Abstract
A Fusion Nuclear Science Facility (FNSF) has been recognized in the fusion community as a necessary facility to resolve the critical technology issues of in-vessel components prior to the construction of a DEMO reactor (Abdou et al., 1996) [1]. Among these components, development of a reliable, low-cost and safe blanket system that provides self-sufficient tritium breeding and efficient conversion of the extracted fusion energy to electricity, while meeting all material, design and configuration limitations is among the most important but still challenging goals. In the recent FNSF study in the US (Kesel et al., 2015) [2], a Dual-Coolant Lead-Lithium (DCLL) blanket has been selected as the main breeding blanket concept. This paper summarizes the most important details of the proposed DCLL blanket design, presents the MHD thermohydraulic analysis for the PbLi flows in the blanket conduits and introduces supporting R&D studies, which are presently ongoing at UCLA. We also discuss the required pre-FNSF R&D in the area of MHD Thermofluids to support the further work on the DCLL blanket design & analysis and its integration into the fusion facility.
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