Radiative divertor experiments in DIII-D with D₂injection

Abstract

Deuterium gas injected into ELMing H mode divertor discharges in the DIII-D tokamak typically reduced the total power at the divertor target ~2 times and the peak heat flux ~3 to 5 times with modest (<10%) degradation in plasma energy confinement. The parameter range for the discharges investigated was: Ip=1.0-2.0 MA, q95 approximately= 2.4-6.0 and total input power (≲20 MW. Most of this reduction in heat flux occurred at the sudden formation of a high density, highly radiating region located between the outboard divertor separatrix strike point and the X point. This divertor behaviour is associated with a `partially detached' divertor plasma condition, which is referred to in this paper as the partially detached divertor (PDD) regime. With the onset of the PDD, typically at a line averaged density of 0.6 to 0.7 times the Greenwald density limit, an abrupt reduction in plasma electron pressure (≳4 times) was observed at the outboard divertor separatrix strike point; at the same time, however, only a modest (≲30%) change in the electron pressure was observed upstream near the outboard midplane separatrix. The data suggest that significant plasma momentum loss occurred between the high density, highly radiative region and the (downstream) divertor separatrix target. Plasma performance showed little degradation with the onset of the PDD regime. Deuterium injection made only modest changes in the temperature and density profile shapes near the midplane separatrix of the main plasma. The PDD approach is shown to be compatible with discharges operating at low safety factor (i.e. q95 equivalent to 2.9) and to be effective in significantly reducing toroidal asymmetry in observed divertor plasma properties (e.g., heat flux). The potential for operating in a steady state has been demonstrated using feedback control of the neutral pressure outside the main plasma

ARIES-RS divertor system selection and analysis

“…In this estimation, the expected bremsstrahlung and line radiation losses in the core, and the up/down and inboard/outboard distributions of scrape-off layer power flow are consistent with the experimental database from ASDEX, PDX and DIII-D double-null operation [1]. The in/out power balance at the divertor is consistent with the single-null ELMing H-mode results from DIII-D [2].…”

Density limits in toroidal plasmas

“…ASDEX [9] and JET [84] report complete detachment at densities just below the limit. In DIII-D, detachment is seen at 0.6-0.7 x n G [85], with some weak dependence on input power, n DETACHMENT ~ P 0. 15 .…”

NSTX Report on FES Joint Facilities Research Milestone 2010

“…Figure N1-3. Note that the heat flux profiles in the inner divertor are often broad, due to a phenomenon known as partial detachment 12,13 The final part of NSTX research involved obtaining a comprehensive dataset in strongly shaped discharges with high δ and elongation, as these are of interest to NSTXUpgrade (Design: I P = 2 MA, B t = 1 T, P NBI = 10 MW, 5 sec pulse length) in particular and future spherical tokamaks in general. This involved fine scans in I p , toroidal field B t , neutral beam heating power P NBI , radial separation between the two X-points at the outer midplane, δ r sep , and measurements in a poloidal cross-section shape that matches a scaled C-Mod and DIII-D shape 19,20 .…”