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DIII-D experiments have achieved promising core–edge integrated plasma scenarios which combine a high-temperature low-collisionality pedestal (pedestal top temperature T _e _,ped > 0.8 keV and collisionality ν * _ped < 1) with a partially detached divertor by leveraging the benefits of a low-density-gradient pedestal in a closed divertor. It is found that with a closed divertor and high heating power, strong gas puffing to achieve detachment moves the peak density gradient outward with respect to the maximum gradient of electron temperature and reduces the density gradient at the pedestal region, which correlates with shallow pedestal fuelling due to the closed divertor geometry. In high-current plasmas in particular, the pedestal top density is found to change little with gas puffing while the separatrix, density increases to allow access for divertor detachment. The separation between density and temperature pedestals results in a high- η _e well above the electron-temperature-gradient stability threshold. Electron turbulence is found to be enhanced in the pedestal and correlated with high η _e resulting from the pedestal shift. The pedestal is wider than the EPED scaling. A revised empirical width scaling is derived based on the combination of EPED scaling with η _e and highlights the important role of additional turbulence on the pedestal structure. The wide temperature pedestal facilitates the achievement of a high-temperature, low-collisionality pedestal and high global performance. Simultaneously, the outward shift of the density pedestal facilitates access to detached divertor conditions with low temperature and heat flux towards the target plate. This approach may be promising for closing the core–edge integration gap for future fusion reactors, which may have a weak-gradient density pedestal due to the highly opaque boundary plasmas.

The transition from small edge-localized modes (ELMs) to large ELMs has been repetitively observed in minute-scale long-pulse high-confinement mode (H-mode) discharges during the 2017 EAST campaign. The appearance of large ELMs is found to be strongly correlated with the decrease in separatrix density due to the gradual decrease in fuel recycling during long-pulse H-mode operations (LPHOs). By the numerical scanning of separatrix density with a fixed temperature profile, it has been found that the dependence of ELM instability on separatrix density is related to the competition between the ion diamagnetic stabilizing effect and destabilizing effect of pressure gradient and current density in the pedestal region. This sheds light on a comprehensive understanding of the different roles of separatrix density in ELM instability observed during EAST experiments. With a high separatrix density, the ideal ballooning mode can be destabilized near the separatrix, which is thought to help achieve small ELMs in EAST LPHOs. During the 2021 EAST campaign, an experiment of large ELM control was performed through actively changing fuel recycling by moving the strike point location on the lower tungsten divertor target plate. It has been demonstrated that the mitigation of large ELMs is strongly correlated with the significant increase in separatrix density, which is thought to be attributed to a higher ionization source in the scrape-off layer (SOL) region by SOLPS-ITER simulation. The high ionization source in the SOL region is believed to provide a strong fueling effect near the separatrix and thus raise the local density, which is considered an important reason for triggering ballooning instabilities near the separatrix and achieving small ELMs.

To address the needs for a fusion pilot plan design, DIII-D/EAST joint experiments on DIII-D have demonstrated high normalized beta β _N ∼ 4.2, toroidal beta β _T ∼ 3.3% with q _min > 2, q _95 ⩽ 8 sustained for more than six energy confinement times in high poloidal beta regime. The excellent energy confinement quality ( H _98y2 ∼ 1.8) is achieved with an internal transport barrier at high line-averaged Greenwald density fraction f _Gr > 0.9. The trapped gyro-Landau fluid (TGLF) modeling of the transport characteristics shows that the beam-driven rotation does not play an important role in the high confinement quality. The modeling also captures very well several transport features, giving us confidence in using integrated modeling to project these experimental results to future machines. The high-performance phase is terminated by fast-growing modes triggered near the n = 1 ideal-wall kink stability limit. New radio frequency (RF) capabilities for off-axis current drive could remove the residual ohmic current to achieve a fully non-inductive state, and improve the mode–wall coupling to increase the ideal-wall β _N limit, enabling sustainment of the fully non-inductive high performance plasma in stationary conditions.

Integration of transient and steady-state divertor heat fluxes control with a high-performance core is necessary for future fusion reactors. In recent DIII-D high- ${\beta _{\text{p}}}$ experiments, divertor detachment and simultaneous edge localized mode (ELM) suppression are demonstrated while the plasma confinement quality is maintained high in ITER-similar shape. By optimizing the neon injection in high- ${\beta _{\text{p}}}$ scenario with ITER-similar shape, deep detachment and ELM suppression are achieved with a high-performance core ( ${\beta _{\text{N}}}$ ∼ 2.8, ${\beta _{\text{p}}}$ ∼ 2.3) at ${q_{95}}$ ∼ 7.5. Partial divertor detachment and suppression of large ELMs are achieved at ${q_{95}}$ ∼ 6. The stability analyses suggest that with low neon injection, the density pedestal becomes higher and steeper and the ${T_{\text{i}}}$ profile also increases, therefore the increased edge pressure and higher current density destabilize the Peeling-Ballooning mode (PBM), which would lead to a large ELM collapse. With strong neon gas puffing, the significantly reduced pedestal pressure and current density, due to the degraded ${T_{\text{e}}}$ pedestal, lead to the stabilization of PBM and ELMs are suppressed. For both cases, the coupling between the large radius internal transport barrier (ITB) and edge pedestal is the key reason for maintaining high global performance. The formation of large radius ITB compensates for pedestal degradation. Such results could provide an attractive scenario to well control the transient and steady-state heat flux onto the divertor plates while maintaining good plasma performance, which is an important step toward the steady-state operation of future fusion reactors.

