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The tokamak à configuration variable (TCV) continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to address critical issues in preparation for ITER and a fusion power plant. For the 2019–20 campaign its configurational flexibility has been enhanced with the installation of removable divertor gas baffles, its diagnostic capabilities with an extensive set of upgrades and its heating systems with new dual frequency gyrotrons. The gas baffles reduce coupling between the divertor and the main chamber and allow for detailed investigations on the role of fuelling in general and, together with upgraded boundary diagnostics, test divertor and edge models in particular. The increased heating capabilities broaden the operational regime to include T e/T i ∼ 1 and have stimulated refocussing studies from L-mode to H-mode across a range of research topics. ITER baseline parameters were reached in type-I ELMy H-modes and alternative regimes with ‘small’ (or no) ELMs explored. Most prominently, negative triangularity was investigated in detail and confirmed as an attractive scenario with H-mode level core confinement but an L-mode edge. Emphasis was also placed on control, where an increased number of observers, actuators and control solutions became available and are now integrated into a generic control framework as will be needed in future devices. The quantity and quality of results of the 2019–20 TCV campaign are a testament to its successful integration within the European research effort alongside a vibrant domestic programme and international collaborations.

Self-consistent full-size turbulent-transport simulations of the divertor and scrape-off-layer (SOL) of existing tokamaks have recently become feasible. This enables the direct comparison of turbulence simulations against experimental measurements. In this work, we perform a series of diverted ohmic L-mode discharges on the tokamak à configuration variable (TCV) tokamak, building a first-of-a-kind dataset for the validation of edge turbulence models. This dataset, referred to as TCV-X21, contains measurements from five diagnostic systems from the outboard midplane (OMP) to the divertor targets—giving a total of 45 one- and two-dimensional comparison observables in two toroidal magnetic field directions. The experimental dataset is used to validate three flux-driven 3D fluid-turbulence models—GBS, GRILLIX and TOKAM3X. With each model, we perform simulations of the TCV-X21 scenario, individually tuning the particle and power source rates to achieve a reasonable match of the upstream separatrix value of density and electron temperature. We find that the simulations match the experimental profiles for most observables at the OMP—both in terms of profile shape and absolute magnitude—while a comparatively poorer agreement is found towards the divertor targets. The match between simulation and experiment is seen to be sensitive to the value of the resistivity, the heat conductivities, the power injection rate and the choice of sheath boundary conditions. Additionally, despite targeting a sheath-limited regime, the discrepancy between simulations and experiment also suggests that the neutral dynamics should be included. The results of this validation show that turbulence models are able to perform simulations of existing devices and achieve reasonable agreement with experimental measurements. Where disagreement is found, the validation helps to identify how the models can be improved. By publicly releasing the experimental dataset and validation analysis, this work should help to guide and accelerate the development of predictive turbulence simulations of the edge and SOL.

Parallel convection and E × B drifts act together to redistribute heat between the strike-points in the low field side snowflake minus (LFS SF−). The cumulative heat convection from both mechanisms is enhanced near the secondary X-point and is shown to dominate over heat conduction, partly explaining why the LFS SF− distributes power more evenly than the single null (SN) or other snowflake (SF) configurations. Pressure profiles at the entrance of the divertor are strongly affected by the position of the secondary X-point and magnetic field direction indicating the importance of E × B drifts. Pressure drops of up to 50% appear between the outer-midplane (OMP) and the divertor entrance enhancing the role of parallel heat convection. The electron temperature and density profiles and the radial turbulent fluxes measured at the OMP are largely unaffected by the changes in divertor geometry, even on flux surfaces where the connection length is infinite.

An overview of recent Plasma-Material Interactions (PMI) research at the DIII-D tokamak using the Divertor Material Evaluation System (DiMES) is presented. The DiMES manipulator allows for exposure of material samples in the lower divertor of DIII-D under well-diagnosed ITER-relevant plasma conditions. Plasma parameters during the exposures are characterized by an extensive diagnostic suite including a number of spectroscopic diagnostics, Langmuir probes, IR imaging, and Divertor Thomson Scattering. Post-mortem measurements of net erosion/deposition on the samples are done by Ion Beam Analysis, and results are modelled by the ERO and REDEP/WBC codes with plasma background reproduced by OEDGE/DIVIMP modelling based on experimental inputs. This article highlights experiments studying sputtering erosion, re-deposition and migration of high-Z elements, mostly tungsten and molybdenum, as well as some alternative materials. Results are generally encouraging for use of high-Z PFCs in ITER and beyond, showing high redeposition and reduced net sputter erosion. Two methods of high-Z PFC surface erosion control, with (i) external electrical biasing and (ii) local gas injection, are also discussed. These techniques may find applications in the future devices.

