In this study, a neutronic performance of the Laser Inertial Confinement Fusion Fission Energy (LIFE) molten salt blanket is investigated. Neutronic calculations are performed by using XSDRNPM/SCALE5 codes in S-8-P-3 approximation. The thorium molten salt composition considered in this calculation is 75 % LiF-25 % ThF4, 75 % LiF-24 % ThF4-1 % (UF4)-U-233, 75 % LiF-23 % ThF4-2 % (UF4)-U-233. Also, effects of the Li-6 enrichment in molten salt are performed for all heavy metal salt. The radiation damage behaviors of SS-304 structural material with respect to higher fissionable fuel content and Li-6 enrichment are computed. By higher fissionable fuel content in molten salt and with Li-6 enrichment (20 and 50 %) in the coolant in form of 75 % LiF-23 % ThF4-2 % (UF4)-U-233, an initial TBR > 1.05 can be realized. On the other hand, the 75 % LiF-25 % ThF4 or 75 % LiF-24 % ThF4-1 % (UF4)-U-233 molten salt fuel as regards maintained tritium self-sufficiency is not suitable as regards improving neutronic performance of LIFE engine. A high quality fissile fuel with a rate of similar to 2,850 kg/year of U-233 can be produced with 75 % LiF-23 % ThF4-2 % (UF4)-U-233. The energy multiplication factor is increased with high rate fission reactions of U-233 occurring in the molten salt zone. Major damage mechanisms in SS-304 first wall stell have been computed as DPA = 48 and He = 132 appm per year with 75 % LiF-23 % ThF4-2 % (UF4)-U-233. This implies a replacement of the SS-304 first wall stell of every between 3 and 4 years.