Experiments performed during strongly-shaped high-power diverted negative triangularity (NT) experiments in DIII-D achieved detached divertor conditions and a transient-free edge, showcasing the potential for application of NT to a core-edge integrated reactor-like scenario and providing the first characterization of the parametric dependencies for detachment onset. Detached divertor conditions will be required in future devices to mitigate divertor heat fluxes. Access to dissipative divertor conditions was investigated via an increase in upstream density. Detachment onset at the outer strike point was achieved with H-mode level confinement $H_{98-y2}\sim 1$ and reactor-relevant normalized pressures $\beta_\mathrm N \sim 2$ . Confinement degradation was observed with deeper detachment, associated with the loss of an electron temperature pedestal. Differences in geometry, radial transport, impact of cross field drifts are discussed to explain differences in access to detachment in NT discharges. Higher normalized densities, with respect to equivalent discharges in positive triangularity, were necessary to achieve detachment, partially explained by the shorter parallel connection length to the targets. The effect of cross-field particle drifts ( E × B , B $\times \nabla$ B ) on access to detachment was demonstrated by the lower upstream density needed to access detachment with ion B $\times \nabla$ B drift directed outside of the active divertor (Greenwald fraction $f_{\mathrm{Gw}}\sim$ 0.9–1.0 vs $f_{\mathrm{Gw}}\sim$ 1.3). The upstream density at detachment onset was observed to increase linearly with plasma current with ion B $\times \nabla$ B drift into the divertor, consistent with the observed narrowing of the scrape-off layer heat flux width λ _q . Edge fluid simulations capture separatrix densities needed to achieve detachment in NT plasma and their dependence on drift direction. The ability to reproduce detachment dynamics in NT plasma increases the confidence in future design studies for NT divertors.

Experiments in DIII-D demonstrate that the upstream plasma density to detach an un-pumped slot divertor is similar for a V-shaped and a flat-end slot, despite significantly higher neutral pressure in the V-shaped slot and in contrast to SOLPS-ITER predictions. The detachment threshold can be reduced by using in-slot instead of main-chamber gas fuelling or by placing the strike point on the inner slanted slot baffle instead of the slot end, as described by simulations with full drift physics. When increasing the plasma line-averaged density (without extrinsic impurities), the transition to detachment in DIII-D slot divertor is sharp and requires a high value of plasma density with the ion $\overrightarrow {\mathbf{B}} {{ \times }}\nabla B$ drift into the slot, whereas it is smooth and requires a lower value of plasma density with the opposite drift direction, in accord with detachment experiments in the DIII-D open lower divertor. Unique experiments on DIII-D and comparison to advanced simulations expand the scientific understanding of slot-shaped divertors, considered highly desirable for next step fusion devices.

The integration of a high-performance core and a dissipative divertor, or the so-called ‘core–edge integration,’ has been widely identified as a critical gap in the design of future fusion reactors. In this letter, we report, for the first time, direct experimental evidence of electron turbulence at the DIII-D H-mode pedestal that correlates with the broadening of the pedestal and thus facilitates core–edge integration. In agreement with gyrokinetic simulations, this electron turbulence is enhanced by high η _e ( η _e = L _n / L _T _e , where L _n is the density scale length and L _T _e is the electron temperature scale length), which is due to a strong shift between the density and temperature pedestal profiles associated with a closed divertor. The modeled turbulence drives significant heat transport with a lower pressure gradient that may broaden the pedestal to a greater degree than the empirical and theoretically predicted pedestal width scalings. Such a wide pedestal, coupled with a closed divertor, enables us to achieve a good core–edge scenario that integrates a high-temperature low-collisionality pedestal (pedestal top temperature T _e,ped > 0.8 keV and a pedestal top collisionality ν * _ped < 1) under detached divertor conditions. This paves a new path toward solving the core–edge integration issue in future fusion reactors.

Experiments have been conducted in the DIII-D tokamak to explore the in-situ growth of silicon-rich layers as a potential technique for real-time replenishment of surface coatings on plasma-facing components (PFCs) during steady-state long-pulse reactor operation. Silicon (Si) pellets of 1 mm diameter were injected into low- and high-confinement (L-mode and H-mode) plasma discharges with densities ranging from 3.9– $7.5\times10^{19}$ m ^−3 and input powers ranging from 5.5 to 9 MW. The small Si pellets were delivered with the impurity granule injector at frequencies ranging from 4 to 16 Hz corresponding to mass flow rates of 5–19 mg s ^−1 (1– $4.2\times10^{20}$ Si s ^−1 ) at cumulative amounts of up to 34 mg of Si per five-second discharge. Graphite samples were exposed to the scrape-off layer and private flux region plasmas through the divertor material evaluation system to evaluate the Si deposition on the divertor targets. The Si II emission at the sample correlates with silicon injection and suggests net surface Si-deposition in measurable amounts. Post-mortem analysis showed Si-rich coatings containing silicon oxides, of which SiO _2 is the dominant component. No evidence of SiC was found, which is attributed to low divertor surface temperatures. The in-situ and ex-situ analysis found that Si-rich coatings of at least 0.4–1.2 nm thickness have been deposited at 0.4–0.7 nm s ^−1 . The technique is estimated to coat a surface area of at least 0.94 m ^2 on the outer divertor. These results demonstrate the potential of using real-time material injection to form Si-enriched layers on divertor PFCs during reactor operation.

Experiments performed in DIII-D demonstrate that higher plasma current and heating power combined with impurity seeding facilitate the achievement of divertor detachment with a higher pedestal pressure and higher plasma performance in H-mode plasmas with a baffled closed divertor compared with an open divertor. Dedicated experiments were carried out to study the impact of power, plasma current and impurity seeding on divertor detachment with ion $B \times \nabla B$ directed into the divertor favorable for the L–H transition. With a factor of three variation in heating power and with only D _2 puffing, no significant difference in the separatrix density at detachment onset was found. The higher heating power leads to higher impurity concentration and wider scrape-off layer (SOL) width, and reduces the detachment onset density to one similar to that in lower-power plasmas. Higher current requires higher pedestal and line-averaged densities to achieve divertor detachment; however, the increase in separatrix density at increasing plasma current is found to be less pronounced. Initial calculations found that both power scan and plasma current scan datasets are qualitatively consistent with theory after considering the change in impurity concentration and heat flux width. This also motivates the future extensive study of transport and divertor impurity behavior in order to have a quantitative comparison between experiment and theory. Compared with an open divertor, a closed divertor facilitates detachment onset at ∼40% lower line-averaged plasma density. Additional N _2 seeding facilitates the achievement of detachment at a lower separatrix density and thus a higher pedestal temperature, which is beneficial for advanced tokamak scenarios. Higher heating power requires a higher N _2 puffing rate to achieve the same degree of detachment, while a higher N _2 puffing rate leads to lower detachment onset line-averaged density, both of which agree with theory. In contrast to the narrower pedestal in an open divertor approaching detachment, the pedestal density width in a closed divertor increases with density. The density gradient increases with line-averaged density at higher plasma current, but remains nearly unchanged at lower plasma current. In particular, compared with discharges with low power, at high heating power the pedestal density gradient is much weaker, while the SOL density is significantly higher and wider. At the same plasma current, both pedestal pressure gradient and temperature gradient decrease linearly with the line-averaged density but remain similar across different heating powers. Even with different plasma current and heating power, the normalized pressure gradient remains identical. As a result, achievement of divertor detachment with a higher pedestal pressure and higher plasma performance is shown in a closed divertor, which is important for improving core–edge integration as one of the critical issues for future tokamak fusion reactors.