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.

A simple analytic model for the repartition of the Scrape-Off Layer (SOL) exhaust power between the inner and outer divertors in a diverted low-density tokamak plasma is introduced. Electron heat conduction is assumed to dominate the heat transport, from the outboard mid-plane to the divertor targets, with no heat sinks or sources in the SOL. Both divertor channels are in the attached, high-recycling regime. The model is in reasonable qualitative agreement with recent TCV experimental data and EMC3-Eirene simulations. For the Single Null divertor, it reproduces the experimentally observed increase in the power ratio between the inner and outer divertor plates of TCV, Pin/Pout, with increasing the outer divertor leg length or the outer target flux expansion. For the Snowflake Minus configuration, it reproduces the observed variation of Pin/Pout with X-point separation, although only for the reversed magnetic field direction. Within the model limitations, it provides a basic understanding of the power sharing in alternative divertor geometries. Keywords: Heat exhaust, SOL transport, Long-legged Single Null Divertor, Snowflake Divertor, In-out power sharing, TCV

A study to examine some current issues in the physics of the plasma sheath has been recently carried out in DIII-D low power Ohmic plasmas using both flush and domed Langmuir probes, divertor Thomson scattering (DTS), an infrared camera (IRTV), and a new calorimeter triple probe assembly mounted on the Divertor Materials Evaluation System (DIMES). The sheath power transmission factor was found to be consistent with the theoretically predicted value of 7 (±2) for low power plasmas. Using this factor, the three heat flux profiles derived from the LP, DTS, and calorimeter diagnostic measurements agree. Comparison of flush and domed Langmuir probes and divertor Thomson scattering indicates that proper interpretation of flush probe data to get target plate density and temperature is feasible and could potentially yield accurate measurements of target plate conditions where the probes are located.

ITER will use the shaped inner column as startup/rampdown limiter. An exploratory experiment was performed in DIII-D using a similarly shaped inner wall limiter, looking for evidence of a short fall-off feature, λshort, near the last closed flux surface. Measurements were made on the high field side (HFS) using infrared thermography and a Langmuir swing-probe. In some cases clear evidence was found for a narrow feature λ short ∼ ρ pol D + ∼ a few mm. The ratio q | | 0 short / q | | 0 long was mostly between 0.1 and 0.6. Measurements made on the low field side and bottom side showed little clear indication of any narrow feature.

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.

We present a method to calculate the ion saturation current, Isat, for Langmuir probes at high frequency (>100 kHz) using the harmonics technique and we compare that to a direct measurement of Isat. It is noted that the Isat estimation can be made directly by the ratio of harmonic amplitudes, without explicitly calculating Te. We also demonstrate that since the probe tips using the harmonic method are oscillating near the floating potential, drawing little power, this method reduces tip heating and arcing and allows plasma density measurements at a plasma power flux that would cause continuously biased tips to arc. A multi-probe array is used, with two spatially separated tips employing the harmonics technique and measuring the amplitude of at least two harmonics per tip. A third tip, located between the other two, measures the ion saturation current directly. We compare the measured and calculated ion saturation currents for a variety of plasma conditions and demonstrate the validity of the technique and...

Intrinsic toroidal rotation of the deuterium main ions in the core of the DIII-D tokamak is observed to transition from flat to hollow, forming an off-axis peak, above a threshold level of direct electron heating. Nonlinear gyrokinetic simulations show that the residual stress associated with electrostatic ion temperature gradient turbulence possesses the correct radial location and stress structure to cause the observed hollow rotation profile. Residual stress momentum flux in the gyrokinetic simulations is balanced by turbulent momentum diffusion, with negligible contributions from turbulent pinch. The prediction of the velocity profile by integrating the momentum balance equation produces a rotation profile that qualitatively and quantitatively agrees with the measured main-ion profile, demonstrating that fluctuation-induced residual stress can drive the observed intrinsic velocity profile.

