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Radiative divertor experiments and 2D fluid simulations show strong impact of cross-field drifts on the low field side (LFS) divertor target heat flux and volumetric radiation profiles when evolving to detached conditions in DIII-D high confinement mode (H-mode) plasmas with forward and reversed toroidal field configurations. In both field configurations, the peak heat flux is reduced by about a factor of 2 in detachment by D 2 -injection and by factor of 3 – 4 in detachment by N 2 -injection. Operating with the B × ∇B-drift towards the X-point (fwd. BT), the LFS divertor radiation front is observed to shift step-like from the target to near to the X-point at the onset of detachment. In contrast, operating with the B × ∇B-drift away from the X-point (rev. BT), the radiation front is observed to remain closer to the target plate and to be radially shifted towards the far SOL. These phenomena occur with detachment induced both with N 2 - and D 2 -injection. The step-like detachment onset is consistent with recently published theory of the role of poloidal E × B-drift in the private flux region in driving highly non-linear detachment onset in fwd. BT [22].

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.

The ITER Electron Cyclotron Heating (ECH) system provides 20 MW of microwave power from 24 gyrotron sources. The power is transmitted through evacuated, corrugated waveguide transmission lines. The...

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 T e ≲ 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.

DIII-D research is addressing critical challenges in preparation for ITER and the next generation of fusion devices through focusing on plasma physics fundamentals that underpin key fusion goals, understanding the interaction of disparate core and boundary plasma physics, and developing integrated scenarios for achieving high performance fusion regimes. Fundamental investigations into fusion energy science find that anomalous dissipation of runaway electrons (RE) that arise following a disruption is likely due to interactions with RE-driven kinetic instabilities, some of which have been directly observed, opening a new avenue for RE energy dissipation using naturally excited waves. Dimensionless parameter scaling of intrinsic rotation and gyrokinetic simulations give a predicted ITER rotation profile with significant turbulence stabilization. Coherence imaging spectroscopy confirms near sonic flow throughout the divertor towards the target, which may account for the convection-dominated parallel heat flux. Core-boundary integration studies show that the small angle slot divertor achieves detachment at lower density and extends plasma cooling across the divertor target plate, which is essential for controlling heat flux and erosion. The Super H-mode regime has been extended to high plasma current (2.0 MA) and density to achieve very high pedestal pressures (~30 kPa) and stored energy (3.2 MJ) with H 98y2 ≈ 1.6–2.4. In scenario work, the ITER baseline Q = 10 scenario with zero injected torque is found to have a fusion gain metric independent of current between q 95 = 2.8–3.7, and a lower limit of pedestal rotation for RMP ELM suppression has been found. In the wide pedestal QH-mode regime that exhibits improved performance and no ELMs, the start-up counter torque has been eliminated so that the entire discharge uses ≈0 injected torque and the operating space is more ITER-relevant. Finally, the high- (⩽3.8) hybrid scenario has been extended to the high-density levels necessary for radiating divertor operation, achieving ~40% divertor heat flux reduction using either argon or neon with P tot up to 15 MW.

A bifurcative step transition from low-density, high-temperature, attached divertor conditions to high-density, low-temperature, detached divertor conditions is experimentally observed in DIII-D tokamak plasmas as density is increased. The step transition is only observed in the high confinement mode and only when the $B\ifmmode\times\else\texttimes\fi{}\ensuremath{\nabla}B$ drift is directed towards the divertor. This work reports for the first time a theoretical explanation and numerical simulations that qualitatively reproduce this bifurcation and its dependence on the toroidal field direction. According to the model, the bifurcation is primarily driven by the interdependence of the $E\ifmmode\times\else\texttimes\fi{}B$-drift fluxes, divertor electric potential structure, and divertor conditions. In the attached conditions, strong potential gradients in the low field side (LFS) divertor drive $E\ifmmode\times\else\texttimes\fi{}B$-drift flux towards the high field side divertor, reinforcing low density, high temperature conditions in the LFS divertor leg. At the onset of detachment, reduction in the potential gradients in the LFS divertor leg reduce the $E\ifmmode\times\else\texttimes\fi{}B$-drift flux as well, such that the divertor plasma evolves nonlinearly to high density, strongly detached conditions. Experimental estimates of the $E\ifmmode\times\else\texttimes\fi{}B$-drift fluxes, based on divertor Thomson scattering measurements, and their dependence on the divertor conditions are qualitatively consistent with the numerical predictions. The implications for divertor power exhaust and detachment control in the next step fusion devices are discussed.

Numerical simulations of toroidal asymmetries in a tightly baffled small angle slot (SAS) divertor on the DIII-D tokamak show that toroidal asymmetries in divertor closure result in (non-axisymmetric) local onset of detachment within a density window of 10–15% on top of the nominal threshold separatrix density. The SAS divertor is explored at DIII-D for improving access to cold, dissipative/detached divertor conditions. The narrow width of the slot divertor coupled with a small magnetic field line-to-target angle facilitates the buildup of neutral density, thereby increasing radiative and neutrals-related (atoms and molecules) losses in the divertor. Small changes in the strike point location can be expected to have a large impact on divertor conditions. The combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field configuration causes the strike point to move along the divertor target plate, possibly leaving the divertor slot at some locations. The latter extreme case essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade the performance of the slot divertor. Such a strike point dislocation is approximated by a finite gap in the divertor baffle for which 3D edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.

