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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]

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]

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]

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]

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]