Power flattening of an inertial fusion energy breeder with mixed ThO2-UO2 fuel

Neutronic performance of a blanket-driven ICF (inertial confinement fusion) neutron and based on SiCf/SiC composite material has been investigated for fissile fuel breeding and a flat fission power density. The blanket is fueled with ThO2 and UO2 mixed by various mixing methods for achieving a flat fission power density, and cooled with different coolants, natural lithium, (LiF)2BeF2, Li17Pb83 and 4He for the nuclear heat transfer. MCNP4B code is used for calculations of neutronic data per DT fusion neutron. M (fusion energy multiplication rate) increases to ∼2.6 in the 4He coolant blanket. On the other hand, this value is ∼2.00 in the Li17Pb83 coolant blanket because of more capture reactions than fission reactions in this blanket. Therefore, the investigated reactor can produce substantial electricity in situ. Total fissile fuel breading ratio (FFBR) (232Thγ+238Uγ) is increased from ∼0.16 to ∼0.37 by using coolants having high neutron multiplier cross-section. Values of TBR (tritium breeding ratio) being one of the main neutronic parameters for a fusion reactor are greater than 1.05 in all case of mixed fuels and type of coolants. The highest and the lowest TBR values are ∼1.34 and ∼1.07 in the natural Li and the 4He coolant blanket, respectively. Consequently, the breeder hybrid reactor is self-sufficient in the tritium required for the DT fusion driver. Γ (peek-to-average fission power density ratio) of the blanket is reduced to ∼1.1. Furthermore, quasi-constant fission power density profile is obtained in FFB (fissile fuel breeding) zone for case of PMF (parabolic mixed fuel) and all type of coolants. Values of neutron leakage out of the blanket in all case of mixed fuels and type of coolants are quite low due to SiC reflector and B4C shielding. The maximum neutron leakage is only ∼0.025