Recent DIII-D experiments assessed the snowflake divertor (SF) configuration in a radiative regime in H-mode discharges with D 2 D 2 seeding. The SF configuration was maintained for many energy confinement times (2–3 s) in H-mode discharges ( I p =1.2 I p = 1.2 MA, P NBI =4–5 P NBI = 4 – 5 MW, and B×∇B B × ∇ B down (favorable direction toward the divertor)), and found to be compatible with high performance operation (H98y2 ⩾ 1). The two studied SF configurations, the SF-plus and the SF-minus, have a small finite distance between the primary X-point and the secondary B p B p null located in the private flux region or the common flux region, respectively. In H-mode discharges with the SF configurations (cf. H-mode discharges with the standard divertor with similar conditions) the stored energy lost per the edge localized mode (ELM) was reduced, and significant divertor heat flux reduction between and during ELMs was observed over a range of collisionalities, from lower density conditions toward a higher density H-modes with the radiative SF divertor.

Upgrades to core and divertor Thomson scattering (DTS) diagnostics at DIII-D have provided measurements of electron pressure profiles in the lower divertor from attached- to fully-detached divertor plasma conditions. Detailed, multistep sequences of discharges with increasing line-averaged density were run at several levels of Pinj. Strike point sweeping allowed 2D divertor characterization using DTS optimized to measure Te down to 0.5 eV. The ionization front at the onset of detachment is found to move upwards in a controlled manner consistent with the indication that scrape-off layer parallel power flux is converted from conducted to convective heat transport. Measurements of ne, Te and pe in the divertor versus Lparallel demonstrate a rapid transition from Te ⩾ 15 eV to ⩽3 eV occurring both at the outer strike point and upstream of the X-point. These observations provide a strong benchmark for ongoing modeling of divertor detachment for existing and future tokamak devices.

Simulations using the multi-fluid code UEDGE indicates that, in low confinement (L-mode) plasmas in DIII-D, poloidal projection of the ionization driven flows dominate poloidal particle flows in the divertor near the divertor plates, whereas E × B drift flows dominate the radial particle flows. In contrast, in high confinement (H-mode) conditions E × B drift flows dominate both poloidal and radial particle flows in the divertor in the vicinity of the strong gradient region near the separatrix. UEDGE indicates that the toroidal C2+ flow velocities in the divertor plasmas are mainly entrained within 30% to the background deuterium flow in both L- and H-mode plasmas in the plasma region where the CIII 465 nm emission is measured. Therefore, UEDGE indicates that the Doppler Coherence Imaging Spectroscopy (CIS), measuring the toroidal velocity of the C2+ ions, can provide insight to the deuterium flows in the divertor. Parallel-to-B velocity dominates the toroidal divertor flow; direct drift impact being less than 1%. Toroidal divertor flow is predicted to reverse when the magnetic field is reversed. This is explained by the parallel-B flow towards the nearest divertor plate corresponding to opposite toroidal directions in opposite toroidal field configurations.

An accelerated fusion energy development program, a fast-track approach, requires proceeding with a nuclear and materials testing program in parallel with research on burning plasmas, ITER. A Fusion Nuclear Science Facility (FNSF) would address many of the key issues that need to be addressed prior to DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues, and the direct relevance to an attractive target power plant. The standard aspect ratio provides space for a solenoid, assuring robust plasma current initiation,and for an inboard blanket, assuring robust tritium breeding ratio (TBR) >1 for FNSF tritium self-sufficiency and building of inventory needed to start up DEMO. An example design point gives a moderate sized Cu-coil device with R/a = 2.7 m/0.77 κ = 2.3, BT= 5.4 T, IP = 6.6 MA, βN = 2.75, Pfus = 127 MW. The modest bootstrap fraction of fBS = 0.55 provides an opportunity to develop steady state with sufficient current drive for adequate control. Lastly,more » proceeding with a FNSF in parallel with ITER provides a strong basis to begin construction of DEMO upon the achievement of Q ~ 10 in ITER.« less

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

New scalings of the dependence of divertor heat flux peak and profile width, important parameters for the design of future large tokamaks, have been obtained from recent DIII-D experiments. We find the peak heat flux depends linearly on input power, decreases linearly with increasing density, and increases linearly with plasma current. The profile width has a weak dependence on input power, is independent of density up to the onset of detachment, and is inversely proportional to the plasma current. We compare these results with previously published scalings, and present mathematical expressions incorporating these results.

