2024 – Validation of edge fluid codes for degree of detachment of the high-field side divertor + Quantification of Plasma-Molecular Interaction Effects on Divertor Detachment in L-mode and H-mode

Validation of edge fluid codes for degree of detachment of the high-field side divertor + Quantification of Plasma-Molecular Interaction Effects on Divertor Detachment in L-mode and H-mode

2024 Research Campaign, Divertor Science and Innovation

Purpose of Experiment

The primary purpose of the proposed experiment is to quantify the divertor detachment by measuring the degree of detachment and understanding plasma-molecular interactions on the DIIID low-field side (LFS) divertor using (DTS), divertor imaging and spectroscopy measurements. The degree of detachment then further predicted and validated with UEDGE, SOLPS-ITER and EGDE2D-EIRENE. Along with that, through extensive spectroscopic measurements of high-n Balmer lines and molecular bands, the Bayesian inference Spectroscopy Technique for Plasma Molecular Interaction (BaySPMI) codes will be validated at DIII-D tokamak. Plasma ion-molecular interaction gives rise to plasma momentum loss in divertor region that affect the degree of detachment. The processes include Molecular Activated Recombination (MAR) and Molecular Activated Ionization (MAI), which also affect the power and particle balance in detached divertor plasma conditions. Multiple experiments in and simulations of DIII-D and other tokamaks have quantitatively, predictions of the degree of detachment of either divertor leg vary between the edge fluid codes by up to a factor of 5 in the plasma particle and heat fluxes, resulting in significantly different predictions of neutral fueling of the core plasma. Since all three codes are used to predict the SOL in ITER and future power plants, isolating the root cause(s) for these disparate predictions is necessary for credible predictions of the SOL in these future devices.

Experimental Approach

The proposal builds up on a series of previous experiments in the Divertor Science and Innovation area in a DTS optimized, lower single-null configuration, e.g., fwd. BT, 160299 at 2000 ms, 187142 at 1200 ms, rev. BT, 160322 at 2000 ms. In these discharges, the plasma current and toroidal magnetic fields were 1.3 MA and 2.1 T, respectively. To adapt to the new high-field side LHCD limiter, the gap to that limiter is to be increased by 2 cm to maintain large gaps to main chamber surfaces for ease of setting up the edge fluid codes and interpretating their predictions. Here, the 6-cm flux surface is to be limited by the toroidally continuous upper outer limiter. The feasibility of this change was already checked and cleared by the Plasma Control Group (Jason Barr). First plasma discharge will be fueling ramp/density scan to identify the divertor detachment. Second would be also fueling ramp/density by changing squareness of LSN by reproducing shape from #186420/#152835 (For better LFS coverage to LLAMA). After this there would be decision point that, further experiments would be done with changed squareness or not. In both the density ramps and further experiments, the primary method to assess different degrees of detachment is to set the core plasma density via gas injection (GASA/B) to low-recycling (Te at the LFS strike point ≈ 40 eV), high-recycling (2-3 eV) and detached conditions (≈ 0.5 eV). In addition, the density limit is determined by ramping the core plasma density dynamically to high/maximum fueling. Next, in individual discharges, the core plasma density is held constant (gas injection in feedback-mode on core density) and the strike points are swept for approx. 20 cm across the DTS and spectroscopy chords to obtain the 2D Te and ne profiles, and the 1D poloidal emission profiles. To warrant density control, the lower and upper cryo pumps are held at liquid He temperature. One last plasma, if time permits, would be the same plasma with GASC. The primary challenge of this experiment, potentially requiring repeat discharges, is operating at constant core plasma density, in particular for plasmas close to the density limit. The emphasis of this experiment is optimized diagnosis of the plasma and neutral conditions: repeat discharges may be necessary for accidentally saturated measurements, and additional discharges are proposed as backup if the core of the shot plan is completed.

Interested in a behind-the-scenes look at DIII-D? Join us for a virtual OR in-person tour during Fusion Energy Week (May 5-9)! Sign up for a tour here.

X