment of the neutronic performance of some alternative fluids in a fusion-fission hybrid reactor by using Monte Carlo method
Küçük Resim Yok
Tarih
2013
Yazarlar
Dergi Başlığı
Dergi ISSN
Cilt Başlığı
Yayıncı
Pergamon-Elsevier Science Ltd
Erişim Hakkı
info:eu-repo/semantics/closedAccess
Özet
In this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVFa Ferritic steel structural material and the molten salt-heavy metal mixtures 99-95% Li20Sn80-1-5% SFG-Pu, 99-95% Li20Sn80-1-5% SFG-PuF4, or 99-95% Li20Sn80-1-5% SFG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as neutron flux, tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fissile fuel breeding in liquid first wall, blanket and shield zones and investigates effects of spent fuel grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.O. Published by Elsevier Ltd.
Açıklama
Anahtar Kelimeler
Hybrid reactor, TBR, M, Fissile fuel, MCNPX-2.7.0
Kaynak
Annals of Nuclear Energy
WoS Q Değeri
Q2
Scopus Q Değeri
Q1
Cilt
60