Neutronic analysis of lithium hydride (LiH) material IN A (D, T) driven hybrid blanket

Abstract

In this study, neutronic performance of Lithium Hydride (LiH) material is analyzed in a D‐T driven hybrid blanket cooled by flibe (Li₂BeF₄). The hybrid blanket is fuelled by UO₂ from LWR fuel rod, LWR spent fuel rods and CANDU spent fuel rods. Energy production, tritium breeding, neutron leakage and fissile fuel breeding are considered. Volume fractions are selected as 1, 2, 3, 4 and 5. LiH thickness is increased from 0 to 80 cm. The number of rows is selected as 10 to 20. When the volume fractions increase, TBR values decrease. When the LiH thickness reaches 50 cm, TBR values reach the point of saturation. At this thickness, TBR values of Model‐ II (P. 4) are higher than those of Model‐I (P. 4). The M energy multiplication factor has nearly the same tendency in both models. Neutron leakage values of Model‐II (for DR_LiH = 50 cm) are lower than that Model‐I. Although the fissile fuel breeding rate values of Model‐I and Model‐II are almost the same, the values of Model‐I are a little higher than Model‐II. Therefore, it is concluded that LiH material is suitable for using, when neutronic behaviour is considered.

Key words:

• fusion blanket
• LiH material
• tritium