Plasma Shape Research Articles

Pedestal particle balance studies in JET-ILW H-mode plasmas

JET-ILW type I ELMy H-modes at 2.5 MA/2.8 T with constant NBI heating (23 MW) and gas fuelling rate were performed, utilising edge localised mode (ELM) pacing by vertical kicks and plasma shaping (triangularity, δ) as tools to disentangle the effects of ELMs, inter-ELM transport and edge stability on the pedestal particle balance. In agreement with previous studies, the pedestal confinement improves with increasing δ, mostly due to a significant increase in pedestal density while the ELM frequency () is decreased. Improved pedestal confinement with increasing δ was observed even when the pedestal MHD stability was degraded artificially by vertical kicks, implying that increased triangularity may favourably affect the inter-ELM pedestal recovery. The workflow developed to quantify the pedestal particle balance uses high time-resolution profile reflectometry to characterise the inter-ELM evolution of the plasma particle content (), the NEO drift-kinetic solver to evaluate the neoclassical fluxes and interpretative EDGE2D-EIRENE simulations to estimate the edge particle source. The edge particle source is then constrained by deuterium Balmer-α line intensity measurements in the main chamber, which are, however, strongly affected by reflections from the metal walls. The reflections are accounted for by the CHERAB code taking the divertor emission (the brightest light source in the torus) distribution from imaging spectroscopy measurements as input. Our analysis shows that in the second half of the ELM cycle, the volume-integrated particle source is larger than , indicating that transport plays a key role in the inter-ELM pedestal recovery.

Effect of negative triangularity on peeling-ballooning instability

Experiments have achieved high confinement discharges in tokamaks with a negative triangularity (NT) plasma shape accompanied by a lower pedestal and smaller and more frequent edge localized modes (ELMs) compared with positive triangularity (PT). Some existing theories emphasize the linear instability variations result from the change of pedestal. However, NT can directly bring significant changes on magnetic field structures which may also influence the instability of ELMs. Based on a series of equilibria constructed with different triangularities and pressure profiles, the influence of NT on peeling-ballooning mode (P–B mode) is investigated. It is found that NT can increase the growth rates of low to intermediate n (toroidal mode number) modes in the linear stage and lead to a larger pedestal collapse in the nonlinear stage if its pressure profile is the same with the PT shape. Further analyses demonstrate that NT enlarges the unfavorable curvature area, which provides stronger driving source and larger unstable region for the instability. Meanwhile, the diamagnetic effect and local magnetic shear helps to stabilize high n modes in the linear phase, and the E × B shearing rate at the top of the pedestal contributes to suppress the transport of turbulence into the plasma core in the nonlinear phase for the NT shape. What’s more, further simulations with different pedestal heights demonstrate that there exists a threshold value of pressure ratio, below which the ELM energy loss in NT shapes can be smaller than that in PT shapes, suggesting that the smaller energy loss with NT in experiment mainly results from the lower pedestal heigh. The results reveal behaviors of P–B modes and provide possible mechanisms for the phenomenon of lower pedestal height with negative triangularities in experiments.

Development of magnetic control for the EU-DEMO flight simulator and application to transient phenomena

This paper presents a first implementation of DEMO tokamak (Donné, 2019) [1] configuration including CREATE magnetic controllers in the Fenix flight simulator environment. Fenix is based on the 1-D transport code ASTRA (Pereverzev and Yushmanov, 2002 [2] and Fable et al., 2013) [3] , coupled with the 2-D equilibrium solver SPIDER (Ivanov et al., 2005) [4] . Following the successful demonstration on ASDEX Upgrade (Herrmann and Gruber, 2003) [5]) , it was clear that this environment could be fruitfully used to simulate DEMO scenarios including both magnetic and kinetic control strategies. In order to design magnetic controllers, first a systematic validation of the plasma linearized dynamic response has been successfully carried out. This dynamic response has been used to design and tune a first version of magnetic controllers including vertical stability, plasma shape and plasma current control. Simulations of both the plasma flat-top phase including H to L transition, and of the ramp-up phase are presented to show the effectiveness of the proposed simulator and of the magnetic control solution.

