Abstract
The radial width of the exhaust heat flux flowing in the SOL of DIII-D is found to expand at high input power and plasma density, consistent with MHD ballooning stability limits. At low heating power, ~3 MW, the SOL width remains constant and consistent with established empirical scaling dependent only on the midplane poloidal field. At high heating power, ~ 13 MW a higher separatrix density, and resulting higher separatrix pressure is required for divertor detachment. The separatrix pressure gradient at the separatrix continues to increase with density until it begins to saturate at levels ~50% above the calculated ideal MHD ballooning limit. Examination of the separate contributions to the pressure gradient from electrons and ions reveals the ion pressure gradient to saturate more strongly than the electron pressure gradient. Potential analysis issues leading to the measured pressure gradient exceeding the ballooning limit are discussed. At high density, particularly for detached divertor plasmas, the SOL width for temperature and density expand modestly, ~30–50%. The divertor plasma density profile in detachment also reflects this trend, expanding in the radial direction a factor of 2–3. Despite the SOL width expansion at the highest power and density no degradation of the pedestal and resulting core confinement is observed with the additional density at high power. These results imply a more favorable scaling for divertor heat flux control in future reactor-scale tokamaks than predicted by existing empirical scaling.
| Original language | English |
|---|---|
| Article number | 100869 |
| Journal | Nuclear Materials and Energy |
| Volume | 25 |
| DOIs | |
| Publication status | Published - Dec 2020 |
| MoE publication type | A1 Journal article-refereed |
Funding
This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Award(s) DE-FC02-04ER54698 and DE-AC52-07NA27344. This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Award(s) DE-FC02-04ER54698 and DE-AC52-07NA27344.
