Design and Research and Development of Front-Face Remote Handling Targets for CFETR Divertor

The Chinese Fusion Engineering Testing Reactor (CFETR) divertor development is ongoing by further optimizing a water-cooled engineering design scheme and implementing the research and development of small-and medium-sized test mock-ups and prototype fabrication. To increase the tritium breeding ratio (TBR), the design scheme adopts a hybrid divertor–blanket integration concept where a breeding blanket is placed underneath targets instead of a cassette. The novel front-face remote handling (RH) targets are designed aiming to explore whether the targets could be maintained from the plasma side. Three unique independent RH design schemes are developed separately for the inner, dome, and outer targets. Hypervaportron flat-tile units are optimized in the high heat flux (HHF) area with an engineering requirement of 20 MW/m $^{2}$ steady-state heat load. All the targets have good RH compatibility with RH tools in the maintenance procedure. Preliminary analysis with respect to pumping performance, hydraulics, heat removal ability, and structure meets the requirements. The research and development is based on advanced fusion reactor materials and realizable materials’ connection technology. The advanced materials of potassium-doped tungsten (KW), oxide dispersion-strengthened copper (ODS-Cu) alloy, and reduced activation ferritic/martensitic (RAFM) steel are used as armor, heat sink, and structural materials, respectively. One of the realizable material connection routes is that RAFM steel is explosively welded with ODS-Cu and then the KW/Cu flat tiles are brazed with ODS-Cu. The small-sized mock-up successfully endured 1000 cycles with 20 MW/m $^{2}$ steady-state heat load. The maximum temperature of the KW surface was 882 $^{\circ}$ C, much lower than the supposed recrystallization temperature. The ultrasonic testing (UT) showed no defects exceeding $\Phi $ 2 mm equivalent before/after the testing. The medium-sized mock-up showed that the UT results meet the requirements and the profile error of the plasma-facing surface is less than 1 mm. All this work will be beneficial for the final engineering design of the CFETR divertor.