ITER tokamak devices

AP/Wide World Photos. Reproduced by permission p. 192 Cutaway of the ITER tokamak device, photograph. Photo Researchers, Inc. Reproduced by permission p. 193 Svedberg, Theodor, portrait. The Library of Congress p. 195 Synge, Richard... 

A cutaway of the ITER tokamak device. A goal of the ITER project is to create sustainable energy by means of fusion. 

Fusion has already been achieved in several devices, but not beyond the break-even point, where the amount of energy produced is the same as the amount consumed. Much basic research is still required and is the focus of a number of international collaborative efforts. As discussed in Chapter 4, foremost among these efforts is the International Thermonuclear Experimental Reactor (ITER), which will be a scale-up of the Princeton Tokamak Fusion Test Reactor shown in Figure 19.17. 

A major challenge to experiments is the understanding of erosion mechanisms in the complex environment of a fusion device. Predictions about erosion in ITER by numerical modeling are based on related experimental findings in tokamaks, on laboratory measurements (e.g., ion beams or linear plasma devices) and the theoretical understanding of underlying mechanisms. Most experimental data on erosion yields in tokamaks have been obtained from... 

The complex environment of a tokamak does not only contain carbon and tungsten. In devices which apply boronization we find also boron. JET evaporates routinely beryllium in the main chamber leading to a flow of beryllium into the divertor. A similar behaviour we also expect from ITER. Moreover, as in all vacuum devices, there is always some residual oxygen in the machine. For a reliable prediction of erosion and deposition it is necessary to obtain knowledge about sources, transport and deposition of all those species in a fusion device. 

Fig. 2.3. Principle of 2D projection in edge modeling of axisymmetric devices, such as ideal tokamaks (pictures from an ITER coil construction site). Atomic and surface processes are most relevant in the lower section ( divertor ) of this cross-section (marked in middle part of upper figure). Also shown typical plasma flow field from 2D edge modelling in this divertor region (bottom). See also figures in Sect. 2.3.2...

Co-deposition entails the formation of hydrogenated carbon layers via the redeposition of eroded C atoms and C-containing molecules/radicals in combination with the fuel H, on both plasma facing and out of line-of-sight surfaces in the device. Such layers have been observed to exceed tens of tm (e.g., in TFTR [26,27]), much thicker than the ion-implantation region, which only extends tens of nm. In addition, the co-deposited layer does not appear to have a limit to its thickness. The H/C ratio in the co-deposited layers is similar to that seen in the implantation zone, viz, 0.4 at room temperature. Based on experience with current tokamaks [7, 28, 29] and predictions for ITER [3], most of the T in an ITER-type machine is expected to be trapped in co-deposits. Consequently, the removal of such layers has recently become a high priority issue. This will be addressed in Sect. 10.4. 

Tritium is a very sensitive subject for public acceptance of fusion and will play a central role in the operation of a next-step experimental fusion facility, which will routinely use large amounts of tritium as fuel (e.g., 100 times more in ITER than in present experiments) in a mixture with deuterium. Tritium retention is a regulatory issue since the amount that can potentially be released in an accident sets the limits on plasma operation without removal. Fuel economy has never been an issue in deuterium-fuelled experiments and only recently have the limitations associated with the use of tritium, and its incomplete recovery in experiments in TFTR and in JET, brought the issue of fuel retention under closer scrutiny [56,57]. Table 12.3 provides a list of key quantities related to tritium in existing tokamaks and a next-step device [18,57-59]. 

ITER. This will remain a major difficulty unless experimentally validated in tokamaks with impurities and relevant wall materials to provide a realistic test-bed which would closely mirror options proposed for the next-step device (e.g., beryllium walls and carbon and/or tungsten divertor proposed for ITER). Such experiments would indeed help answer questions including the magnitudes of erosion and tritium co-deposition, dust formation in the vessel, the ease of tritium removal from mixed-materials, as well as operational aspects (e.g., of using beryllium on the first wall). 

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