An Integrated Design Study for an Advanced Tokamak to Close Physics Gaps in Energy Confinement and Power Exhaust

A high-level design study for a new experimental tokamak shows that advances in fusion science and engineering can be leveraged to narrow the gaps in energy confinement and exhaust power handling that remain between present devices and a future fusion pilot plant (FPP). This potential new U.S. facility, an Exhaust and Confinement Integration Tokamak Experiment (EXCITE), will access an operational space close to the projected FPP performance regime via a compact, high-field, high-power-density approach that utilizes advanced tokamak scenarios and high-temperature superconductor magnets. Full-device optimization via system code calculations, physics-based core-edge modeling, plasma control simulations, and finite element structural and thermal analysis has converged on a T, MA, m, , D-D tokamak with strong plasma shaping, long-legged divertors, and 50 MW of auxiliary power. Such a device will match several absolute FPP parameters: plasma pressure, exhaust heat flux, and toroidal magnetic field. It will also narrow or close the gap in key dimensionless parameters: toroidal beta, bootstrap fraction, collisionality, and edge neutral opacity. Integrated neutron shielding preserves personnel access by limiting nuclear activation and maximizes experimental run time by reducing site radiation. In addition to design study results and optimization details, parameter sensitivities and uncertainties are also discussed.

A diagnostics slim cassette for reflectometry measurements in DEMO: Design and simulation studies

The reflectometry diagnostic proposed for DEMO aims to measure the radial edge density profile and to provide data for the feedback control of plasma position and shape. Microwave (MW) reflectometry measurements are foreseen at 16 poloidal locations (gaps) and require plasma-facing antennas and waveguides (WGs) to route the microwaves through the upper port (UP) to the diagnostic hall. The current integration proposal is based on the Diagnostics Slim Cassette (DSC) concept, a full, 25-cm thick poloidal sector dedicated to house the antennas and WGs. This paper translates the main findings from previous studies into an updated, coherent design of the DSC, while discussing integration aspects not considered before. Neutronics simulations, thermo-mechanical analyses and electromagnetic simulations are employed to evaluate the WG deformation under operating conditions and its effect on the reflectometry measurements. Although the estimated deformations are not expected to compromise the performance of the diagnostic, this work shows that the operating conditions must be considered when optimizing the shapes of the WGs to minimize losses.

Improved FCDI algorithm for tokamak plasma equilibrium reconstruction

The paper presents the improved FCDI plasma equilibrium reconstruction algorithm implemented at the Globus-M2 tokamak. The algorithm has been improved to be able to operate in real time. The FCDI-FF (Fixed Filaments) operation mode was added, in which the plasma is modeled by current filaments at fixed positions. New user interface for the algorithm displays reconstructed plasma separatrix and contours of the internal magnetic surfaces, the poloidal flux distribution, and graphs of the reconstructed plasma parameters, including profiles of plasma pressure, poloidal current and safety factor q, plasma energy, poloidal beta and internal inductance, which can be exported in *.mat, *.json and G-eqdsk formats. Keywords: tokamak, equilibrium reconstruction, plasma shape.

Effect of aspect ratio on plasma response to resonant magnetic perturbations in tokamak devices

A systematic numerical study is carried out, computing and comparing the plasma response to the resonant magnetic perturbation (RMP) field, applied for controlling edge localized modes (ELMs), in a series of tokamak plasmas with varying aspect ratio and utilizing the MARS-F code. The aspect ratio is scanned either by varying the plasma major radius at a fixed minor radius or by varying the latter while fixing the former. Both approaches yield similar results when compared in terms of quantities with proper normalizations. In general, a non-monotonic dependence of the resonant response field (normalized by the vacuum counterpart) near the plasma edge is found with varying aspect ratio, indicating that a given ELM control coil current configuration strongly favors plasmas with a certain aspect ratio. This optimal aspect ratio, on the other hand, depends on the toroidal as well as poloidal (i.e., coil phasing) spectra of the applied RMP field. The equilibrium (edge) safety factor, the plasma shape, and the plasma toroidal flow are all fixed to ensure that the effects identified here are predominantly due to the plasma aspect ratio.

