Fusion with divertor

The requirements for long pulse operation in the next step fusion device ITER and beyond, like acceptable power exhaust, peak load for steady state, transient loads, sufficient target lifetime, limited long term tritium retention in wall surfaces, acceptable impurity contamination in central plasma and efficient helium exhaust, depend on complex processes. The input to the numerical codes, which are used for the optimization of divertor and wall components, relies to a large extend on our understanding of the major processes related to erosion and deposition, tritium retention, impurity sources and erosion processes. The reliability of predictions made with these codes depends crucially on the accuracy of the atomic and plasma-material interaction data available. 

Tungsten is presently one of the favorite metals for a PFC in a fusion reactor as it even exceeds the properties of molybdenum concerning melting and sputtering. In [59] the inner wall and divertor were covered with tungsten-coated carbon tiles without a serious confinement in plasma opera-... 

Additionally, the incident impurity ions in a fusion device will be multiply charged, e.g., a charge state of 4 can be assumed for Be, C, O, and even higher values for W ions. This will result in increased acceleration of the ions in the sheath potential such that the most probable energies for multiply charged ions in a divertor plasma with Te = 10 eV will be around 200 eV, i.e., well above the threshold energy. 

Next-step D-T burning fusion reactors, such as the International Thermonuclear Experimental Reactor (ITER), will require several kilograms of tritium [1,2]. While most of the tritium will be contained in the fuel process loop, the interaction of the plasma with plasma-facing components (first-wall armour, limiters, and divertors) will lead to accumulation of tritium in the torus. Based on the amounts and distribution of D retention in TFTR and... 

Simulating erosion and re-deposition processes in fusion devices lead to a better understanding of the processes involved. The 3-dimensional Monte-Carlo code ERO-TEXTOR [35,36] has been developed to model the plasma-wall interaction and the transport of eroded particles in the vicinity of test limiters exposed to the edge plasma of TEXTOR. Important problems concerning the lifetime of various wall materials (high Z vs. low Z) under different plasma conditions and the transport of eroded impurities into the main plasma can be treated with the ERO-TEXTOR code. Recently, the divertor geometries have been implemented to carry out simulations for JET, ASDEX and ITER [37], In addition, first attempts have been made to simulate erosion and re-deposition processes in the linear plasma device PISCES to analyze the effect of beryllium. 

Even if using special fusion materials, it cannot be expected that the blanket of the fusion power plant would survive the full lifetime of the reactor. At present, it is estimated that the blanket and the divertor will have to be replaced every 2-5 years. This is a major intervention, which has to be done completely with remote handling operations, as the materials will be strongly activated. It has been found that the engineering design of the blanket system is extremely important and will largely determine the time needed for blanket replacement (Ihli et al. 2007), and through this the availability of the power plant. 

Existing tokamak fusion reactor designs with magnetic divertors have therefore exclusively poloidal field or multipole divertors. But using versions with multipoles installed inside, as in ASDEX, it has not yet been possible to find a feasible solution compatible with reactor access, manufacture and service requirements. 

Resume POLOIDAL FIELD SYSTEM FOR COMMERCIAL FUSION REACTORS WITH MULTIPOLE DIVERTOR COILS ARE NOT YET FEASIBLE. 

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