Overview of the US ITER Dual Coolant Lead Lithium (DCLL) Test Blanket Module Program

With the US rejoining ITER, the US chamber technology community has resumed participation in discussion in the ITER Test Blanket Working Group (TBWG) and has proposed to develop, in collaboration with other parties, liquid and solid breeder blanket concepts to be tested in ITER. Presently, the US focus on the liquid breeder option is the dual coolant helium-cooled reduced activation ferritic steel structure with self-cooled Pb-17Li breeder (DCLL) that uses flow channel insert (FCI) as the MHD and thermal insulator. When projected for a reference tokamak power reactor design, it has the potential for a gross thermal efficiency of >40%. The US is planning for an independent test blanket module (TBM) that will occupy half an ITER test port with corresponding supporting ancillary equipment. An initial design, testing strategy and corresponding test plan have been completed for the DCLL concept. The DCLL TBM conceptual design for the integrated testing phase, including the choice of configuration, relevant design analyses, ancillary equipment, testing strategy and corresponding test plan, have been prepared for the transition into the preliminary design phase