Predictions of the recently developed paleoclassical transport model are compared with data from many toroidal plasma experiments: electron heat diffusivity in DIII-D, C-Mod and NSTX ohmic and near-ohmic plasmas; transport modeling of DIII-D ohmic-level discharges and of the RTP ECH 'stair-step' experiments with eITBs at low order rational surfaces; investigation of a strong eITB in JT-60U; H-mode Te edge pedestal properties in DIII-D; and electron heat diffusivities in non-tokamak experiments (NSTX/ST, MST/RFP, SSPX/spheromak). The radial electron heat transport predicted by the paleoclassical model is found to agree with a wide variety of ohmic-level experimental results and to set the lower limit (within a factor {approx} 2) for the radial electron heat transport in most resistive, current-carrying toroidal plasmas -- unless it is exceeded by fluctuation-induced transport, which often occurs in the edge of L-mode plasmas and when the electron temperature is high ({approx}>T{sub e}{sup crit} {approx}B{sup 2/3}{bar {alpha}}{sup 1/2} keV) because then paleoclassical transport becomes less than gyro-Bohm-level anomalous transport.

A new means of operating flux-core spheromaks with possibly increased stability, confinement and pulse length is analysed by a resistive magnetohydrodynamic (MHD) model. High amplification of the bias poloidal flux, required to minimize ohmic losses, is achieved by reducing the bias rapidly in a plasma formed at a lower amplification. The plasma separatrix is predicted to expand and incorporate the removed bias flux maintaining the total poloidal flux within the spheromak's flux-conserving wall. MHD energy on open magnetic field lines is reduced, reducing magnetic fluctuation levels. A means of experimental verification is suggested that may point the way to fusion-relevant spheromaks.

Magnetic reconnection in the spheromak changes magnetic topology by conversion of injected toroidal flux into poloidal flux and by magnetic surface closure (or opening) in a slowly decaying spheromak. Results from the Sustained Spheromak Physics Experiment, SSPX, are compared with resistive MHD simulations using the NIMROD code. Voltage spikes on the SSPX gun during spheromak formation are interpreted as reconnection across a negative-current layer close to the mean-field x-point. Field lines are chaotic during these events, resulting in rapid electron energy loss to the walls and the low T e < 50 eV seen in experiment and simulation during strong helicity injection. Closure of flux sufaces (and high T e ) can occur between voltage spikes if they are sufficiently far apart in time; these topology changes are not reflected in the impedance of the axisymmetric gun. Possible future experimental scenarios in SSPX are examined in the presence of the constraints imposed by reconnection physics.

We present significant advances in the performance and understanding of the sustained spheromak physics experiment (SSPX). High performance discharges are obtained by adjusting the injected current and bias magnetic flux to produce a flat to slightly peaked radial current profile, consistent with reducing the drive for tearing and other MHD modes. Wall conditioning and density control are also essential for obtaining high performance discharges. With these optimizations, plasmas with peak electron temperatures of ~350 eV, edge magnetic fluctuation levels less than 1% and core thermal diffusivities (χe)

The addition of inexpensive high-speed oscilloscopes has enabled higher Te Thomson scattering measurements on the SSPX spheromak. Along with signal correlation techniques, the scopes allow new analyses based on the shape of the scattered laser pulse to discriminate against fluctuating background plasma light that often make gated-integrator measurements unreliable. A 1.4 J Nd:YAG laser at 1064 nm is the scattering source. Spatial locations are coupled by viewing optics and fibers to 4-wavelength-channel filter polychrometers. Ratios between the channels determine Te while summations of the channels determine density. Typically, the channel that provides scattered signal at higher Te is contaminated by fluctuating background light. Individual channels are correlated with either a modeled representation of the laser pulse or a noise-free stray light signal to extract channel amplitudes.

We report here results from power balance measurements for ohmically heated plasmas in the sustained spheromak physics experiment. The plasma is formed inside a close-fitting tungsten-coated copper shell; wall conditioning by baking, glow discharge cleaning (GDC), Ti gettering, and helium shot conditioning produces clean plasmas (Z eff < 2.5) and reduces impurity radiation to a small fraction of the input energy, except when the molybdenum divertor plate has been overheated. We find that most of the input energy is lost by conduction to the walls (the divertor plate and the inner electrode in the coaxial source region). Recently, carborane was added during GDC to boronize the plasma-facing surfaces, but little benefit was obtained.

In the DIII-D tokamak, we measure the downstream target plate heat flux with an IRTV camera and relate it to Thomson and Langmuir probe profile measurements of ne and Te in the scrape-off layer (SOL) by projecting all measurements to the out-board midplane in order to assess the competition between parallel and cross-field heat transport. We analyze the measured characteristic widths associated with the SOL and model the results with the UEDGE code to provide insight into the mechanisms defining the various widths and the implications they have on transport. Analysis of the scaling of the heat flux width reveals an essentially inverse dependence on Ip. The Bt dependence is extremely weak to non-existent. No dependence was found on the shear and normalized pressure gradient at the 95% flux surface (s95, α95), nor on PSOL, ne or Pinj.

