Session CO06: Magnetic Confinement: Divertor & Scrape-Off-Layer

Live

We generalize the Heuristic Drift (HD) model of the scrape-off layer width, taking into account both the enhanced SOL confinement time at high collisionality and the increase in upstream temperature at very low collisionality. We find that there is a wide range of upstream separatix density over which the original HD model is applicable, but at high collisionality the SOL widens, in agreement with experimental data from ASDEX-Upgrade and JET. We further find that for typical low-gas-puff H-mode conditions, the projected \textit{ExB} flow shearing rate in the SOL dominates over the interchange growth rate, while for typical L-Mode conditions, and at the high densities where H-Modes return to L-Mode, the interchange growth rate dominates. The result may be related to that of Halpern and Ricci (2017) with respect to the steep gradient region of the SOL in inner-wall-limiter discharges. Taking $\omega_{\mathrm{s}}$ \textgreater $\gamma _{\mathrm{int}}$ as the criterion for the H-Mode, we can use the generalized HD model to predict the scaling of the H$\to $L Mode transition. We find a stronger scaling of power with density, approximately squared, than the conventional L$\to $H Mode threshold. The values for existing machines, and for ITER, are somewhat below the L$\to $H Mode predictions, consistent with significant hysteresis, especially at lower densities.   

Live

The BOUT$++$ simulations with the attached divertor conditions for H-mode discharges well follow the Heuristic-Drift-based (HD) divertor heat flux width scaling of inverse dependence on the poloidal magnetic field in the drift dominant regime. However, both ASDEX-Upgrade data and the generalized HD (GHD) model showed that the scrape-off width broadens as the density/collisionality increases. A series of BOUT$++$ transport simulations are performed to study the physics of the scaling characteristics of the divertor heat flux width vs density/collisionality via a plasma density scan. The simulations show that even in the drift dominated regime, the divertor heat flux width can be broadened due to the transition of the SOL residence time from the parallel particle flow time to the enhanced parallel conduction time as the collisionality/density increases as posited in the GHD model. In addition, the heat flux width is found to be proportional to the square root of ion mass for low collisionality while it has a weakly dependence on ion mass for high collisionality. Furthermore, our simulations show that as the density increases, the radial electric field (Er) well shallows, which potentially weakens ErxB flow shear stabilization of turbulence at high density.   

Live

BOUT$++$ has been developed and applied for a range of problems that impact on boundary plasma fluctuation and resulting the SOL power width scaling. A summary of simulation progress and results will be presented including, but not limited to: (1) Simulating the DIII-D and EAST grassy ELM regime; (2) Analysis of edge turbulent transport and divertor heat load for ITER hybrid scenario; (3) Prediction of divertor heat flux width for ITER pre-fusion power operations; (4) Impact of plasma density/collisionality on divertor heat flux width; (5) Divertor power loads during thermal quench; (6) Deep learning surrogate model for kinetic Landau-fluid closure with collision. Simulation results show that the peak heat flux decreases while the corresponding width increases as the SOL fluctuation-driven transport increases due to larger turbulent fluxes ejected from the pedestal into the SOL when operating in either a small and grassy ELM regime, type-I ELMs or during thermal quench or in a high-density regime. In addition to the enhanced fluctuation-driven radial transport, the SOL power width can also be broadened due to the transition of the SOL residence time from the parallel particle flow time to the parallel conduction time in a high-density operation.   

Live

Present day SOL scalings can be explained by a simple model linking the SOL width to magnetic drifts. The question, then, is if this scaling trend will persist. To this end, it is natural to model the SOL as a \textit{lossy thermal boundary layer} (BL) which is driven by core heat flux, and which balances drift transport, turbulent transport and parallel losses. A puzzle here is that the SOL is turbulent but also remarkably stable --- with FLR, line-tying and drift excursion (analogous to finite banana width) acting to weaken or quench the usual suspects for turbulence generation. The logical deduction is that SOL turbulence originates \textit{inside} the separatrix and subsequently enters the SOL by ``turbulence spreading''. Thus, the SOL BL is seen to be driven by \textit{both} a flux of turbulence intensity as well as heat, both emanating from the core. Note that these are in principle \textit{independent}, thus constituting two separate control parameters for the SOL. Also, since the turbulence flux is determined by pedestal dynamics and turbulence, it can introduce the appearance of ``nonlocality'' to SOL transport. Spreading effects on the SOL width are under study and will be discussed.   