A new, comprehensive set of evidence reveals that Micro-Tearing Modes (MTMs) dominate pedestal electron heat transport in an H-mode experiment in the DIII-D tokamak. The experiment investigates the role of MTMs by scanning pedestal collisionality, a main drive of MTM instability, from 0.43 to 0.84 on the pedestal top. Broadband (150–800 kHz) magnetic and density fluctuations originating from the pedestal gradient region and highly consistent with MTMs are observed, with amplitude increasing during the scan. The higher magnetic fluctuation amplitude correlates with a lower pedestal electron temperature gradient, implying MTMs may regulate the pedestal electron heat transport. The collisionality scan results in profile and transport changes consistent with predicted transport capability of MTMs: (1) experimentally-determined electron heat diffusivity increases ∼40% at the location where the broadband density fluctuations peak; (2) ion heat diffusivity has less increase (

We assess the toroidal magnetic field B _t asymmetry in DIII-D due to a misalignment of the toroidal field coils with respect to the poloidal magnetic field coils and vacuum vessel. The peak-to-peak variation of the divertor strike point (SP) radius is measured to be 1 cm, with an n = 1 toroidal pattern. We use the centre of a narrow carbon deposition band on tungsten-coated divertor tiles just inside the outer strike point (OSP) as a proxy for the divertor SP location. The band occurred in a series of reverse B _t discharges with the OSP positioned on the divertor inserts due to strong E × B drift transport of C from the inner to the outer SP through the private flux region. The variation in band radius (and hence the magnetic SP) is a (4.89 ± 0.31) mm shift toward (310 ± 4)° toroidal direction. These measurements agree well with previous measurements of the 3D magnetic field distribution (Luxon 2003 Nucl. Fusion 43 1813), simulations performed by the mafot field line integration code, and recent Langmuir probe measurements in the small-angle-slot (SAS) divertor (Watkins et al 2019 Nucl. Mater. Energy 18 46). Comparison of these measurements in the SAS divertor also indicates that there is the possibility of a tilt (in conjunction with the shift) of the B _t coil field of (0.04 ± 0.07)° towards the toroidal angle of (215 ± 25)°. Previous measurements suggested a field misalignment of (4.6 ± 0.3) mm in the 270° toroidal direction, and a tilt of (0.06 ± 0.02)° toward the 114° toroidal direction, which is similar to the results reported here. These studies will be important for better understanding the radial variation of the toroidal strike line in DIII-D, for designing the new generation of SAS divertor, and for developing an understanding of the impact of error fields on tokamaks with tightly baffled slot divertors.

The Experimental Advanced Superconducting Tokamak (EAST) has recently achieved a variety of H-mode regimes with different edge-localized mode (ELM) dynamics, including type-I ELMs, compound ELMs, which are manifested by the onset of a large spike followed by a sequence of small spikes on Da emissions, usual type- III ELMs, and very small ELMs. This newly observed very small ELMy H-mode appears to be similar to the type-II ELMy H-mode, with higher repetition frequency (~1 kHz) and lower amplitude than the type-III ELMy H-mode, exhibiting an intermediate confinement level between type-I and type-III ELMy H-modes. The energy loss and divertor power load are systematically characterized for these different ELMy H-modes to provide a physics basis for the next-step high-power long-pulse operations in EAST. Both type-I and compound ELMs exhibit good confinement (H98y,2) ~ 1). A significant loss of the plasma stored energy occurs at the onset of type-I ELMs (~8%) and compound ELMs (~5%), while no noticeable change in the plasma stored energy is observed for the small ELMs, including both type-III ELMs and very small ELMs. The peak heat flux on divertor targets for type-I ELMs currently achieved in EAST is about 10 MW m-2, as determined from the divertor-embedded triple Langmuir probe system with high time resolution. As expected, type-III ELMs lead to much smaller divertor power loads with a peak heat flux of about 2MWm-2. Peak power loads for compound ELMs are between those for type-I and type-III ELMs. It is remarkable that the new very small ELMy H-modes exhibit even lower target power deposition than type-III ELMs, with the peak heat flux generally below 1 MWm-2. These very small ELMs are usually accompanied by broadband fluctuations with frequencies ranging from 20 to 50 kHz, which may promote particle and power exhaust throughout the very small ELMy H-mode regime. [ABSTRACT FROM AUTHOR]

Edge plasma code SOLPS modeling has been used to assess the effects of divertor detachment vs attachment and closure vs openness on the main scrape-off layer (SOL) plasma in DIII-D. The differences between the inner divertor and outer divertor, and the effects of divertor closure on the outer-midplane SOL plasma, have been analysed. It is found that the divertor shape and plasma state have significant influence on the SOL and pedestal plasma. (i) The inner divertor has a more significant effect on the high-field side (HFS) SOL than the outer divertor does on the low-field side (LFS) SOL, and for increasing main plasma density the ionization region in the HFS moves gradually from the divertor target into the HFS SOL, forming a high-density front in the high density discharge case. Also, particle flux reversal occurs near the separatrix. (ii) It is confirmed that divertor closure has a significant effect on the main SOL and pedestal plasma. Comparing an open and a closed divertor, it is found that for the closed divertor the ne radial profile at the OMP is higher outside the separatrix and lower inside it, due to the combined effects of the D+ and C impurity ion density. Moreover, the open divertor has higher core fueling than the closed divertor, and the closed divertor has better fuel and impurity neutral screening than the open divertor. (iii) The in/out asymmetry is also influenced by the divertor closure, increase of the closure of the outer divertor can substantially reduce the asymmetry. The closed divertor also has much higher HFS SOL density than that of the LFS SOL. These results highlight the strong dependence of neutral/impurity screening and main SOL properties on divertor geometry and plasma regimes, which needs to be taken into account in advanced divertor design for fusion reactors. [ABSTRACT FROM AUTHOR]