In inboard-limited plasmas, the Scrape-Off Layer (SOL) shows two regions: the near SOL, extending a few mm from the Last Closed Flux Surface (LCFS), characterized by a steep gradient of the parallel heat flux radial profile, and a far SOL, typically some cm wide, with flatter heat flux profiles. The physics of the near SOL is investigated in TCV with two series of experiments featuring deuterium and helium plasmas, in which the plasma current, density and elongation have been varied. The parallel heat flux profiles are measured on the limiter by means of infrared thermography. For the first time, the near SOL is reported to disappear for low plasma current or at high density, for values of the SOL collisionality νlowast corresponding to a conduction-limited regime. The power in the near SOL ∆PSOL is shown to decrease with the normalized Spitzer resistivity ν as ∆PSOL ν−1. The floating potential profiles, measured at the limiter using flush-mounted Langmuir probes (LP), show the presence of non-ambipolar currents, and their relation to the presence of a velocity shear layer is discussed. The shearing rate is shown to strictly correlate with the power in the near SOL ∆PSOL, consistently with a recent theoretical model. Measurements of the near SOL on the Low Field Side (LFS) are performed using a reciprocating Langmuir probe (RP). The near SOL is reported to vanish simultaneously at the LFS and at the limiter. The near and far SOL widths are compared with the predictions from existing theoretical models, to which empirical corrections with resistivity and elongation are proposed.

A set of Ohmic density ramp experiments addressing the role of parallel connection length in modifying scrape off layer (SOL) properties has been performed on the TCV tokamak. The parallel connection length has been modified by varying the poloidal flux expansion f x . It will be shown that this modification does not influence neither the detachment density threshold, nor the development of a flat SOL density profile which instead depends strongly on the increase of the core line average density. The modification of the SOL upstream profile, with the appearance of what is generally called a density shoulder , has been related to the properties of filamentary blobs. Blob size increases with density, without any dependence on the parallel connection length both in the near and far SOL. The increase of the density decay length, corresponding to a profile flattening, has been related to the variation of the divertor normalized collisionality ##IMG## [http://ej.iop.org/images/0029-5515/57/11/116014/nfaa7db3ieqn001.gif] {$\Lambda_{\rm div}$} (Myra et al 2006 Phys. Plasmas 13 112502, Carralero et al , ASDEX Upgrade Team, JET Contributors and EUROfusion MST1 Team 2015 Phys. Rev. Let . 115 215002), showing that in TCV the increase of ##IMG## [http://ej.iop.org/images/0029-5515/57/11/116014/nfaa7db3ieqn002.gif] {$\Lambda_{\rm div}$} is not sufficient to guarantee the SOL upstream profile flattening.

The snowflake (SF) magnetic configuration is investigated as a potential power exhaust solution for a fusion reactor, but also to improve our understanding of divertor physics in general. Unlike a conventional single null (SN), it features an additional nearby x-point, which deeply modifies the magnetic field in the scrape-off layer (SOL) and can thereby affect parallel and cross-field transport of heat and particles. This paper investigates the power exhaust properties of the snowflake minus (SF-) configuration on the TCV tokamak, for Ohmically heated, L-mode, low-density, attached plasmas for a range of x-point separations, magnetic field directions and locations of the secondary x-point (low-field-side (LFS) or high-field-side (HFS)). Due to the relatively large x-point separation in physical space, this study probes x-point transport features in general, rather than reactor relevant aspects of the SF configuration. The target heat fluxes at all strike points are simultaneously monitored with an infrared (IR) thermography system and the kinetic profiles of the SOL at the outer mid-plane with a reciprocating probe (RCP). The placement of the additional x-point is seen to affect the inner-outer divertor power balance. An effective heat flux width lambda(eff)(q,u) for the SOL in the low poloidal field region is inferred from the measured power repartition between the two SOL manifolds created by the secondary x-point. For the HFS SF- configuration (secondary null in the LFS SOL), the lambda(eff)(q,u) is two times larger than that measured by the RCP at the outer mid-plane and by IR at the outer target of a comparable SN, while the outer mid-plane SOL profiles are similar to the SN. This is interpreted as the effect of increased effective cross-field diffusivity chi(null)(perpendicular to) in the intra-null region relative to the rest of the SOL. For the HFS SF-configuration (secondary null in the HFS SOL), no such increased transport is observed. The pressure-driven plasma convection expected near the primary null cannot explain the increased chi(null)(perpendicular to), which is instead consistent with interchange ballooning-like turbulence enhanced by the low poloidal field.

