Numerical neutronic analysis of a natural lithium cooled fusion breeder fueled with UO2
Küçük Resim Yok
Tarih
2000
Yazarlar
Dergi Başlığı
Dergi ISSN
Cilt Başlığı
Yayıncı
KING FAHD UNIV PETROLEUM MINERALS
Erişim Hakkı
info:eu-repo/semantics/closedAccess
Özet
The fissile breeding capability of a catalyzed-(D,D) and (D,T) fusion-fission (hybrid) reactor, fueled with spent UO2, is analyzed under first wall fusion neutron load of 5 MW/m(2) to provide nuclear fuel for LWRs. This can be a prospective alternative to the existing methods of nuclear fuel enrichment. Lithium (Li) and lithium beryllium mixtures are chosen for the nuclear heat transfer out from the fissile fuel-breeding zone. The behavior of the UO2 fuel is observed during the 48 months for discrete time intervals of Delta t = 15 days and over a plant factor of 75%. Calculations show that a residence time of 12 to 42 months in possible for spent UO2 kept inside a fusion-fission reactor so as to accumulate cumulative fissile fuel enrichment values that would meet an acceptable quality level for deployment of the irradiated UO2 as fuel in LWRs. Enrichment grades between 4.5% and 6.5% can be achieved during a plant operation over four years depending on the type of fusion driver and coolant. In all types, the tritium breeding ratio (TBR) exceeds unity. Therefore, the blanket is self-sufficient with respect to tritium breeding. Mathematical models have been established for important nuclear engineering criteria depending on the type of fusion driver and coolant in spent fuel rejuvenation.
Açıklama
Anahtar Kelimeler
Kaynak
ARABIAN JOURNAL FOR SCIENCE AND ENGINEERING
WoS Q Değeri
Q4
Scopus Q Değeri
N/A
Cilt
25
Sayı
2A