Plasma shape effects on the Alfvén eigenmode spectrum through Alfvén slow-magnetosonic wave coupling

The effect of the plasma shaping (triangularity and elongation) on the continuous spectrum, frequency gaps, and number of eigenmodes is studied. The plasma shaping affects the coupling between Alfvén and slow magnetosonic waves via the geodesic magnetic curvature and the plasma pressure. The Alfvén-slow coupling creates a large number of new gaps in the continuous spectrum below the Alfvén frequency where discrete modes can reside. In ideal magnetohydrodynamic simulations a large number (a few hundred or more) of potential eigenmodes are found. The number of eigenmodes is correlated with the maximum geodesic curvature and a minimum number of possible discrete eigenmodes is found at a negative triangularity of −0.3. It is hypothesized that these possible eigenmodes form a low amplitude and dense discrete spectrum, which can be studied experimentally.

Drift kinetic theory of the NTM magnetic islands in a finite beta general geometry tokamak plasma

In (Imada et al 2019 Nucl. Fusion 59 046016 and references therein) a new 4D drift kinetic nonlinear theory, valid in the limit of a low beta, small inverse aspect ratio, circular cross section, toroidal geometry, to describe the plasma response to the neoclassical tearing mode (NTM) magnetic perturbation is derived. In (Dudkovskaia et al 2021 Plasma Phys. Control. Fusion 63 054001) this theory is reduced in a low collisionality limit, which allows a dimensionality reduction to a 3D problem to efficiently resolve the collisional dissipation layer in the vicinity of the trapped-passing boundary. (Dudkovskaia et al 2021 Plasma Phys. Control. Fusion 63 054001) adopts an improved model for the magnetic drift frequency, which reduces the threshold magnetic island half-width from , where is the trapped ion banana orbit width, to , making it in closer agreement with experimental observations for the large aspect ratio tokamak equilibrium. In the present paper, the theory is extended to a high beta, arbitrary tokamak geometry to capture the plasma shaping effects on the NTM threshold physics with the focus on the non-zero triangularity discharges that are known to have a strong impact on the plasma MHD stability. First, it is found that the higher triangularity plasma is more prone to NTMs which is in agreement with the tearing mode onset relative frequency measurements in DIII-D. Second, the NTM threshold dependence on the tokamak inverse aspect ratio obtained in (Dudkovskaia et al 2021 Plasma Phys. Control. Fusion 63 054001) is refined and extended to a finite aspect ratio limit. Third, the NTM threshold dependence on poloidal beta is obtained and benchmarked against the EAST threshold island width measurements.

Local gyrokinetic simulations of tokamaks with non-uniform magnetic shear

In this work, we modify the standard flux tube simulation domain to include arbitrary ion gyroradius-scale variation in the radial profile of the safety factor. To determine how to appropriately include such a modification, we add a strong ion gyroradius-scale source (inspired by electron cyclotron current drive) to the Fokker–Planck equation, then perform a multi-scale analysis that distinguishes the fast electrons driven by the source from the slow bulk thermal electrons. This allows us to systematically derive the needed changes to the gyrokinetic model. We find new terms that adjust the ion and electron parallel streaming to be along the modified field lines. These terms have been successfully implemented in a gyrokinetic code (while retaining the typical Fourier representation), which enables flux tube studies of non-monotonic safety factor profiles and the associated profile shearing. As an illustrative example, we investigate tokamaks with positive versus negative triangularity plasma shaping and find that the importance of profile shearing is not significantly affected by the change in shape.

Methodology of Plasma Shape Reachability Area Estimation in D-Shaped Tokamaks

This paper suggests and develops a new methodology of estimation for a multivariable reachability region of a plasma separatrix shape on the divertor phase of a plasma discharge in D-shaped tokamaks. The methodology is applied to a spherical Globus-M/M2 tokamak, including the estimation of a controllability region of a vertical unstable plasma position on the basis of the experimental data. An assessment of the controllability region and the reachability region of the plasma is important for the design of tokamak poloidal field coils and the synthesis of a plasma magnetic control system. When designing a D-shaped tokamak, it is necessary to avoid the small controllability region of the vertically unstable plasma, because such cases occur in practice at a restricted voltage on a horizon field coil. To make the estimations mentioned above robust, PID-controllers for vertical and horizontal plasma position control were designed using the Quantitative Feedback Theory approach, which stabilizes the system and provides satisfactory control indexes (stability margins, setting time, overshoot) during plasma discharges. The controllers were tested on a series of plasma models and nonlinear models of current inverters in auto-oscillation mode as actuators for plasma position control. The estimations were made on these models, taking into account limitations on control actions, i.e., voltages on poloidal field coils. This research is the first step in the design of the plasma shape feedback control system for the operation of the Globus-M2 spherical tokamak. The developed methodology may be used in the design of poloidal field coil systems in tokamak projects in order to avoid weak achievability and controllability regions in magnetic plasma control. It was found that there is a strong cross-influence from the PF-coils currents and the CC current on the plasma shape; hence, these coils should be used to control the plasma shape simultaneously.