In this paper we report on density and impurity measurements in the Sustained Spheromak Physics Experiment (SSPX) which has recently started operation. The SSPX spheromak plasma is sustained by coaxial helicity injection for a duration of 2msec with peak toroidal currents of up to 0.5MA. The plasma-facing components consist of tungsten-coated copper to minimize sputtering. The surfaces are conditioned by a combination of baking at 150 C, glow discharge cleaning, Titanium gettering, and pulse-discharge cleaning with helium plasmas. In this way we can achieve density control so that the plasma density ({approx} 1-4 x 10{sup 20}m{sup -3}) matches the gas input. Low-density operation is presently limited by breakdown requirements, but we hope that new gas valves with supersonic nozzles will allow for a further reduction in density. We find that the conditioning reduces the impurity radiation to the point where it is no longer important to the energy balance, and long-lived spheromak plasmas are obtained (decay times of 1.5msec).

The Sustained Spheromak Physics Experiment (SSPX) will examine the confinement properties of spheromak plasmas sustained by DC helicity injection. Understanding the plasma-surface interactions is an important component of the experimental program since the spheromak plasma is in close contact with a stabilizing wall (flux conserver) and is maintained by a high current discharge in the coaxial injector region. Peak electron temperatures in the range of 400 eV are expected, so the copper plasma facing surfaces in SSPX have been coated with tungsten to minimize sputtering and plasma contamination. Here, we report on the characterization and conditioning of these surfaces used for the initial studies of spheromak formation in SSPX. The high pressure plasma-sprayed tungsten facing the SSPX plasma was characterized in situ using β-backscattering and ex situ using laboratory measurements on similarly prepared samples. Measurements showed that water can be desorbed effectively through baking while the removal rates of volatile impurity gases during glow discharge and shot conditioning indicated a large source of carbon and oxygen in the porous coating.

We examine the scaling of the peak divertor heat flux and total divertor plate power in partially detached divertor (PDD) discharges in DIII-D, as a function of input power and radiated power. The peak divertor heat flux in the attached plasma increases linearly with input power, but saturates in the detached cases. The total divertor plate power remains linear with input power in both the attached and detached plasmas. This is consistent with the fact that the heat flux peak is reduced from the attached case but other areas receive increased radiant heating from the detached plasma. The divertor plate radiant heating is linear with input power because the total radiated power from the entire plasma is a linear function of input power in both attached and detached plasma. In the private flux region, radiated heat flux absorbed on the target plate calculated from bolometer data is enough to account for the measured plate heating. Approximately half of the overall plate heating power in detached plasma is due to absorbed radiation.

The role of neutrals in triggering the H–L back transition in high density ELMing H-mode plasmas is explored in double-null (DN) and single-null (SN) divertors. We propose that the neutral particle buildup below the X-point may play an important role in triggering the H–L transition at high density.

Under certain conditions, high temperature electrons diffusing across the separatrix near the midplane can propagate to the divertor without collision and produce a high energy tail in the electron distribution function near the separatrix. We evaluate a kinetic criterion along the field using measurements from divertor Thomson scattering and compare temperature measurements from target plate Langmuir probes to estimate the magnitude of the high energy tail. A Fokker–Planck kinetic model [O. Batishchev, Phys. Plasmas 4 (1998) 1672], which includes electron–electron, electron-ion, electron-neutral, and charge exchange interactions is used to estimate the size of the high energy tail. For the cases considered, we found less than 1% of non-thermal particles near the target plate at typical densities and power levels for the DIII-D scrape-off layer plasma. ELMs violate our kinetic criterion.

Experiments on DIII-D have demonstrated the efficacy of using induced scrape-off-layer (SOL) flow to preferentially enrich impurities in the divertor plasma. This SOL flow is produced through simultaneous deuterium gas injection at the midplane and divertor exhaust. Using this SOL flow, an improvement in enrichment (defined as the ratio of impurity fraction in the divertor to that in the plasma core) has been observed for all impurities in trace-level experiments (i.e., impurity level is non-perturbative), with the degree of improvement increasing with impurity atomic number. In the case of argon, exhaust gas enrichment using modest SOL flow is as high as 17. Using this induced SOL flow technique and argon injection, radiative plasmas have been produced that combine high radiation losses ( P rad / P input > 70%), low core fuel dilution ( Z eff τ E > 1.0 τ E,ITER93H ).

In this paper we describe the newly installed divertor Thomson scattering system for the DIII-D tokamak and present initial results from plasma discharges. Measured plasma densities have ranged from 5 × 1018 to 5 × 1020 m−3 and divertor plasma temperatures from 1 to 500 eV. These data are compared with earlier Langmuir probe data and qualitatively compared with UEDGE computer simulations. The divertor Thomson system uses one of the eight existing core Thomson scattering lasers (1 J, 20 Hz) which has been re-directed to probe the divertor region of the DIII-D vessel. Scattered light from this multipulse Nd:Yag laser is viewed with an f/6.8 collection optics system which provides eight spatial channels from 1–21 cm above the vessel floor (divertor target), each with 1.5 cm vertical resolution. Translating the plasma across the vessel floor using position controls provides a full scan of the divertor plasma.

