9.3 Divertor Plasma Detachment
In the first subsection, we already discussed detached plasma regimes. However, that
was toroidally and poloidaly symmetric detachment of plasma situated on “closed”
magnetic flux surfaces, where plasma interaction with material surfaces
(e.g. limiters, main chamber wall) is driven largely by cross-field transport. Here
we consider detachment of plasma situated in the divertor volume on “open”
magnetic field lines intersecting the divertor targets. In this case, plasma interaction
with the divertor target material is mainly driven by plasma transport along the
magnetic field lines.
There are different ways to define the “depth” of divertor plasma detachment.
Here we will call the divertor plasma detached when there is a rollover of the ion flux
to the divertor target, similar to that shown in Fig. 1.5a. Note that experimental data
show that similarly to poloidaly symmetric plasma detachment, divertor plasma
detachment does also occur at plasma densities close to the density limit (e.g. see
[25]).
Although some signatures of divertor plasma detachment were observed a long
time ago (e.g. see Fig. 8 in [35]), intensive study of such regimes became one of the
focal points of the magnetic fusion research only since 1990 (see the corresponding
references in [25]). The interest to the detached divertor regime was driven by the
need to reduce the power loading on the divertor targets in future tokamak-reactors,
including ITER, to a tolerable level. The only way to do this is to re-radiate a
significant fraction of the power, generated in the reactor, with impurity. However,
the concept of strong impurity radiation from the core plasma has two issues. First,
due to peculiarities of impurity cross-field transport in the core, strongly radiating,
high-Z impurities (recall Fig. 2.14) have a tendency of accumulation in the very core
plasma, cooling it down and reducing the rate of fusion reactions. The radiation from
low-Z impurities that are less prone to accumulation in the core plasma is rather
weak, so for a sizeable effect, the concentration of the low-Z impurity must be high,
with the corresponding dilution of the fusion D-T plasma and reduction of its
performance. Secondly, a strong reduction of the heat flux from the core to the
edge can prevent the transition to the improved confinement regime (H-mode),
which may be needed for self-sustained burning of the fusion plasma. Therefore, it
is widely accepted that it is necessary to increase power dissipation by impurity
radiation in the divertor region as much as possible. There are two possible ways to
do this: (i) to increase the plasma/impurity density in the divertor region (recall that
the density of the impurity radiation loss is proportional to the product of the electron
and impurity densities) and (ii) to increase the divertor volume by implementing
so-called advanced divertor geometry (see Fig. 1.6). It is very likely that in practice
both ways will be combined.
However, an increase of the divertor plasma density will likely result in the
increase of the plasma flux to the target and to the increase of the so-called
“irreducible” power flux to the target, associated with the deposition of the internal
energy of electron-ion pair – the ionization potential. Estimates made for ITER have
9.3 Divertor Plasma Detachment
241
In the first subsection, we already discussed detached plasma regimes. However, that
was toroidally and poloidaly symmetric detachment of plasma situated on “closed”
magnetic flux surfaces, where plasma interaction with material surfaces
(e.g. limiters, main chamber wall) is driven largely by cross-field transport. Here
we consider detachment of plasma situated in the divertor volume on “open”
magnetic field lines intersecting the divertor targets. In this case, plasma interaction
with the divertor target material is mainly driven by plasma transport along the
magnetic field lines.
There are different ways to define the “depth” of divertor plasma detachment.
Here we will call the divertor plasma detached when there is a rollover of the ion flux
to the divertor target, similar to that shown in Fig. 1.5a. Note that experimental data
show that similarly to poloidaly symmetric plasma detachment, divertor plasma
detachment does also occur at plasma densities close to the density limit (e.g. see
[25]).
Although some signatures of divertor plasma detachment were observed a long
time ago (e.g. see Fig. 8 in [35]), intensive study of such regimes became one of the
focal points of the magnetic fusion research only since 1990 (see the corresponding
references in [25]). The interest to the detached divertor regime was driven by the
need to reduce the power loading on the divertor targets in future tokamak-reactors,
including ITER, to a tolerable level. The only way to do this is to re-radiate a
significant fraction of the power, generated in the reactor, with impurity. However,
the concept of strong impurity radiation from the core plasma has two issues. First,
due to peculiarities of impurity cross-field transport in the core, strongly radiating,
high-Z impurities (recall Fig. 2.14) have a tendency of accumulation in the very core
plasma, cooling it down and reducing the rate of fusion reactions. The radiation from
low-Z impurities that are less prone to accumulation in the core plasma is rather
weak, so for a sizeable effect, the concentration of the low-Z impurity must be high,
with the corresponding dilution of the fusion D-T plasma and reduction of its
performance. Secondly, a strong reduction of the heat flux from the core to the
edge can prevent the transition to the improved confinement regime (H-mode),
which may be needed for self-sustained burning of the fusion plasma. Therefore, it
is widely accepted that it is necessary to increase power dissipation by impurity
radiation in the divertor region as much as possible. There are two possible ways to
do this: (i) to increase the plasma/impurity density in the divertor region (recall that
the density of the impurity radiation loss is proportional to the product of the electron
and impurity densities) and (ii) to increase the divertor volume by implementing
so-called advanced divertor geometry (see Fig. 1.6). It is very likely that in practice
both ways will be combined.
However, an increase of the divertor plasma density will likely result in the
increase of the plasma flux to the target and to the increase of the so-called
“irreducible” power flux to the target, associated with the deposition of the internal
energy of electron-ion pair – the ionization potential. Estimates made for ITER have
9.3 Divertor Plasma Detachment
241