interacting with the material surface would be so low that virtually no material
erosion would be possible [3].
Interestingly, today’s concept of handling the issue of plasma interaction with the
PFCs in magnetic fusion reactors is essentially a symbiosis of these two fundamental
ideas adapted to the particular features of the magnetic devices.
The original Spitzer’s divertor design was focused on a stellarator magnetic
configuration (e.g. see [2]), where, because of the complexity of magnetic geometry,
the formation of divertor was only possible by the reversal of the strong toroidal
magnetic field. Nonetheless, the implementation of such a divertor in B-65
stellarator [4] has demonstrated encouraging results showing a significant impurity
reduction in the core plasma.
In much simpler, toroidally symmetric magnetic geometry of a tokamak, the
divertor (the so-called poloidal divertor) can be relatively easy formed by proper
arrangement of the electric current(Às) in additional (or even existing) magnetic
coils, reversing some relatively weak poloidal magnetic field.
However, experimental studies of the impact of poloidal divertors on tokamak
performance have started only at the end of 1970th. Before that, the most common
approach was the designation of some parts of the tokamak PFC – the “limiters” - to
handle the plasma-material interaction. For example, Fig. 1.3a shows the sketch of a
toroidally symmetric limiter which is designated to accommodate the most severe
plasma interaction with the PFCs. The so-called “last closed magnetic flux surface”
(LCFS) separates the nested, closed magnetic flux surfaces in the core, which are
Fig. 1.3 (a) Sketch of a toroidally symmetric limiter which is designated to accommodate the most
severe plasma interaction with the PFCs. The LCFS that separates the nested closed magnetic flux
surfaces in the core, which are occupied by hot, fusion grade plasma, from the open ones where the
magnetic field lines intersect the material targets; and (b) Schematic view of the poloidal crosssection of a tokamak with a poloidal divertor formed by the plasma current and the currents in
toroidal magnetic coils. In both figures, Q SOL is the power coming from the core into the SOL, the
red and blue arrows show the directions of the poloidal magnetic field and the orange arrows
indicate the direction of the heat flux
1 Edge Plasma Issues in Magnetic Fusion Devices
3
erosion would be possible [3].
Interestingly, today’s concept of handling the issue of plasma interaction with the
PFCs in magnetic fusion reactors is essentially a symbiosis of these two fundamental
ideas adapted to the particular features of the magnetic devices.
The original Spitzer’s divertor design was focused on a stellarator magnetic
configuration (e.g. see [2]), where, because of the complexity of magnetic geometry,
the formation of divertor was only possible by the reversal of the strong toroidal
magnetic field. Nonetheless, the implementation of such a divertor in B-65
stellarator [4] has demonstrated encouraging results showing a significant impurity
reduction in the core plasma.
In much simpler, toroidally symmetric magnetic geometry of a tokamak, the
divertor (the so-called poloidal divertor) can be relatively easy formed by proper
arrangement of the electric current(Às) in additional (or even existing) magnetic
coils, reversing some relatively weak poloidal magnetic field.
However, experimental studies of the impact of poloidal divertors on tokamak
performance have started only at the end of 1970th. Before that, the most common
approach was the designation of some parts of the tokamak PFC – the “limiters” - to
handle the plasma-material interaction. For example, Fig. 1.3a shows the sketch of a
toroidally symmetric limiter which is designated to accommodate the most severe
plasma interaction with the PFCs. The so-called “last closed magnetic flux surface”
(LCFS) separates the nested, closed magnetic flux surfaces in the core, which are
Fig. 1.3 (a) Sketch of a toroidally symmetric limiter which is designated to accommodate the most
severe plasma interaction with the PFCs. The LCFS that separates the nested closed magnetic flux
surfaces in the core, which are occupied by hot, fusion grade plasma, from the open ones where the
magnetic field lines intersect the material targets; and (b) Schematic view of the poloidal crosssection of a tokamak with a poloidal divertor formed by the plasma current and the currents in
toroidal magnetic coils. In both figures, Q SOL is the power coming from the core into the SOL, the
red and blue arrows show the directions of the poloidal magnetic field and the orange arrows
indicate the direction of the heat flux
1 Edge Plasma Issues in Magnetic Fusion Devices
3