Improvements in the H-mode pedestal pressure and global energy confinement were observed on DIII-D with both 3 and 6 MW of neutral beam (NBI) heating with neon injection. The pedestal pressure increased, primarily associated with a significant increase in the pedestal density with a relatively smaller decrease in the pedestal temperature. At the same neon injection rate, a rapid decrease in edge localized mode (ELM) frequency and ∼35% increase in plasma stored energy were observed in the 3 MW discharge, while in the 6 MW discharge, the ELM frequency was gradually decreased and the plasma stored energy gradually increased by up to 20%. The measured pedestal widths from both 3 and 6 MW discharges matched the EPED1.0 width scaling, , within 20% deviations before and after neon injection. The peeling-ballooning mode (PBM) stability analysis showed that after neon injection, the ballooning stability boundary was increased, while the peeling stability boundary only dropped once the neon level was relatively high. The ballooning stability boundary increase with neon injection is consistent with increase of the ion diamagnetic stabilization (i.e. including the contribution from carbon, neon as well as deuterium ions) which is used in the criterion to determine when the calculated linear growth rates of PBM are unstable. Consequently, the new ELITE computed PBM boundaries were more consistent with the operating points than when only deuterium was included. This in turn indicates that the impurity may help to improve pedestal pressure through affecting ion diamagnetic frequency in the pedestal region, which could positively lead to the ballooning stability boundary extension. [ABSTRACT FROM AUTHOR]

UEDGE simulations highlight the role of cross-field drifts on the onset of detached conditions, and new calibrated divertor vacuum ultra violet (VUV) spectroscopy is used to challenge the predictions of radiative constituents in these simulations. UEDGE simulations for DIII-D H-mode plasmas with the open divertor with the ion ∇B-drift towards the X-point show a bifurcated onset of the low field side (LFS) divertor detachment, consistent with experimentally observed step-like detachment onset (Jaervinen A.E. et al 2018 Phys. Rev. Lett.121 075001). The divertor plasma in the simulations exhibits hysteresis in upstream separatrix density between attached and detached solution branches. Reducing the drift magnitude by a factor of 3 eliminates the step-like detachment onset in the simulations, confirming the strong role of drifts in the bifurcated detachment onset. When measured local plasma densities and temperatures are within proximity of predicted values in the simulations, there is no shortfall of the local emission of the dominant resonant radiating lines. However, the simulations systematically predict a factor of two lower total integrated radiated power than measured by the bolometer with the difference lost through radial heat flow out of the computational domain. Even though there is no shortfall in the emission of the dominant lines, a shortfall of total radiated power can be caused by underpredicted spatial extent of the radiation front, indicating a potential upstream or divertor transport physics origin for the radiation shortfall, or shortfall of radiated power in the spectrum between the dominant lines. In addition to the underpredicted spatial extent, in detached conditions, the simulations overpredict the peak radiation and dominant carbon lines near the X-point, which can be alleviated by manually increasing divertor diffusivity in the simulations, highlighting the ad hoc cross-field transport as one of the key limitations of the predictive capability of these divertor fluid codes. [ABSTRACT FROM AUTHOR]

A toroidally symmetric tungsten ring inserted in the lower outer divertor of DIII-D was exposed to 25 repeated, attached L-mode shots in reverse- configuration. Radial profiles of the W gross erosion flux inferred in situ from spectroscopic measurements of the WI line (400.9 nm) during these experiments are well reproduced by ERO-D3D simulations of carbon and tungsten impurity erosion, transport and redeposition in the outer divertor region. Tungsten gross erosion is mainly induced by physical sputtering of tungsten by carbon impurities. The outward radial transport of carbon impurities in the outer divertor is shown to be mainly governed by drifts in the sheath region. In addition, the erosion and redeposition of carbon on tungsten, induced by the implantation of carbon into tungsten modeled with the homogeneous mixed material model, increases the effective flux of carbon impurities onto the tungsten ring (carbon recycling on tungsten). The dynamics of carbon implantation in tungsten is shown to be consistent with the plasma shot duration in DIII-D. Moreover, it is shown that the localized deposition of tungsten measured experimentally in the outboard region away from the tungsten ring is caused by the long-range radial transport of tungsten impurities in the outer divertor region induced by the interplay between poloidal and radial drifts. Such experimental measurements might provide direct quantitative estimations of tungsten net erosion. The modeling and analysis of carbon and tungsten erosion and redeposition presented in this paper demonstrates that various physical mechanisms and their synergistic effects need to be taken into account to accurately describe erosion, transport and redeposition of impurities in tokamak divertors. [ABSTRACT FROM AUTHOR]

The relationship between the tungsten gross erosion and the edge-localized mode (ELM) characteristics is investigated for high performance discharges during the dedicated metal rings campaign in DIII-D, using the 1D ‘free-streaming’ ELM model, coupled to a simple analytical TRIM.SP sputtering model, while the OEDGE code is employed to calculate the inter-ELM impurity distribution and tungsten gross erosion. A carbon–tungsten mixed target material is used for both inter- and intra-ELM simulations. The OEDGE results reveal that the inter-ELM tungsten gross erosion rate is not sensitive to the carbon fraction on the tungsten surface layer; a wider range of carbon fractions from 0.3 to 0.6 gives similar tungsten gross erosion profiles and matches the diagnostic data, which is consistent with the simple analytical model. However, the intra-ELM simulation indicates that the value of the surface carbon fraction can be as low as 0.3 to reproduce the measured tungsten gross erosion during ELMs, based on the effective ionizations/photon (S/XB) value of WI emissions. For both inter- and intra-ELM cases, carbon is predicted to dominate the tungsten erosion. However, in contrast to the inter-ELM tungsten erosion, which is dominated by locally redeposited C2+, the energetic C6+ originating from the pedestal can contribute substantial tungsten erosion in the near-separatrix region during ELMs, and the width of this C6+ dominated region exhibits a negative correlation with the heat flux density carried by ELMs. [ABSTRACT FROM AUTHOR]