Experimental and Modeling Study of the Divertor Heat Flux Width on EAST (PhD Oral-24)

A comprehensive study of the divertor heat flux width is carried out on the experimental advanced superconducting tokamak (EAST). Experimentally, factors like plasma current, heating scheme and plasma operating regime are found to have significant effects on the divertor heat flux width on EAST. Edge plasma fluid codes BOUT$++$ and SOLPS are employed to simulate the EAST discharges to figure out the potential reasons to the effects of these factors on the divertor heat flux width on EAST. Detailed results and analysis will be presented in the “PhD Oral-24” section.   

Live

Magnetically confined fusion plasmas with a negative triangularity ($\delta )$ core shape are known to feature enhanced confinement as compared to standard, D-shaped plasmas. Recently, correlation electron cyclotron emission measurements on the TCV tokamak revealed that the confinement improvement is accompanied by reduced temperature and density fluctuations across most of the confined plasma. In this contribution, we extend these studies towards the edge/SOL region. Fluctuations in this region are measured with a newly commissioned Gas Puff Imaging diagnostic, a reciprocating probe, and wall Langmuir probes, and for triangularities in the range -0.7\textless $\delta $\textless $+$0.68 in both limited and diverted ohmic L-mode plasmas. These measurements reveal a strong reduction in SOL fluctuation at sufficiently negative $\delta $ plasmas ($\delta $\textless -0.25), and, surprisingly, an almost full suppression of plasma interaction with the first wall. Reasons for this suppression, which could have important implications for the prospects of negative $\delta $ as a reactor solution, are explored, pointing towards the role of reduced connection length intrinsic to negative $\delta $.   

Live

A new SAS-V divertor configuration will be explored in DIII-D to optimize detachment for both Bt directions by leveraging the strong synergy between the neutral recycling benefits of the SAS geometry and the effects of ExB drifts. SOLPS-ITER modeling finds that for the favorable Bt direction, ExB drifts can circumvent the benefits of closed divertor configurations by carrying particles out of the closed outer divertor and into the inner divertor. A V-shaped geometry near the outer strike point is found to mitigate this effect by: (a) increasing neutral recycling at the wall of the slot in the private flux region (PFR) due to -- and causing -- strong radial ExB ion flux from the divertor scrape-off layer to the PFR; (b) decreasing ExB loss of ions out of the outer divertor into the inner divertor via the PFR due to reduction of the radial gradient of electron temperature at the outer target caused by the increased particle retention in the outer divertor. This work points to a promising divertor optimization path to explore for power exhaust in fusion reactors.   

Live

The divertor detachment is one promising solution to reduce the heat flux on the target. Usually the detachment is obtained by increasing the upstream density but it results in deteriorated core confinement. Hence it is necessary to devise a mechanism that can onset the detachment at low upstream densities. From numerous previous studies, it has been found that the magnetic configuration and PFCs material affect the detachment phenomenon. However researches on the effect of divertor shape has been rarely reported. In order to investigate the effect of divertor shape on detachment, we analyzed the effects of SAS-like shape on the KSTAR environment using SOLPS-ITER package. In the SAS-like divertor, highly concentrated deuterium was observed in the SOL region near the separatrix. The increased volumetric power loss by the deuterium near the separatrix lead to significantly lowered heat flux on the target and the detachment onset at considerably low upstream density. Interestingly such reduction of heat load could be obtained with a very shallow SAS compared to the original DIII-D. We discovered that the heat reduction was more sensitive to the slot angle than the depth.   

Live

We propose a new understanding of how the radial electric field ($E_r$) impacts the edge magnetohydrodynamic (MHD) instabilities. The analysis uncovered that $E_r$-shear stabilizes the Peeling-Ballooning modes, while $E_r$-curvature destabilizes the low-n kink/peeling modes. The underlying physical mechanism is that the perturbed radial velocity and displacement become stronger dephasing or phase pinning. More specifically, the ratio of $E_r$-curvature to $E_r$-shear could be measured to quantify their relative competition strength.   

