Transport of tritium (T) and heat is calculated to estimate the performance of liquid Li17Pb83 (Li-Pb) or Li2BeF 4 (Flibe) as a T-breeder in a fusion reactor blanket. This bred in such a way of low leak to facilities outside and continuous recovery by a removal system, and heat is transferred through structural walls to He coolant efficiently. In this paper, T permeation in a blanket composed of structural materials and liquid Li-Pb or Flibe is calculated based on data of previous experiment. The effects of T recovery ratio by the outside removal apparatus and fluid-film T diffusion resistance in the liquid blanket on overall T permeation rates are analytically clarified. Design of a liquid blanket with low T leak and high T recovery is discussed here. In addition, possibility in that microbubbles may be generated at interfaces between a liquid blanket and a structural wall is investigated.
All Science Journal Classification (ASJC) codes
- Civil and Structural Engineering
- Nuclear Energy and Engineering
- Materials Science(all)
- Mechanical Engineering