The detached divertor has recently been achieved in H-mode plasmas in EAST with an ITER-like tungsten divertor, which is essential for high power long pulse operation in the near future. The divertor detachment is indicated by the rollover of particle flux and the drop in electron temperate down to ~5 eV, as measured by the Langmuir probes at the divertor target, as the plasma density increases. The density threshold of detachment increases with heating power, and is higher with combined neutral beam injection (NBI) and radiofrequency heating than NBI only. The divertor detachment exhibits a strong in–out asymmetry for favorable BT, i.e. with the ion B  ×   drift direction toward the X-point. In contrast, with reverse toroidal field, the particle flux near the outer strike point first increases with density, then maintains nearly constant, without pronounced roll-over, when the divertor plasma enters the detachment regime, and the onset of detachment between inner and outer divertors becomes nearly balanced, as expected from the effect of drifts. [ABSTRACT FROM AUTHOR]

A new small angle slot (SAS) divertor concept has been developed to enhance neutral cooling across the divertor target by coupling a closed slot structure with appropriate target shaping. Initial tests on DIII-D find a strong interplay between such anticipated ‘SAS’ effects and cross-field drifts, favouring operation with the ion B  ×  ∇B drift away from the X-point, as currently employed for advanced tokamaks. This offers the following key improvements relative to DIII-D’s open lower divertor or partially-closed upper divertor: (i) SAS allows for transition to low temperature moderately detached divertor conditions with Te  ≲  10 eV at very low main plasma densities, lower than are usually attainable at all in DIII-D high confinement (H-mode) plasmas as used in these tests; (ii) Pedestal performance and core confinement are significantly improved with SAS. The final confinement collapse associated with the onset of X-point MARFE (multifaceted asymmetric radiation from the edge) following deep detachment occurs at significantly higher pedestal densities, thus widening the window of H-mode operation compatible with a dissipative divertor. For operation with the ion B  ×  ∇B drift toward the X-point, the divertor plasma transitions to a bifurcative detached state at much higher densities, similar to other divertor configurations in DIII-D. These results highlight the strong interplay between divertor closure and drifts, and point to an interesting divertor optimization path to explore that offers potential for future fusion reactors. [ABSTRACT FROM AUTHOR]

The ‘Super H-Mode’ regime is predicted to enable pedestal height and fusion performance substantially higher than standard H-Mode operation. This regime exists due to a bifurcation of the pedestal pressure, as a function of density, that is predicted by the EPED model to occur in strongly shaped plasmas above a critical pedestal density. Experiments on Alcator C-Mod and DIII-D have achieved access to the Super H-Mode (and Near Super H) regime, and obtained very high pedestal pressure, including the highest achieved on a tokamak (p ped ~ 80 kPa) in C-Mod experiments operating near the ITER magnetic field. DIII-D Super H experiments have demonstrated strong performance, including the highest stored energy in the present configuration of DIII-D (W ~ 2.2–3.2 MJ), while utilizing only about half of the available heating power (Pheat ~ 7–12 MW). These DIII-D experiments have obtained the highest value of peak fusion gain, QDT,equiv ~ 0.5, achieved on a medium scale (R  <  2 m) tokamak. Sustained high performance operation (βN ~ 2.9, H98 ~ 1.6) has been achieved utilizing n  =  3 magnetic perturbations for density and impurity control. Pedestal and global confinement has been maintained in the presence of deuterium and nitrogen gas puffing, which enables a more radiative divertor condition. A pair of simple performance metrics is developed to assess and compare regimes. Super H-Mode access is predicted for ITER and expected, based on both theoretical prediction and observed normalized performance, to allow ITER to achieve its goals (Q  =  10) at Ip  <  15 MA, and to potentially enable more compact, cost effective pilot plant and reactor designs. [ABSTRACT FROM AUTHOR]

A total power injection up to 0.3 GJ has been achieved in EAST long pulse H-mode operation of 101.2 s with an ITER-like water-cooled tungsten (W) mono-block divertor, which has steady-state power exhaust capability of 10 MWm−2. The peak temperature of W target saturated at 12 s to the value T ~ 500 °C with a heat flux ~3.3 MW m−2 being maintained during the discharge. By tailoring the 3D divertor plasma footprint through edge magnetic topology change, the heat load was broadly dispersed and thus peak heat flux and W sputtering were well controlled. Active feedback control of H-mode detachment with D2 fuelling or divertor impurity seeding has been achieved successfully, with excellent compatibility with the core plasma performance. Active feedback control of radiative power utilizing neon seeding was achieved with f rad  =  18%–41% in H-mode operation, exhibiting potential for heat flux reduction with divertor and edge radiation. This has been further demonstrated in DIII-D high βP H-mode scenario within the joint DIII-D/EAST experiment using impurity seeding from the divertor volume. Steady-state particle control and impurity exhaust has been achieved for long pulse H-mode operation over 100 s with the W divertor by leveraging the effect of drifts and optimized divertor configuration, coupled with strong pumping and extensive wall conditioning. Approaches toward the reduction of divertor W sourcing, which is of crucial importance for a metal-wall tokamak, are also explored. These advances provide important experimental information on favourable core-edge integration for high power, long-pulse H-mode operation in EAST, ITER and CFETR. [ABSTRACT FROM AUTHOR]

Wendelstein 7-X (W7-X) was operated successfully with the first divertor plasma in the operation phase 1.2a (OP1.2a). A new combined probe head, developed and installed on the multiple-purpose manipulator, is able to measure the edge plasma profiles (), variation of magnetic field, poloidal and radial turbulence structures. The scrape-off layer (SOL) plasma parameters in two magnetic configurations (standard and high mirror) are in good agreement with the magnetic island structure and the field line connection length calculated by the field line tracer. In both the standard and high mirror configurations, the radial turbulent heat flux and particle flux have strong dependence on the local magnetic topology, revealing two distinct transport patterns: a broadband turbulence dominant region in the outer SOL and a low frequency dominant region in the inner SOL. In the standard divertor configuration, the broadband turbulence with a frequency range of 40–120 kHz is located near the island center along the probe path, leading to outward transport. These broadband fluctuations propagate with a velocity of 2.3–4.4 km s−1 poloidally along the ion diamagnetic drift direction in the plasma frame, with close to 0.1. The large radial transport induced by the broadband turbulence is accompanied by a steep electron density gradient. The low frequency (5–30 kHz) dominant transport exhibits obvious intermittent structure. Some statistical techniques are applied to the characterization of the intermittent transport. [ABSTRACT FROM AUTHOR]