In this paper the authors present measurements of the global power and particle balance in the high-performance phase of negative central magnetic shear (NCS) discharges and compare with reference VH-mode discharges. The principal differences observed are that NCS has a much lower fraction of the total input power flowing into the boundary, less core radiation, and larger rate of stored energy increase as a fraction of total power. Scrape-off layer (SOL) temperature and divertor heat flux profiles, and radiation profiles at the midplane, are similar to VH-mode. Due to the good core particle confinement and efficient fueling by neutral beam injection (NBI), with little gas puffing, the gas load on the walls and the recycling are very low during the NCS discharges. This results in a rate of density rise relative to beam fueling at the L to H transition time which is 1/3 of the value for VH transitions, which is in turn 1/2 that for L-to-ELMing-H-mode transitions.

Electron density, temperature and parallel pressure measurements at several locations along field lines connecting the midplane scrapeoff layer (SOL) with the outer divertor are presented for both attached and partially-detached divertor cases: Ip = 1.4 MA, q95 = 4.2 and Pinput ∼ 6.7 MW under ELMing H-mode conditions. At the onset of the Partially Detached Divertor (PDD), a high density, low temperature plasma forms in the divertor SOL (divertor MARFE). The electron pressure drops by a factor of 2 between the midplane separatrix and the X-point and then an additional ∼ 3–5 times between the X-point and the outboard separatrix strike point. These results are in contrast to the attached (non-PDD) case, where electron pressure in the SOL is reduced by, at most, a factor of two between the midplane and the divertor target. Divertor MARFEs generally have only marginal adverse impact on important H-mode characteristics, such as confinement time. In fact, PDD discharges at low input power (i.e., approximately twice the L-H-mode threshold power) maintain good H-mode characteristics until a high density, low temperature plasma abruptly forms inside the separatrix near the X-point (X-point MARFE). Concurrent with the appearance of this X-point MARFE is a degradation in both energy confinement and the plasma fueling rate and an increase in the carbon impurity concentration inside the core plasma. The formation of the X-point MARFE is consistent with a thermal instability resulting from the temperature dependence of the carbon radiative cooling rate in the range ∼ 7–30 eV.

We have obtained the first measurements of n{sub e} and T{sub e} in the DIII-D divertor region with a multi-pulse (20 Hz) Divertor Thomson Scattering (DTS) system. Eight measurement locations are distributed vertically up to 21 cm above the divertor plate. Two-dimensional distributions have been obtained by sweeping the divertor plasma across the DTS measurement location. Several operating modes have been studied, including ohmic, L-mode, Elming H-mode, and Radiative Divertor operation with puffing of D{sub 2} and impurities. Mapping of the data to either the (L{sub pol}, {phi}) or (R, Z) planes with the EFIT equilibrium is used to analyze the 2D profiles. We find that in ELMing H-mode: n{sub e}, T{sub e}, and P{sub e} are relatively constant along field lines from the X-point to the divertor plate, especially near the separatrix field line. With D{sub 2} puffing, the DTS profiles indicate that T{sub e} in a large part of divertor region below the X-point is dramatically reduced from {approximately}30-40 eV in ELMing H-mode to 1-2 eV. This results in a fairly uniform low-T{sub e} divertor, with an increased electron density in the range of 2 to 4 x 10{sup 20} m{sup -3}. Detailed comparisons of the spatial profiles of n{sub e}, T{sub e}, and electron pressure P{sub e}, are presented for several operating modes. In addition, these data are compared with initial calculations from the UEDGE fluid code.

The authors characterize the divertor target plate heat and particle fluxes that occur due to Edge-Localized-Modes (ELMs) during H-mode in DIII-D and ASDEX-Upgrade. During steady-state ELMing H-mode the fraction of main plasma stored energy lost with each ELM varies from 6% to 2% as input power increases above the H-mode power threshold. The ELM energy is deposited near the strikepoints on the divertor target plates in a fast time scale of {le} 1 ms. The spatial profile of the ELM heat pulse is flatter and broader, up to about a factor of 2, than that of the heat flux between ELMs. On ASDEX-Upgrade the inboard strike-point receives the greatest fraction, {ge} 75%, of ELM divertor heat flux, while on DIII-D the in/out split is nearly equal. The toroidal asymmetry of the heat pulse has produced a peaking factor on DIII-D of no more than 1.5. The particle flux, as measured by Langmuir probes, has also been found to be localized near the divertor strike-points. The increased particle flux during ELMs is a significant fraction of the total time-integrated divertor plate particle flux.