Revisit of recombination processes of low-charge-state ion generation during picosecond intense laser–gas interaction

The origin of He+ ions in picosecond intense laser–gas interaction was experimentally and numerically investigated. A nanosecond laser pulse was used to modulate the plasma condition, which helps to distinguish the origin of the helium ions with different charge states. By comparing the energy spectra of He2+ and He+ ions in the experiments with/without plasma shaping, we found neither ion–atom nor electron–ion recombination is responsible for the He+ generation, which is supported by two-dimensional hydrodynamic simulations and theoretical calculations.

Triangularity effects on global flux-driven gyrokinetic simulations

On the road to fusion energy production, many alternative scenarios have been investigated in order to address certain well-known problems of tokamak devices; among which, anomalous transport, ELMs and disruptions. The studies on plasma shaping fall into this effort.In particular, it has been experimentally observed that when operating in L mode, negative triangularity (NT) features better confinement properties than positive triangularity (PT). However, even though the trend is quite clear, a complete and satisfying theoretical explanation for this experimental findings is still lacking.With the aim of understanding and describing these improvements starting from first principles, we present the first comparison between PT and NT with global flux-driven gyrokinetic simulations performed with the ORB5 code.The numerical setup includes: electrostatic turbulence, kinetic trapped electrons, non-linear collisional operator, ECRH source, limiter and wall as boundary conditions. The simulations have been performed on ideal MHD equilibria and kinetic profiles inspired by TCV experiments, in a mixed ITG-TEM regime.First analysis reveal a strong reduction of transport in NT; while at the edge PT shows superdiffusivity, NT does not. The limiter plays an important role that has to be further clarified.

Impact of Neutral Beam and Plasma Shaping on the Current Drive Efficiency and Fast Ion Losses in a Compact Tokamak FNS-ST

Fusion neutron source (FNS)-spherical tokamak (ST) will be a FNS based on ST ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$R =0.5$ </tex-math></inline-formula> m, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$a =0.3$ </tex-math></inline-formula> m, <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$k =2.75$ </tex-math></inline-formula> , and <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$B =1.5$ </tex-math></inline-formula> T) with a steady-state neutron yield ~1018 n/s. The noninductive current drive and neutron generation is to be maintained by neutral beam injection (NBI). Beam in toroids (BTOR) code (Python) is used to reproduce NBI and plasma shaping with high degree of detail and to perform parametric studies of neutral beam (NB) driven effects. The beam model light neutral beam (LNB) is inherited from the original beam transmission (BTR) code, used since 2005 for NBI design and optimizations. The BTOR methodology combines NB statistical description with high-performance analytical models of particle tracing, which allows one to get results several orders faster, if compared to the conventional (Monte–Carlo) approach, and still with a good agreement. The results obtained for the FNS-ST device prove that the entire beam efficiency highly depends on the beam–plasma mutual size and shaping, including the beam cross section and aiming, plasma aspect ratio, elongation, and triangularity. The results are especially important for a steady-state current drive and fusion control in low aspect ratio tokamaks; the BTOR method is also applicable to a conventional tokamak design.