Resonant magnetic perturbations (n  =  3 RMPs) are used to suppress large amplitude ELMs and mitigate naturally occurring ‘grassy’-ELMs in DIII-D plasmas relevant to the ITER steady-state mission. Fully non-inductive discharges in the ITER shape and pedestal collisionality (  ≈  0.05–0.15) are routinely achieved in DIII-D with RMP suppression of the Type-I ELMs. The residual grassy-ELMs deliver a low peak heat flux to the divertor as low as 1.2×  the inter-ELM heat flux in plasmas with sustained high H-factor (H98y2  ≈  1.2). The operating window for the RMP grassy-ELM regime is q95  =  5.3–7.1 and external torque in the range 9–0.7 Nm in the co-Ip direction, which is in the range required for a steady-state tokamak reactor. The RMP grassy-ELM regime is associated with a two-step pedestal, with strong flattening of the density around the zero crossing in the E  ×  B shear. The edge magnetic response of the plasma to the n  =  3 RMP is found to be  ≈2–3×  larger than for comparable ITER baseline plasmas (βN  ≈  1.8, q95  ≈  3.1). The amplification of the RMP is consistent with the weak magnetic perturbation level (δB/B  ≈  1  ×  10−4) required for effective Type-I ELM suppression. Cyclic variations in the pedestal pressure, width, and toroidal rotation are observed in these plasmas, correlated with cyclic variations in the strength and frequency of the grassy-ELMs. Extended MHD analysis and magnetic measurements indicate that these pedestal pulsations are driven by cyclic variations in the resonant field strength at the top of the pedestal. These pedestal pulsations reveal that the grassy-ELMs is correlated with the proximity of the pedestal to the low-n peeling-ballooning mode stability boundary. The use of low amplitude magnetic fields to access grassy-ELM conditions free of Type-I ELMs in high beta poloidal plasmas (βP  ≈  1.5–2.0) opens the possibility for the further optimization of the steady-state tokamak by use of edge resonant magnetic perturbations. [ABSTRACT FROM AUTHOR]

Dedicated experiments have been performed in the DIII-D tokamak to assess the influence of divertor geometry on the H-mode pedestal structure. It has been found that in both attached and detached plasmas, compared to the open divertor, the more closed divertor results in lower density at the pedestal top. Higher pedestal performance can be achieved with the closed divertor in detached plasmas, as manifested by a higher pedestal temperature at the same pedestal density, by about a factor of two, than the open divertor. In addition, approaching divertor detachment by increasing the gas-puffing rate, for different divertor geometries, the pedestal width exhibits different trends. In the attached plasma, the pedestal width agrees well with the theoretical and empirical pedestal-poloidal-beta scaling. However, during divertor detachment, in the open divertor the pedestal width is significantly reduced. In contrast, for detached plasmas with the more closed divertor, the pedestal is significantly wider, by up to 50%, compared to the theoretical scaling. Moreover, near divertor detachment, the open diverted plasma exhibits a more aligned density and temperature pedestal profile, while in the closed divertor the detachment results in a relative shift (up to 50% of the pedestal width) between the density and temperature pedestal profiles. [ABSTRACT FROM AUTHOR]

A stationary edge-localized mode (ELM)-absent H-mode regime, with an electrostatic edge coherent mode (ECM) which resides in the pedestal region, has been achieved in the EAST tokamak recently. This regime allows the operation of a nearly fully noninductive long pulse (>15 s), exhibiting a relatively high pedestal and good global energy confinement with near 1.2, and excellent impurity control. Furthermore, this regime is mostly obtained with a 4.6 GHz lower hybrid current drive (LHCD) or counter-current neutral beam injection (NBI), plus electron cyclotron resonance heating, and an extensive lithium wall coating. This stationary ELM-absent H-mode regime transits to a stationary small ELM H-mode regime, and upon additional heating power from the 2.45 GHz LHCD, an ion cyclotron resonant frequency or co-current NBI is applied (under 4.6 GHz LHCD heating background). A slight change of the plasma configuration also makes the small ELMs reappear. The experimental observations suggest that a long-pulse ELM-absent regime can be induced by the ECM, which exhibits strong electrostatic fluctuations and may provide a channel for continuous particle (especially impurities) and heat exhaust across the pedestal. The ECM exists in the collisionality of   =  2.5–4 and the pressure gradient   =    =  100–200 (kPa), which is in good agreement with the previous simulation of GYRO. This ELM-absent H-mode regime with ECM may offer a suitable candidate for high-performance, steady-state H-mode operation in future fusion reactors. [ABSTRACT FROM AUTHOR]

A new model for the edge harmonic oscillations (EHOs) in the quiescent H-mode regime has been developed, which successfully reproduces the recent observations in the DIII-D tokamak. In particular, at high E  ×  B flow shear only a few low-n kink modes remain unstable at the plasma edge, consistent with the EHO behavior, while at low E  ×  B flow shear, the unstable mode spectrum is significantly broadened, consistent with the low-n broadband electromagnetic turbulence behavior. The model is based on a new mechanism for destabilizing low-n kink/peeling modes by the E  ×  B flow shear, which underlies the EHOs, separately from the previously found Kelvin–Helmholtz drive. We find that the differential advection of mode vorticity by sheared E  ×  B flows modifies the 2D pattern of mode electrostatic potential perpendicular to the magnetic field lines, which in turn causes a radial expansion of the mode structure, an increase of field line bending away from the mode rational surface, and a reduction of inertial stabilization. This enhances the kink drive as the parallel wavenumber increases significantly away from the rational surface at the plasma edge where the magnetic shear is also strong. This destabilization is also shown to be independent of the sign of the flow shear, as observed experimentally, and has not been taken into account in previous pedestal linear stability analyses. Verification of the veracity of this EHO mechanism will require analysis of the nonlinear evolution of low-n kink/peeling modes so destabilized in the linear regime. [ABSTRACT FROM AUTHOR]