Edge localized modes (ELMs) are the focus of increasing attention by the edge physics community because of the potential impact that the large divertor heat pulses due to ELMs would have on the divertor design of future high power tokamaks such as ITER. This paper reviews what is known about ELMs, with an emphasis on their effect on the scrape-off layer and divertor plasmas. ELM effects have been measured in the ASDEX-U, C-Mod, COMPASS-D, DIII-D, JET, JFT-2M, JT-60U and TCV tokamaks, and are reported here. At least three types of ELMs have been identified and their salient features determined. Type 1 giant ELMs can cause the sudden loss of up to 10–15% of the plasma stored energy, but their amplitude ( ΔW W ) does not increase with heating power. Type 3 ELMs are observed near the H-mode power threshold and produce small energy dumps (1–3% of the stored energy). All ELMs increase the scrape-off layer plasma and produce particle fluxes on the divertor targets which are as much as ten times larger than the quiescent phase between ELMs. The divertor heat pulse is largest on the inner target, unlike that of L-mode or quiescent H-mode; some tokamaks report radial structure in the heat flux profile which is suggestive of islands or helical structures. The power scaling of type 1 ELM amplitude and frequency has been measured in several tokamaks and has recently been applied to predictions of the ELM size in ITER. Concern over the expected ELM amplitude has led to a number of experiments aimed at demonstrating active control of ELMs. Impurity gas injection with feedback control on the radiation loss in ASDEX-U suggests that a promising mode of operation (the CDH-mode) with very small type 3 ELMs can be maintained with heating power well above the H-mode threshold, where giant type 1 ELMs are normally observed. While ELMs have many potential negative effects, the beneficial effect of ELMs in providing density control and limiting the core plasma impurity content in high confinement H-mode discharges should not be overlooked.

A new reciprocating Langmuir probe was used to measure density and temperature profiles, ion flow, and potential fluctuation levels from the lower divertor floor up to the X point on the DIII-D Tokamak. This probe is designed to make fast (2 kHz swept, 20 kHz Mach, 500 kHz Vfloat) measurements with 2 mm spatial resolution in the region where the largest gradients on the plasma open flux tubes are found and therefore provide the best benchmarks for scrap-off layer and divertor numerical models. Profiles are constructed using the 300 ms time history of the probe measurements during the 25 cm reciprocating stroke. Both single and double null plasmas can be measured and compared with a 20 Hz divertor Thomson scattering system. The probe head is constructed of four different kinds of graphite to optimize the electrical and thermal characteristics. Electrically insulated pyrolytic graphite rings act as a heat shield to absorb the plasma heat flux on the probe shaft and are mounted on a carbon/carbon composite core for mechanical strength. The Langmuir probe sampling tips are made of a linear carbon fiber composite. The mechanical, electrical, data acquisition, and power supply systems will be described. Initial measurements will also be presented.

The DIII–D Thomson scattering system has been expanded to measure divertor plasma temperatures from 1 to 500 eV and densities from 0.05 to 8×1020 m−3. To complete this system, a difficult stray light problem was overcome to allow for an accurate Rayleigh scattering density calibration. The initial stray light levels were over 500 times higher than the expected Rayleigh scattered signal. Using a charge-coupled device (CCD) camera, various portions of the vessel interior were examined while the laser was fired through the vessel in air at atmospheric pressure. Image relaying, exit window tilting, entrance and exit baffle modifications, and a beam polarizer were then used to reduce the stray light to acceptable levels. The CCD camera gave prompt feedback on the effectiveness of each modification, without the need to reestablish vacuum conditions required when using the normal avalanche photodiode detectors (APD). Once the stray light was sufficiently reduced, the APD detectors provided the signal time history to more accurately identify the source location. We have also found that certain types of high reflectance dielectric coatings produce 10–15 times more scatter than other types of more conventional coatings. By using low-scatter mirror coatings and these new stray light reduction techniques, we now have more flexibility in the design of complex Thomson scattering configurations required to probe the central core and the new radiative divertor regions of the DIII–D vessel.

As fusion experiments progress towards long pulse or steady state operation, plasma facing components are undergoing a significant change in their design. This change represents the transition from inertially cooled pulsed systems to steady state designs of significant power handling capacity. A limited number of Plasma Facing Component (PFC) systems are in operation or planning to address this steady state challenge at low heat flux. However in most divertor designs components are required to operate at heat fluxes at 5 MW/m{sup 2} or above. The need for data in this area has resulted in a significant amount of thermal/hydraulic and thermal fatigue testing being done on prototypical elements. Short pulse design solutions are not adequate for longer pulse experiments and the areas of thermal design, structural design, material selection, maintainability, and lifetime prediction are undergoing significant changes. A prudent engineering approach will guide us through the transitional phase of divertor design to steady-state power plant components. This paper reviews the design implications in this transition to steady state machines and the status of the community efforts to meet evolving design requirements. 54 refs., 5 figs., 2 tabs.

A line-of-sight vapour deposition process is used to form a low-voltage field emission cathode structure on the surface of an oxide-metal eutectic composite etched to expose arrays of single-crystal tungsten fibres. A SiO2 insulating film is first deposited, forming cone-shaped deposits on the tips of the exposed fibres. These cones act as a shadow mask for the subsequent deposition of a metal film. After removal of the cones, the metal film forms a grid-like structure with concentric apertures centered on each fibre tip. Application of a potential of less than 100 V between the metal grid and the base of the fibres produces an electric field at the fibre tips of sufficient intensity to produce field emission. The equilibrium geometry of the deposits which form on the fibre tips (cathode cones) is such that there is a linear relationship between the height of the cone and its diameter. There is also a direct correlation between the cathode cone angle, which depends on the material being deposited, and the cone angle of the insulator film. For SiO2, the cathode cone angle was determined to be 37°, while the cone angle of the insulator film was 26°. Comparison of the theoretical values of the lateral growth velocity of the cathode cones with values determined from experimental measurements indicated that the maximum sticking coefficient was independent of the growth angle, η, for SiO2, but varied as a function of cos η for molybdenum. A multiple deposition process was developed which permitted independent control of the interelectrode spacing and insulator film thickness.

