Divertors

To illustrate the usefulness of such an algorithm, and the seriousness of the issue of thermal conductivity degradation to the design and operation of PFCs, the algorithm discussed above has been used to construct Fig. 9 [34], which shows the isotherms for a monoblock divertor element in the unirradiated and irradiated state and the "flat plate" divertor element in the irradiated state. In constmcting Fig. 9, the thermal conductivity saturation level of 1 dpa given in Fig. 8 is assumed, and the flat plate and monoblock divertor shown are receiving a steady state flux of... 

Fig. 9. Temperature contour for irradiated and unirradiated graphite divertor tiles.

Table 1. Materials and heat loads for the major fusion machines world wide (see Section 1.3 for definitions of divertor and first wall)... 

Fusion Machine Country Fuel System First Wall Heat Load (MW/m2) First Wall Material Divertor or Limiter Heat Load (MW/m2) Divertor of Limiter Material... 

ASDEX Axially Symmetric Divertor Experiment Max Planck Institute for Plasma Physics, Garching, Germany... 

Theoretical estimates of the erosion of the first wall and plasma contamination due to sputtering requires a knowledge of particle and photon fluxes to the wall, as well as data on the erosion yields. Sputtering will be discussed in a later part of this chapter. Here we shall briefly summarize some of the calculations done on primary fluxes in future fusion reactors. The calculations are rather uncertain because of the poor understanding of various parameters such as divertor efficiency, refueling, neutral beam heating, plasma temperature and density profiles, including the scrape-off layer in the case of divertor operated Tokamaks. 

Hydrogenic ion-induced blistering has been observed, but by itself will not likely present a problem to the first wall at elevated reactor operating temperatures. The effects of blistering at high primary fluxes such as those found at beam dumps and components of neutral beam injectors, divertor throats and bombardment plates, are not known. 

As discussed above, blistering will also occur at the divertor throat and particle bombardment plates, and will therefore be important for future reactors employing these devices. Blistering of neutral beam injector components must also be considered and therefore blistering should be considered a potential problem in a broader sense than has traditionally been accepted. 

The previous discussion has shown that plasma impurities present a complex set of problems whose solution is crucial to the successful operation of fusion reactors. The many and often subtle factors that govern plasma-materials interactions are still only partially understood. Consequently, the methods used today to control impurities are to a degree empirical in nature and cannot yet be precisely defined. It is likely that in the end a variety of approaches will be used to keep plasma impurities at minimal levels. The techniques are conveniently divided into divertor and non-divertor methods. The latter depend on modifications of one sort or another of the composition or structure of the surfaces facing the plasma. As will be seen in Sect. 6.5., meth-... 

In order to reduce significantly the impurity influx into the plasma they have to leave the wall as ions (See Sect. 6.5.) or be ionized within the scrape-off layer before reaching the separator (see Fig. 7). The latter denotes the magnetic surface separating the magnetic lines which are closed within the confinement region and those which penetrate into the divertor chambers (Figs. 7 and 19). 

Fig. 19. A schematic view of a-poloidal and 6-bundle divertor (see text)...

Since the plasma is in part pumped away by the divertor, efficient refueling is necessary in order to keep a constant plasma density. An adequate refueling method has to maintain a sufficiently low rate of production of charge exchange neutrals. The possible effects of the divertor on the MHD-stability of Tokamak plasmas and a detailed understanding of the processes in the scrape-off layer require further investigations. 

Experimental measurements on deuterium trapping in solid metal targets5 1S2-1S6) have shown that this may be a promising technique for pumping a useful fraction of the ion flux in the divertor. 

Fig. 21. Schematic cross-section of a divertor-equipped Tokamak showing some of the possible fates of a particle ejected from the wall as a neutral (a),...

The control of impurity release and transport requires a better understanding of the complex phenomena of plasma-wall interactions including the processes occuring in the scrape-off layer in the limiter shadow. In order to establish the feasibility of suggested solutions such as divertors or surface modifications, experiments have to be performed not only in the laboratory but also in-situ in fusion devices. The latter... 

---