The effects of changes in core density on divertor electron temperature, density and heat flux when resonant magnetic perturbations (RMPs) are applied are presented, notably a reduction in RMP induced secondary radial peaks in the electron temperature profile at the target plate is observed when the core density is increased, which is consistent with modeling. RMPs is used here to indicate non-axisymmetric magnetic field perturbations, created using in-vessel control coils, which have at least one but typically many resonances with the rotational transform of the plasma (Evans et al 2006 Phys. Plasmas13 056121). RMPs are found to alter inter-ELM heat flux to the divertor by modifying the core plasma density. It is shown that applying RMPs reduces the core density and increases the inter-ELM heat flux to both the inner and outer targets. Using gas puffing to return the core density to the pre-RMP levels more than eliminates the increase in inter-ELM heat flux, but a broadening of the heat flux to the outer target remains. These measurements were made at a single toroidal location, but the peak in the heat flux profile was found near the outer strike point where simulations indicate little toroidal variation should exist and tangentially viewing diagnostics showed no evidence of strong asymmetries. In experiments where divertor Thomson scattering measurements were available it is shown that local secondary peaks in the divertor electron temperature profile near the target plate are reduced as the core density is increased, while peaks in the divertor electron density profile near the target are increased. These trends observed in the divertor electron temperature and density are qualitatively reproduced by scanning the upstream density in EMC3-Eirene modeling. Measurements are presented showing that higher densities are needed to induce detachment of the outer strike point in a case where an increase in electron temperature, likely due to a change in MHD activity, is seen after RMPs are applied. [ABSTRACT FROM AUTHOR]

A recent experiment on DIII-D, which was conducted by the joint research team from DIII-D and EAST, has extended the previous high , high qmin regime, which has been tested in the 2013 DIII-D/EAST joint experiment, to inductive operation at higher plasma current ( MA) and significantly higher normalized fusion performance (). The experiment aims at exploring high performance scenario with and reduced torque for long pulse operation, which can be potentially extrapolated to EAST. The effort was largely motivated by the interest in developing a feasible scenario for long-pulse high performance operation with low torque on EAST. Very high confinement, H89  =  3.5 or with , has been achieved transiently in this experiment together with and reduced NBI torque ( N m). The excellent confinement is associated with the spontaneous formation of an internal transport barrier (ITB) in plasmas with MA at large minor radius (normalized ) in all channels (ne, Te, Ti, , especially strong in the Te channel). Fluctuation measurements show a significant reduction in the fluctuation levels, including AE modes and broadband turbulence, at the location where an ITB forms. Linear gyrokinetic simulations also support the decrease of the growth rate of the most unstable mode during strong ITB formation. The simulation implies that strong suppression of turbulence and a positive feedback loop may be active in this process and is responsible for the spontaneous formation of large-radius ITB. In an unstable ITB phase, an ELM crash is observed to have a positive effect on transient formation of large-radius ITB. The formation of this kind of ITB is found to have a shielding (protecting) effect on the core plasma while isolating the perturbation due to ELM crash. [ABSTRACT FROM AUTHOR]

The in–out divertor asymmetry in the Experimental Advanced Superconducting Tokamak (EAST), as manifested by particle fluxes measured by the divertor triple Langmuir probe arrays, is significantly enhanced during type-I edge localized modes (ELMs), favoring the inner divertor in lower single null (LSN) for the normal toroidal field (Bt) direction, i.e. with the ion B  ×   B direction towards the lower X-point, while the in–out asymmetry is reversed when the ion B  ×   B is directed away from the lower X-point. The plasma flow measured by the Mach probe at the outer midplane is in the ion Pfirsch–Schlüter (PS) flow direction, opposite to both B  ×   B and E  ×  B drifts, i.e. towards the inner divertor for normal Bt, and the outer divertor for reverse Bt, consistent with the observed in–out divertor asymmetry in particle fluxes. Since the particle source from an ELM event is predominantly located near the outer midplane, this new finding suggests a critical role of the PS flow in driving the in–out divertor asymmetry. The divertor asymmetry during type-III ELMs exhibits a similar trend to that during type-I ELMs. Strong in–out divertor asymmetry is also present during inter-ELM and ELM-free phases for the normal field direction, i.e. with more particle flux to the lower inner divertor target, but the peak particle flux merely becomes more symmetric, or slightly reversed, for reverse Bt, i.e. reversed B  ×   B drift direction. [ABSTRACT FROM AUTHOR]

The power threshold for low (L) to high (H) confinement mode transition achieved by radio-frequency (RF) heating and lithium-wall coating is investigated experimentally on EAST for two sets of walls: an all carbon wall (C) and molybdenum chamber and a carbon divertor (Mo/C). For both sets of walls, a minimum power threshold Pthr of ~0.6 MW was found when the EAST operates in a double null (DN) divertor configuration with intensive lithium-wall coating. When operating in upper single null (USN) or lower single null (LSN), the power threshold depends on the ion  ∇B drift direction. The low density dependence of the L–H power threshold, namely an increase below a minimum density, was identified in the Mo/C wall for the first time. For the C wall only the single-step L–H transition with limited injection power is observed whereas also the so-called dithering L–H transition is observed in the Mo/C wall. The dithering behaves distinctively in a USN, DN and LSN configuration, suggesting the divertor pumping capability is an important ingredient in this transition since the internal cryopump is located underneath the lower divertor. Depending on the chosen divertor configuration, the power across the separatrix Ploss increases with neutral density near the lower X-point in EAST with the Mo/C wall, consistent with previous results in the C wall (Xu et al 2011 Nucl. Fusion51 072001). These findings suggest that the edge neutral density, the ion  ∇B drift as well as the divertor pumping capability play important roles in the L–H power threshold and transition behaviour. [ABSTRACT FROM AUTHOR]