A comprehensive set of experiments has been carried out on the DIII-D tokamak to measure the 2-D (R,Z) structure of the divertor plasma in a systematic way using new diagnostics. Measurements cover the divertor radially from inside the X-point to the outer target plate and vertically from the target plate to above the X-point. Identical, repeatable shots were made, each having radial sweeps of the X-point and divertor strike points, to allow complete plasma and radiation profile measurements. Data have been obtained in ohmic, L-mode, ELMing H-mode, and reversed B{sub T} operation ({gradient}B drift away from the X-point). In addition, complete measurements were made of radiative divertor plasmas with a Partially Detached Divertor (PDD) induced by D{sub 2} injection and with a Radiating Mantle induced by Impurity injection (RMI) using neon and nitrogen. The data set includes first observations of the radial and poloidal profiles of the X-point, inner and outer leg plasmas in PDD and RMI radiative divertor operation. Preliminary data analysis shows that intrinsic impurities play a critical role in determining the SOL and divertor conditions.

DIII-D currently operates with a single- or double-null open divertor and graphite walls. Active particle control with a divertor cryopump has demonstrated density control, efficient helium exhaust, and reduction of the inventory of particles in the wall. Gas puffing of D{sub 2} and impurities has demonstrated reduction of the peak divertor beat flux by factors of 3--5 by radiation. A combination of active cryopumping and feedback-controlled D{sub 2} gas puffing has produced similar divertor heat flux reduction with density control. Experiments with neon puffing have shown that the radiation is equally-divided between a localized zone near the X-point and a mantle around the plasma core. The density in these experiments has also been controlled with cryopumping. These experimental results combined with modeling were used to develop the new Radiative Divertor for DIII-D. This is a double-null slot divertor with four cryopumps to provide particle control and neutral shielding for high-triangularity advanced tokamak discharges. UEDGE and DEGAS simulations, benchmarked to experimental data, have been used to optimize the design.

We account for ≥85% of the injected power in DIII-D during single-null diverted ELMing H-mode discharges. Core plasma radiation accounts for ≤18% with the rest flowing through the scrape-off layer to the inboard and outboard divertors. The total power roughly splits in an in: out ratio of 1:1.3. Of the power entering the outboard divertor 75% leaves as direct heat flux to the divertor target plates with the rest in the form of radiation. In the inboard divertor, however, 75% of the power leaves as radiation with the rest as divertor target plate heat flux and charge exchange losses. The in: out ratio remains constant over a range of injected power and plasma current and even as divertor conditions change due to gas puffing.

In this paper we discuss the present divertor design for the planned TPX tokamak, which will explore the physics and technology of steady-state (1000s pulses) heat and particle removal in high confinement (2--4{times} L-mode), high beta ({beta}{sub N} {ge} 3) divertor plasmas sustained by non-inductive current drive. The TPX device will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.5 m) slot at the outer strike point. The peak heat flux on, the highly tilted (74{degrees} from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4--6 MW/m{sup 2} with 18 MW of neutral beams and RF heating power. The combination of active pumping and gas puffing (deuterium plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities.

We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized (∼ 10 cm diameter) radiation zone which results in substantial reduction (3–10) in the divertor heat flux while τ E remains ∼ 2 times ITER-89P scaling. However, n e increases with D 2 puffing, and Z eff increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity (δ ≈ 0.8) is important for high τ E VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented.

We present a study of the impact of H-mode operation at low density on divertor plasma parameters on the DIII-D tokamak. The line-average density in H-mode was scanned by variation of the particle exhaust rate, using the recently installed divertor cryo-condensation pump. The maximum decrease (50%) in line-average electron density was accompanied by a factor of 2 increase in the edge electron temperature, and 10% and 20% reductions in the measured core and divertor radiated power, respectively. The measured total power to the inboard divertor target increased by a factor of 3, with the major contribution coming from a factor of 5 increase in the peak heat flux very close to the inner strike point. The measured increase in power at the inboard divertor target was approximately equal to the measured decrease in core and divertor radiation.

Non-stationary toroidal asymmetries are observed in the DIII-D divertor heat flux and scrape-off layer (SOL) currents. Using the present DIII-D diagnostics, asymmetries are seen much less frequently in single-null H-modes ( 50%). Divertor heat flux asymmetries are characterized by toroidal variations in the radial profile (i.e., multiple or bifurcated peaks) at some toroidal locations and single peaks at other toroidal locations while SOL currents sometimes have a strongly bipolar toroidal structure. SOL current asymmetries are particularly large during Edge Localized Modes (ELMs). The measurements reported here indicate that the asymmetries are best described by a model in which non-axisymmetric radial magnetic perturbations create magnetic islands in the plasma boundary and SOL which then cause toroidal variation in the divertor heat flux and the SOL currents.