Global gyrokinetic study of shaping effects on electromagnetic modes at NSTX aspect ratio with ad hoc parallel magnetic perturbation effects

Plasma shaping may have a stronger effect on global turbulence in tight-aspect-ratio tokamaks than in conventional-aspect-ratio tokamaks due to the higher toroidicity and more acute poloidal asymmetry in the magnetic field. In addition, previous local gyrokinetic studies have shown that it is necessary to include parallel magnetic field perturbations in order to accurately compute growth rates of electromagnetic modes in tight-aspect-ratio tokamaks. In this work, the effects of elongation and triangularity on global, ion-scale, linear electromagnetic modes are studied at National Spherical Torus Experiment (NSTX) aspect ratio and high plasma β using the global gyrokinetic particle-in-cell code XGC. The effects of compressional magnetic perturbations are approximated via a well-known modification to the particle drifts that was developed for flux-tube simulations [Joiner et al., Phys. Plasmas 17, 072104 (2010)], without proof of its validity in a global simulation, with the gyrokinetic codes GENE and GEM being used for local verification and global cross-verification. Magnetic equilibria are re-constructed for each distinct plasma profile that is used. Coulomb collision effects are not considered. Within the limitations imposed by the present study, it is found that linear growth rates of electromagnetic modes (collisionless microtearing modes and kinetic ballooning modes) are significantly reduced in a high-elongation and high-triangularity NSTX-like geometry compared to a circular NSTX-like geometry. For example, growth rates of kinetic ballooning modes at high-β are reduced to the level of that of collisionless trapped electron modes.

Real-Time Magnetic Sensor Anomaly Detection Using Autoencoder Neural Networks on the DIII-D Tokamak

Magnetic diagnostics in tokamaks are key to plasma equilibrium control (plasma current, plasma shape, and position) and amelioration of plasma instabilities. Thus, real-time identification of the anomalous sensor is mandatory. A new real-time system based on autoencoder (AE) neural networks (NNs) for anomaly detection in magnetics signals, including both flux loops and magnetic probes, has been successfully implemented on the plasma control system (PCS) of the DIII-D tokamak. The AE NN is trained on over 4000 plasma discharges, with an optimized latent space representation of the input signals while minimizing the reconstruction loss. An algorithm determining anomalous sensors based on excessive deviations from accurate reconstruction from the trained NNs is constructed in a MATLAB/Simulink environment and is deployed on the real-time PCS using the embedded MATLAB coder (EMC) environment. The reconstruction performance of the algorithm is quantified in hardware-in-the-loop (HIL) simulation by injecting artificially common fault types, including stuck-at-zero fault, saturated fault, noise fault, and drift fault. The AE NN algorithm has been successfully commissioned on the DIII-D PCS.

Sensitivity Analysis of Capital Cost of European DEMO Design

Conceptual designs for a European demonstration (EU-DEMO) power plant are based on extrapolations of physics scaling laws and current understanding of engineering limits based on available technologies. It is imperative to quantify the impact of uncertainties in physics and engineering parameters on the ability to produce an economically attractive future fusion power plant that meets key design outcomes. In this work, the sensitivity of the expected capital cost of an EU-DEMO power plant has been studied using the systems code PROCESS. A systems code aims to model interactions between subsystems of a fusion power plant and provide consistent solution across a large parameter space. The PROCESS system code allows for user-defined initial conditions and constraints and then optimizes using a given figure of merit to find optimal design parameters. We present a sensitivity analysis on optimizations around the 2018 pulsed EU-DEMO baseline, and this allows for the identification of the most consequential model parameters and the magnitude of the nonlinear interactions between them. We consider the pulsed EU-DEMO baseline, and while fixing the major radius and optimizing for fusion gain <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$Q$ </tex-math></inline-formula> , we present a sensitivity analysis of the role of the physics and engineering parameters and constraints in determining the capital cost of such a device. We identify the dominant physics parameter as the power threshold necessary to enter H-mode, which accounts for 45% of the sensitivity and find high interactions between plasma shaping parameters and other power plant subsystems. This analysis allows for the identification of areas of additional technical focus and uncertainty propagation.