Live

Prediction of divertor heat flux width is performed for the PFPO-I and PFPO-II scenarios in the new ITER Research plan. Both fluid transport and turbulence codes under BOUT++ framework are used to capture the physics of PFPO-II phase on different temporal scales. Two-fluid turbulence code is used to study SOL turbulence dynamic and corresponding transport. Transport simulations are also performed for the PFPO-I phase as comparison. The initial plasma profiles inside the separatrix are taken from CORSICA scenario studies. Transport coefficients in transport code are calculated by inverting the plasma profiles inside the separatrix and are assumed to be constants in SOL. An anomalous thermal diffusivity scan is performed with E x B and magnetic drifts. The results in both scenarios identify two distinct regimes: a drift dominant regime when diffusivity is smaller than the respective critical diffusivity and a turbulence dominant regime when diffusivity is larger than it. The critical diffusivity is 0.5 m$^{2}$/s in 5MA PFPO-I scenario and 0.3 m$^{2}$/s in 7.5MA PFPO-II scenario. The ITPA multi-machine experimental scaling yields a lower limit of the width. By fixing q and T$_{sep}$, the critical diffusivity is \chi$_{c}$ \propto A$^{\(1/2}$/(Z(Z+1)$^{\(1/2}$Bp$^{2}$).   

Live

A metric for divertor closure in terms of the percentage of neutrals escaping the divertor is presented. The effect of divertor closure on neutral leakage is investigated using the SOLPS code in attached and detached divertor conditions for DIII-D open and closed divertors. At the detachment onset the population of neutrals escaping the divertor is 12{\%} for the closed and 35{\%} for the open divertor. The energy transport in the two geometries is analyzed in depth and it is demonstrated that detachment at lower upstream density found for the closed divertor is due to the following physics mechanisms: the reduction of the convective flow, the increase of the pressure loss and the increase of the power loss. The increase of power loss with closure is due more to hydrogenic emission than the inherent carbon impurity radiation [1]. The results of this work emphasizes that the effect of divertor closure to detachment is not just the divertor neutral trapping but also a shift in the parallel pressure balance between upstream and downstream [2]. 1] L. Casali et al. Contrib. Plasma Phys. 58(7-8), 725-731 (2018), 2] L. Casali et al.  Nucl. Fusion  60 076011 (2020).   

Live

The anticipated power increase in ITER and other future tokamaks will increase the power flux on the divertor plates during transient events beyond a critical value that causes surface vaporization, ionization, and development of a secondary plasma from the divertor materials. The generated secondary plasma will evolve around the strike point, intensifying, expanding into the SOL, and intercepting the remaining incoming transient disrupting and ELM particles. The collisions of the incident core particles with the secondary dense plasma and the deflected/scattered energetic particles under the strong modified magnetic field structure can lead to significant energy transfer to nearby component surfaces. Our comprehensive HEIGHTS 3D simulation using the initial magnetic field equilibrium configuration of exact ITER design showed intense heat flux on the outboard strike point for wide range of ELM parameters causing formation of mini-plasma. Subsequently, the incoming core plasma particles, collide and scatter from this dense plasma, transfer part of their kinetic energy in this evolving secondary plasma. It follows to the changing of hydrodynamic evolution and depositing the remaining energy as particle showers on various nearby components causing unexpected damage.   

On Demand

An impact of neutrals on anomalous edge plasma transport and zonal flow (ZF) is considered. As an example, it is assumed that edge plasma turbulence is driven by the resistive drift wave (RDW) instability. It is found that the actual effect of neutrals is not related to a suppression of the instability \textit{per se}, but due to an impact on the ZF. Particularly, it is shown that, whereas the neutrals make very little impact on the linear growth rate of the RDW instability, they can largely reduce the zonal flow generation in the nonlinear stage, which results in an enhancement of the overall anomalous plasma transport. Even though only RDW instability is considered, it seems that such an impact of neutrals on anomalous edge plasma transport has a very generic feature. It is conceivable that such neutral induced enhancement of anomalous plasma transport is observed experimentally in a detached divertor regime, which is accompanied by a strong increase of neutral density.