Dedicated experiments focusing on the influence of lower hybrid waves (LHWs) on edge-localized modes (ELMs) were first performed during the 2012 experimental campaign of EAST, via modulating the input power of LHWs in the high-confinement-mode (H-mode) plasma mainly sustained by ion cyclotron resonant heating. Natural ELMs are effectively mitigated (ELM frequency increases, while its intensity decreases dramatically) as the LHW is applied, observed over a fairly wide range of plasma current or edge safety factor. By scanning the modulation frequency (fm) of LHW injected power in a target plasma dominated by the so-called small ELMs, we conclude that large ELMs with markedly larger amplitudes and lower frequencies are reproduced at low modulation frequencies (fm < 100 Hz). Analysis of the evolution of edge extreme ultraviolet radiation signals further indicates that plasma fluctuations at the pedestal region indistinctively respond to rapid modulation (fm ⩾ 100 Hz) of LHW injected power. This is proposed as the mechanism responsible for the observed fm dependence of the mitigation effect induced by LHWs on large ELMs. In addition, a critical threshold of LHW input power PLHW is estimated as , beyond which the impact of applied LHWs on ELM behaviours can be achieved. Finally, Langmuir probe measurements suggest that, rather than the concentration of free energy into a narrowband quasi-coherent precursor commonly observed growing until the ELM crash, the continuous development of broadband turbulence during the ELM-absent phase with the application of LHWs might contribute to the avoidance of ELM crashes. These results present new insights into existing experiments, and also provide some foundations and references for the next-step research about exploring in more depth and improving this new attractive method to effectively control the ELM-induced very large transient heat and particle flux. [ABSTRACT FROM AUTHOR]

The ITER beryllium (Be) first wall (FW) panels are shaped to protect leading edges between neighbouring panels arising from assembly tolerances. This departure from a perfectly cylindrical surface automatically leads to magnetically shadowed regions where eroded Be can be re-deposited, together with co-deposition of tritium fuel. To provide a benchmark for a series of erosion/re-deposition simulation studies performed for the ITER FW panels, dedicated experiments have been performed on the EAST tokamak using a specially designed, instrumented test limiter acting as a proxy for the FW panel geometry. Carbon coated molybdenum plates forming the limiter front surface were exposed to the outer midplane boundary plasma of helium discharges using the new Material and Plasma Evaluation System (MAPES). Net erosion and deposition patterns are estimated using ion beam analysis to measure the carbon layer thickness variation across the surface after exposure. The highest erosion of about 0.8 µm is found near the midplane, where the surface is closest to the plasma separatrix. No net deposition above the measurement detection limit was found on the proxy wall element, even in shadowed regions. The measured 2D surface erosion distribution has been modelled with the 3D Monte Carlo code ERO, using the local plasma parameter measurements together with a diffusive transport assumption. Excellent agreement between the experimentally observed net erosion and the modelled erosion profile has been obtained. [ABSTRACT FROM AUTHOR]

Two enhanced-recycling H-mode regimes, named low-enhanced-recycling (LER) and high-enhanced-recycling (HER) H-mode regimes, with edge coherent modes, have been achieved by lower hybrid current drive and ion cyclotron resonance heating with lithium-wall conditioning in the EAST superconducting tokamak. In the LER H-mode regime, the density and radiation increase during the ELM-free phase until the onset of edge-localized modes (ELMs), while in the HER H-mode regime, the density and radiation are well controlled without the presence of ELMs. Both LER and HER H-modes exhibit a low-frequency (frequency <100 kHz) edge quasi-coherent mode (ECM) with an initial frequency chirping down phase, following the L–H transition. In addition, an electromagnetic high-frequency coherent mode (HFM) with frequency >170 kHz appears shortly (<1 ms) after the transition during HER H-modes. Both ECM and HFM propagate in the electron diamagnetic drift direction in the lab frame with a low poloidal wavelength and may be responsible for enhanced recycling during the ELM-free phase. These two enhanced-recycling H-mode regimes may have significant implications for long-pulse high-performance operations in future fusion experiments. [ABSTRACT FROM AUTHOR]

Dedicated experiments for the scaling of divertor power footprint width have been performed in the ITER-relevant radio-frequency (RF)-heated H-mode scheme under the lower single null, double null and upper single null divertor configurations in the Experimental Advanced Superconducting Tokamak (EAST) under lithium wall coating conditioning. A strong inverse scaling of the edge localized mode (ELM)-averaged power fall-off width with the plasma current (equivalently the poloidal field) has been demonstrated for the attached type-III ELMy H-mode as by various heat flux diagnostics including the divertor Langmuir probes (LPs), infra-red (IR) thermograph and reciprocating LPs on the low-field side. The IR camera and divertor LP measurements show that , in good agreement with the multi-machine scaling trend during the inter-ELM phase between type-I ELMs or ELM-free enhanced Dα (EDA). H-mode. However, the magnitude is nearly doubled, which may be attributed to the different operation scenarios or heating schemes in EAST, i.e., dominated by electron heating. It is also shown that the type-III ELMs only broaden the power fall-off width slightly, and the ELM-averaged width is representative for the inter-ELM period. Furthermore, the inverse Ip (Bp) scaling appears to be independent of the divertor configurations in EAST. The divertor power footprint integral width, fall-off width and dissipation width derived from EAST IR camera measurements follow the relation, λint ≅ λq + 1.64S, yielding . Detailed analysis of these three characteristic widths was carried out to shed more light on their extrapolation to ITER. [ABSTRACT FROM AUTHOR]

The turbulence and flows at the plasma edge during the L–I–H, L–I–L and single-step L–H transitions have been measured directly using two reciprocating Langmuir probe systems at the outer midplane with several newly designed probe arrays in the EAST superconducting tokamak. The E × B velocity, turbulence level and turbulent Reynolds stress at ∼1 cm inside the separatrix ramp-up in the last ∼20 ms preceding the single-step L–H transition, but remain nearly constant near the separatrix, indicating an increase in the radial gradients at the plasma edge. The kinetic energy transfer rate from the edge turbulence to the E × B flows is significantly enhanced only in the last ∼10 ms and peaks just prior to the L–H transition. The E × B velocity measured inside the separatrix, which is typically in the electron diamagnetic drift direction in the L-mode, decays towards the ion diamagnetic drift direction in response to fluctuation suppression at the onset of the single-step L–H, L–I–L as well as L–I–H transitions. One important distinction between the L–I–H and the L–I–L transitions has been observed, with respect to the evolution of the edge pressure gradient and mean E × B flow during the I-phase. Both of them ramp up gradually during the L–I–H transition, but change little during the L–I–L transition, which may indicate that a gradual buildup of the edge pedestal and mean E × B flow during the I-phase leads to the final transition into the H-mode. In addition, the transition data in EAST strongly suggest that the divertor pumping capability is an important ingredient in determining the transition behaviour and power threshold. [ABSTRACT FROM AUTHOR]