Bus Design for the Poloidal Field Coils of the NSTX-Upgrade Fusion Device

The National Spherical Torus eXperiment (NSTX) has undergone a major upgrade to NSTX-U at Princeton Plasma Physics Laboratory (PPPL). NSTX upgrade (NSTX-U) will double the toroidal field, plasma current, and neutral beam injection heating power, as well as significantly increase the pulse duration. NSTX-U uses three poloidal field (PF) coils at the vessel top and three at the bottom near the divertor areas to control the local plasma shape there. These shaping coils operate at a maximum terminal voltage of 2 kV, corresponding to a maximum current about 20 kA. To supply the power to the PF coils, hard copper bus bars are typically used near the vacuum vessel, while water cooled flexible cables are used away from the vessel. The detailed design of the hard copper bus bars will be covered in this article. During operation, these hard bus bars are subject to high electromagnetic (EM) forces, thermal displacement loads, as well as plasma halo loads due to disruption. The EM, thermal, and structural analysis were performed, and the results revealed that, with the worst loads combined, the new design meets the NSTX-U thermal, structural, and fatigue cycle requirements. The manufacturing and installation process of the bus work will be discussed at the end of this article.

Different roles of elongation and triangularity in the H-mode pedestal height enhancement through strong plasma shaping

While the peeling–ballooning mode is known to be stabilized by strong plasma shaping in a high-confinement mode (H-mode), a significant difference is found in the roles of elongation and triangularity. The elongation mainly stabilizes the high-n ballooning branch, while the triangularity of the low-n peeling one with some destabilization of the high-n ballooning one, where n is the toroidal mode number. This difference is conjectured mainly due to the magnetic shear effect which is substantially enhanced in a high triangularity regime, giving a destabilization on the ballooning branch, while a stabilization on the peeling one. An important consequence of these different roles of the two shape parameters is that they should work together to stabilize simultaneously both branches of the PBM, implying the necessity of synergy between them for the large pedestal height enhancement.

Dependence of particle and power dissipation on divertor geometry and plasma shaping in DIII-D small-angle-slot divertor

• The compact SAS divertor in DIII-D provides a good testbed for understanding the effects of a tightly closed divertor on particle and power dissipation, and for application to core-edge integration solutions. • A longer outer leg facilitates the achievement of divertor dissipation in the SAS divertor, while a shorter leg leads to higher divertor temperatures and requires higher upstream densities to achieve the same level of divertor detachment. • The divertor conditions strongly depend on the divertor target shaping and Bt direction. • In-slot gas puffing can reduce detachment onset density by about 10% when the outer strike point is placed at the inner slanted surface. • Completed divertor detachment and a high confinement core with normalized energy confinement factor H 98 >1.0 can be simultanesouly achieved with the SAS divertor, demonstrating the benefit of a closed divertor for exploration of core-edge integration. Dedicated experiments in DIII-D find that magnetic shaping and divertor target geometry significantly affect the divertor plasma conditions and divertor detachment process in the small-angle-slot (SAS) divertor. The compact SAS divertor in DIII-D provides a good testbed for understanding the effects of a tightly closed divertor on particle and power dissipation, and for application to core-edge integration solutions. A longer outer leg facilitates the achievement of divertor dissipation in the SAS divertor, while a shorter leg leads to higher electron temperatures near the divertor target plate and requires higher upstream densities to achieve the same level of divertor detachment. In addition, with the ion B × ∇ B drift away from the SAS divertor and the outer strike point (OSP) near the outer corner, the target temperature is lower for a particular upstream density than with the OSP on a slanted or flat surface, leading to lower heat flux even when the particle flux remains similar. In contrast, with the ion B × ∇ B drift into the SAS divertor, a strike point at the inner slanted surface exhibits a lower upstream density to achieve divertor detachment than a strike point either at the outer corner or the outer slanted target. Experimental results and SOLPS-ITER simulations with full drifts suggest the strong interplay between drift flows and the neutral distribution resulted from target shaping. Furthermore, in-slot gas puffing has been shown to achieve global divertor detachment with an onset density about 10% lower than that using main-chamber gas puffing when the outer strike point is placed at the inner slanted surface. Corresponding modelling reveals that the local gas puffing enhances the neutral ionization which potentially facilitates the achievement of divertor dissipation. However, such improvement diminishes when the strike point is at the outer corner, which also indicates the geometric dependence on divertor performance in the SAS divertor. Even with different strike point locations, complete divertor detachment with very low particle and heat fluxes at the divertor targets and a high confinement core with normalized energy confinement factor H 98 >1.0 can be simultanesouly achieved with the SAS divertor with ion B × ∇ B drift into SAS divertor, demonstrating the benefit of a closed divertor for exploration of core-edge integration.