The OEDGE code is used to model core fueling for attached L-mode plasmas and between edge localized modes (ELMs) for attached H-mode plasmas in DIII-D. Empirical plasma reconstruction has been used to determine the plasma conditions in these discharges. EIRENE is used to model the hydrogen recycling. Divertor recycling accounts for 65–100% of the core fueling. The fraction of the total divertor target flux ionized inside the separatrix ranges from 5% to 20%. The fraction of total wall flux ionized inside the separatrix ranges from 20% to 50%. Neutrals originating from wall regions closer to the separatrix are more likely to ionize in the confined plasma. Ionization in the confined plasma is concentrated below the midplane with peaks in the poloidal profiles just above the X -point. Radial core ionization in high density H-mode is peaked strongly near the separatrix.

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.

Present research towards a plasma exhaust solution for a fusion power plant aims at validating edge physics models to strengthen the predictive capabilities and improving the operating regime and the divertor configuration. The TCV tokamak is used in particular to investigate to what extent geometric modifications of the configuration can affect the plasma exhaust performance by decreasing the peak heat loads, improving the control of the detached condition and enhancing the compatibility of the divertor solution with core performance. Detachment experiments in TCV have so far been restricted to Ohmically heated L-mode plasmas. Recent TCV experiments connect to previous detachment studies using standard single-null configurations, where the onset of detachment is best observed at sufficiently high density when the direction of the magnetic field results in an ion rB drift away from the X-point. The studies were extended to nitrogen seeding and to an entire suite of alternative magnetic configurations, including the variation of the poloidal flux expansion at the outer target, the introduction of flux flaring towards the target (X divertor), the movement of the secondary X-point inside the vessel (X-point target) as well as various snowflake configurations. The snowflake minus configuration has attracted particular interest as recent EMC3-EIRENE simulations predict a large region of radiation between two X-points. In addition to the detachment characteristics the effect of the altered connection length and magnetic shear on turbulent transport is investigated. The recent experiments benefited from a range of improved diagnostic capabilities including a new divertor spectroscopy system, an upgraded infrared thermography system, an extended set of wall-mounted Langmuir probes and a newly installed reciprocating probe. Experiments in 2016 will also be able to take advantage of the newly installed neutral beam injection (NBI) heating system, which allows for auxiliary heating of high density plasmas and thereby for testing of potential solutions under higher heat fluxes and in regimes with a higher SOL pressure.

An experiment aimed at benchmarking the ITER scrape-off layer (SOL) power width scaling in limited L-mode discharges has been conducted on DIII-D. Scans of the main scaling parameters were performed in an inner-wall-limited (IWL) magnetic configuration. Using the near-SOL density and temperature e-folding lengths, λ n , λ T , determined from reciprocating Langmuir probe measurements, SOL power flux density e-folding lengths, λ q , are derived. A few lower single null (LSN) discharges were also run for comparison. The results are generally in agreement with the ITER design assumptions, finding that λ n and λ T are correlated ( λ T ∼ 1.2 λ n ) and both λ n and λ T are on average 2.1–2.5 times larger in IWL configurations than in LSN. In moderate elongation (κ ∼ 1.4) IWL discharges, λ q is largest and agrees with the assumed ITER scaling within the estimated uncertainty (a factor of ∼2). In IWL discharges λ q measurements are consistent with the expectations of SOL power balance.

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.

