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# 28th IAEA Fusion Energy Conference (FEC 2020)

10-15 May 2021
Virtual Event
Europe/Vienna timezone
The Conference will be held virtually from 10-15 May 2021

## Experiment and modelling of divertor detachment with deuterium injection in KSTAR H-mode plasmas

11 May 2021, 14:00
4h 45m
Virtual Event

#### Virtual Event

Regular Poster Magnetic Fusion Experiments

### Speaker

Junghoo Hwang (Korea Advanced Institute of Science and Technology)

### Description

For future fusion devices including ITER and DEMO, it is crucial to handle the power flux to the divertor targets in order not to exceed the steady-state material limit, q$_{t,⊥}$≤ 10 Wm$^{-2}$ [Ref. 1]. One of the viable effective methods to achieve it under development is divertor detachment via impurity gas seeding or deuterium injection in the divertor region in order to dissipate the excessive power flux. In KSTAR, L-mode detachment studies have been performed extensively, and the in/out divertor asymmetry was found where the detachment occurred at a lower upstream electron density at the outer target than at the inner target [Ref. 2], as opposed to the observations in other tokamaks. In a recent dedicated experiment, the first step toward divertor detachment in H-mode was successfully done by pure deuterium fueling and the study on the asymmetric characteristics of detachment has been extended to more complicated H-mode plasmas. This paper reports on the results of the experiment and numerical modelling using the SOLPS-ITER code to get better understanding of physical processes behind divertor detachment.
The discharges in the H-mode detachment study had a plasma current of I$_{p}$ = 0.7 MA and a forward toroidal magnetic field of B$_{T}$ = 2.0 T with 8 s flattop (1.5 s – 9.5 s). Shot traces of plasma parameters for the experiment (#22849) are shown in figure 1. The total heating power was about 3 MW by neutral beam injection. The outer strike point was swept for 2 – 4 cm during the current flattop phase in order to obtain target ion saturation current (j$_{sat}$) profiles by the tile-embedded Langmuir probe array. D$_{\alpha}$ line emissivity near the outer target (Fig. 1(e)) gradually increased with some periodic fluctuations which are due to the strike point sweeping. The deuterium fueling rate was linearly increased, such that the line averaged density $\bar{n}_{e}$ kept increasing up to 6x10$^{19}$ m$^{-3}$ (f$_{GW}$ ≈ 0.7) during the discharge before the H-L back transition. Consequently, the divertor regime was changed from low recycling to high recycling and then detached states. A partial detachment was achieved on both inner and outer divertor targets during the inter-ELM phase and the similar behavior to L-mode was found in terms of the in/out divertor asymmetry. The peak particle flux at the outer target reduced by a factor of 4 compared to the maximum value shown in the attached divertor regime, although there was no significant reduction at the inner target as illustrated in figure 2. This asymmetry that was generally found in the detachment experiment in KSTAR was explained by the neutral particle dynamics and geometrical effect through the SOLPS-ITER modelling without particle drifts [Ref. 2].
Meanwhile, drifts are considered as one of the driving factors of the asymmetry along with the geometry. Thus, we have performed the SOLPS simulations of L-mode plasmas including drifts to check how the drifts influence the asymmetry. The modeling setup is basically same as in [Ref. 2] but with diamagnetic drift and ExB drift turned on. The multiplying factor to the drift terms was scanned in parallel to the separatrix density scan. According to the preliminary result, the drifts modify the target profiles including particle flux density, electron density and temperature, and their effects are larger at the inner target than at the outer target.
Simulations using the SOLPS-ITER code have been performed for the H-mode discharges with pure deuterium injection. The input power into the simulation domain was 2.4 MW by taking into account that 20% of the heating power is radiated inside the core plasma. The fueling rate was scanned from 2.5x10$^{21}$ s$^{-1}$ to 1.0x10$^{22}$ s$^{-1}$ to mimic the density ramp experiment. Figure 3 presents the parallel particle flux density onto the targets obtained from the modelling, and it indicates that the modelling successfully reproduces the anomalous asymmetry as observed in the experiment. Larger reduction of the particle flux is shown at the outer target than at the inner target. However, contrary to the experiment, the attached heat flux to the outer target well exceeds 1 MWm$^{-2}$ which is rarely seen from the divertor infrared television (IRTV) system in KSTAR. The analysis on the total radiation profiles in the divertor region was also demonstrated. As the separatrix density increased, the maximum radiation front moved upstream from the targets to the X-point and the radiation finally became concentrated at the X-point in the detached state. Refinement of the radial profiles of the perpendicular transport coefficients is underway which will give better quantitative agreement between the simulation results and the experimental data.

Acknowledgements
This work was supported by the National R&D Program through the National Research Foundation of Korea (NRF) funded by the Ministry of Science and ICT (NRF-2019M1A7A1A03087560).

References:
(1) A. Loarte $\textit{et al}$, Nucl. Fusion 47, S203-63 (2007).
(2) J.-S. Park $\textit{et al}$, Nucl. Fusion 58, 126033 (2018).

Country or International Organization Korea, Republic of Korea Advanced Institute of Science and Technology (KAIST)

### Primary author

Junghoo Hwang (Korea Advanced Institute of Science and Technology)

### Co-authors

Jae-Sun Park (ITER Organization) Richard Pitts (ITER Organization) Hyungho LEE (National Fusion Research Institute) Jun Gyo Bak (National Fusion Research Institute) June-Woo Juhn (Seoul National University) Dr Juhyeok Jang (National Fusion Research Institute) Yoon Seong Han (Korea Advanced Institute of Science and Technology) Suk-Ho Hong (National Fusion Research Institute) Wonho Choe (Korea Advanced Institute of Science and Technology)