A moving belt divertor concept with applications to ITER

For near-term fusion devices such as ITER (the International thermonuclear Experimental Reactor), the divertor system represents one of the most challenging problems in the design of the entire machine. During normal operation the divertor surface must remove the plasma transport power (∼ 100 MW) conducted to it along the magnetic field lines. Also, the divertor surface acts to neutralize the impinging plasma ions for subsequent removal by the vacuum pumping system. This paper reports on the current ITER reference divertor design which illustrates the difficulty of designing a feasible divertor. The initial phase of ITER, the Physics Phase, calls for a carbon-carbon (C-C) composite armor tile brazed to a molybdenum- or copper-alloy heat sink, which is water-cooled at 60 degrees C and 3. 5 MPa. For this design, separatrix sweeping is required if the peak plasma heat flux is higher than 15 MW/m2, to maintain adequate burnout margins and armor thickness. Other critical design issues include the neutron-induced thermal conductivity degradation in carbon, the retention of tritium and carbon codeposited on the divertor surface, and the risk of severe damage by runaway electrons. Particularly crucial issues are the fatigue life and reliability of the thousands of brazes joining the C-C armor and the heat sink material, because the loss of a carbon tile will lead to the exposure of the metallic heat sink to the plasma. The performance of the overall tokamak is therefore strongly linked to the performance of the divertor system and its ability to effectively remove a large amount of heat while receiving an enormous particle flux