This paper presents work to develop benchmarked theoretical models of scrape-off-layer (SOL) characteristics in diverted tokamaks by comparing shot simulations, using the UEDGE plasma fluid and DEGAS neutral transport codes, to measurements of the DIII-D SOL plasma. The experimental data include the radial profiles of ne, Te, and Ti, the divertor exhaust power, the intensity of Hα emission, and profiles of the radiated power. A very simple model of the anomalous perpendicular transport rates produces consistency between the calculated and measured radial profiles of the divertor power, and of the midplane densities and temperatures. Experimentally, the measured exhaust power is now 80–90% of the input power. The simulated peak power on the outer leg of the divertor floor is now within 20% of the measured power. Various sensitivies of these comparisons to model assumptions are described. Finally, these benchmarked models have been used to examine the effects of various baffle configurations for the radiative divertor upgrade in DIII-D.

In this paper, we examine a denser and broader scrape-off layer (SOL) plasma, one class of which was first seen in VH mode, in the DIII-D tokamak. An enhanced SOL appears in many types of double-null (DN) discharges and is not a property of VH mode only. The L and H mode DN enhanced SOL density and temperature profiles exhibit a 5–6 cm broad profile outside the separatrix. The VH mode enhanced SOL is associated with a higher separatrix density and temperature and is only 1.5–2 cm wide. Both types of SOL enhancement are only observed in high triangularity discharges. The origin of the enhanced SOL is, however, not yet understood.

The achievement of a long-pulse ignited discharge with over 1000 MW of fusion power in the International Thermonuclear Experimental Reactor will be an important goal for the next phase of the world magnetic fusion program. However, improvements in the physics are needed to design a more economically attractive tokamak power reactor than the present data base would support. Advanced, steady state plasma controls are the key to realizing these improvements. The Tokamak Physics Experiment has a flexible heating and current drive system for profile control; a flexible poloidal field system that supports a strongly shaped double-null poloidal divertor plasma configuration over a wide range of profiles; and a divertor designed for dispersive operation, flexibility, and remote handling. The machine performance in deuterium is sufficient to produce a reactor-like bootstrap current profile and to confine fast electrons for localized current profile control. A conducting structure, plasma rotation, field error compensation coils, and profile control are used to provide stable plasma configurations with β up to twice the Troyon limit and bootstrap current fraction approaching unity. The facility will be designed for 1000 s pulses initially to minimize the influence of initial transients on system behavior, but the pulse length can be extended through upgrades of external systems if necessary.

The quality of plasma produced in a magnetic confinement fusion device is influenced to a large extent by the neutral gas surrounding the plasma. The plasma is fueled by the ionization of neutrals, and charge exchange interactions between edge neutrals and plasma ions are a sink of energy and momentum. Here we describe a diagnostic capable of measuring the spatial distribution of neutral gas in a magnetically confined fusion plasma. A high intensity (5 MW/cm(2)), narrow bandwidth (0.1 cm(-1)) laser is injected into a hydrogen plasma to excite the Lyman β transition via the simultaneous absorption of two 205 nm photons. The absorption rate, determined by measurement of subsequent Balmer α emission, is proportional to the number of particles with a given velocity. Calibration is performed in situ by filling the chamber to a known pressure of neutral krypton and exciting a transition close in wavelength to that used in hydrogen. We present details of the calibration procedure, including a technique for identifying saturation broadening, measurements of the neutral density profile in a hydrogen helicon plasma, and discuss the application of the diagnostic to plasmas in the DIII-D tokamak.

We summarize results from DIII-D in regard to issues for reactor application of H-mode. Recently, DIII-D has begun to operate a cryopump (D{sub 2} pumping speed = 31,000 {ell}/s at a pressure of 2 mTorr). Initial results are very favorable for density control in H-mode. The plasma density could be reduced by 50%. Energy confinement is unchanged so the temperature rises in proportion to the density drop. Ability to access these less collisional plasmas in H-mode is favorable to current drive application. With the all-graphite wall, impurity accumulation has been eliminated. The exceedingly good confinement of VH-mode offers the possibility of retaining good confinement while radiating copious power from the plasma edge using injected noble gas impurities.

The Tokamak Physics Experiment is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. It is based on an emerging class of tokamak operating modes, characterized by beta limits well in excess of the Troyon limit, confinement scaling well in excess of H-mode, and bootstrap current fractions approaching unity. Such modes are attainable through the use of advanced, steady state plasma controls including strong shaping, current profile control, and active particle recycling control. Key design features of the TPX are superconducting toroidal and poloidal field coils; actively-cooled plasma-facing components; a flexible heating and current drive system; and a spacious divertor for flexibility. Substantial deuterium plasma operation is made possible with an in-vessel remote maintenance system, a lowactivation titanium vacuum vessel, and shielding of ex-vessel components. The facility will be constructed as a national project with substantial participation by U.S. industry. Operation will begin with first plasma in the year 2000.