and Divertor Physics
The ITER plasma boundary will
comprise the thin (~0.2-m thick), relatively low-temperature periphery
of the ITER plasma, plus the magnetically-connected plasma scrape-off
layer and lower poloidal divertor , where the fully-ionized boundary
— including thermalized alpha particles —
will be converted to neutral gas, to be exhausted by pumps external to
the torus vacuum vessel. In addition to this 'fusion ash' helium exhaust
function — essential to the maintenance of a sustained fusion burn
in the plasma core — the ITER divertor plasma will also radiate
a substantial fraction — up to 80% — of the highly localized
plasma thermal power flux that would otherwise impinge on the divertor
target surfaces. The combination of particle (D, T and He) and thermal
power exhaust that the divertor provides gives rise to description of
its functions as being ones of power and particle exhaust.
Region Physics and Modelling
The physics basis for the
power and particle exhaust aspects of the ITER plasma boundary deal
primarily with the characteristics of relatively low-temperature (1-100
the atomic physics of plasma recombination and ionization in this low-temperature
regime and the major role that radiation from hydrogenic and trace impurity
(carbon, oxygen, etc.) species and localized plasma convection plays
attainment and control of the thermal and particle exhaust functions
of the divertor. Divertor studies in present tokamakin present tokamaks
elucidated both the complex plasma and atomic physics involved and
the utility of 2-D numerical boundary simulation models (UEDGE, B2-EIRENE)
which can now accurately reproduce present experimental data and which
provide the design basis for extrapolating such data to the ITER/reactor-tokamak
regime. Operation of the ITER divertor will constitute a final 'reactor-regime'
validation of such modelling.
The inside 'plasma-facing'
portion of the ITER plasma boundary comprises the transition region between
the low-temperature divertor plasma and fusion-producing plasma core.
For the planned 'ELMy H-mode' operation scenario, the transition from
'outer' ~100 eV plasma temperature to the inner ~5 keV 'fusion core'
anticipated to occur over a radial dimension of ~0.1 m. The resulting
plasma temperature and pressure gradients in this 'H-mode edge' or 'H-mode
pedestal' region are high and a unique set of physics basis considerations
applies to the 'self-organizing' development of the strong temperature
gradient and energy transport barrier that is characteristic of the H-mode
and also to the regulation of the H-mode edge pressure gradient by the
periodic MHD instabilities called edge localized modes (ELMs). The inner
temperature of the H-mode edge — the so-called pedestal temperature
— and the magnitude and frequency of ELMs have separate but equally
great import for ITER. Present data show a clear correlation between
pedestal temperature (pressure) and better plasma energy confinement,
and models of ITER performance based on numerical micro-turbulence simulations
in the plasma core predict that an edge temperature of about 5 keV will
be required to obtain the desired fusion power and Q, The same data also
shows that the large, low-frequency Type I ELMs, predicted for this edge
temperature regime, will lead to rapid erosion of the ITER divertor targets.
But at the present time, physics understanding of the pedestal region
and the ability to reliably predict ITER pedestal and ELM properties
arguably less well developed than the corresponding understanding of
the plasma core and divertor regions. Accordingly, the urgent need for
of better understanding of pedestal and ELM physics leads to science
opportunities for present fusion experiments and theory and to unique
for validation of such understanding in ITER.
plasmas with an ITER-like shape and divertor show radiation from the
plasma boundary and divertor. In-divertor radiation is strongly localized
near the plasma x-point and outboard target strikepoint
modelling reproduces the 2-D radiation intensity profiles observed
in the ASDEX Upgrade tokamak divertor
in DIII-D plasmas increases with increasing edge pedestal pressure.
But higher pressures yield large 'Type I' ELMs, which may be problematical
MHD code modelling shows edge 'peeling-mode' structure and radial
depth differences for plasmas with large and small ELMs