The three-dimensional (3D) features of plasma edge profiles and wall interaction patterns induced by edge resonant magnetic perturbation fields (RMP) are discussed comparing TEXTOR and DIII-D. We show that the scrape-off layer (SOL) profiles and decay lengths depend the edge safety factor, the RMP base mode as well as on the plasma rotation during RMP application indicating modification of SOL transport by the 3D perturbation fields. This is compatible with channelling of particle and heat efflux along open perturbed field lines in the very edge of the plasma boundary into a completely re-arranged, helically striated 3D divertor footprint. The distribution of the measured divertor heat and particle fluxes at DIII-D match the vacuum modelled magnetic footprint topology in L-mode while in H-mode the striation width exceeds the modelled footprint width by 15–30%. This 3D structure of the measured heat and particle fluxes results in a new situation for the material erosion properties and initial quantification of the net-erosion within the 3D footprint shows in L-mode a 50% decrease of the chemical erosion yield and evidence for a comparably small 15-20% increase in physical sputtering. Extrapolation of these findings to ITER by vacuum modelling of the magnetic footprint for the actual ELM control coils shows a similar vacuum magnetic footprint topology as found at DIII-D during RMP ELM suppression. However, the open field lines escape the CFC covered ITER divertor area potentially transferring net-erosion characteristics from the CFC domain onto the Tungsten including so far unconsidered heat and particle loads on this sensitive material.

Major developments on the Mega Amp Spherical Tokamak (MAST) have enabled important advances in support of ITER and the physics basis of a spherical tokamak (ST) based component test facility (CTF), as well as providing new insight into underlying tokamak physics. For example, L–H transition studies benefit from high spatial and temporal resolution measurements of pedestal profile evolution (temperature, density and radial electric field) and in support of pedestal stability studies the edge current density profile has been inferred from motional Stark effect measurements. The influence of the q-profile and E × B flow shear on transport has been studied in MAST and equilibrium flow shear has been included in gyro-kinetic codes, improving comparisons with the experimental data. H-modes exhibit a weaker q and stronger collisionality dependence of heat diffusivity than implied by IPB98(y,2) scaling, which may have important implications for the design of an ST-based CTF. ELM mitigation, an important issue for ITER, has been demonstrated by applying resonant magnetic perturbations (RMPs) using both internal and external coils, but full stabilization of type-I ELMs has not been observed. Modelling shows the importance of including the plasma response to the RMP fields. MAST plasmas with q > 1 and weak central magnetic shear regularly exhibit a long-lived saturated ideal internal mode. Measured plasma braking in the presence of this mode compares well with neo-classical toroidal viscosity theory. In support of basic physics understanding, high resolution Thomson scattering measurements are providing new insight into sawtooth crash dynamics and neo-classical tearing mode critical island widths. Retarding field analyser measurements show elevated ion temperatures in the scrape-off layer of L-mode plasmas and, in the presence of type-I ELMs, ions with energy greater than 500 eV are detected 20 cm outside the separatrix. Disruption mitigation by massive gas injection has reduced divertor heat loads by up to 70%.

Stochastic boundaries in fusion devices have been investigated in tokamaks, stellarators and reversed field pinch experiments for many years. However, since edge localized modes (ELMs) have been successfully eliminated in H-mode plasmas at the DIII-D tokamak [1,2] with small, edge resonant magnetic perturbations, they have become a widely investigated topic in tokamaks. In DIII-D stochastic boundaries are produced by coils external to the plasma. The magnetic field there consists of field lines with very different connection lengths, which produces a three dimensional, heterogenous structure of stochastic volume. The most obvious manifestation of the perturbed plasma edge is the strike line splitting observable in heat and particle fluxes, which changes with collisionality. The interaction of the magnetic perturbation and the magnetic equilibrium is of a resonant nature and the structure of the stochastic volume is a strong function of q(95). This is observed as a modulation of e.g. electron temperature as measured by ECE or Thomson scattering. In this work we summarize recent experimental findings on properties of the stochastic boundary in DIII-D. (C) 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim

A Child-Langmuir law-based method for accounting for Debye sheath expansion while fitting the current-voltage I-V characteristic of proud Langmuir probes (electrodes that extend into the volume of the plasma) is described. For Langmuir probes of a typical size used in tokamak plasmas, these new estimates of electron temperature and ion saturation current density values decreased by up to 60% compared to methods that did not account for sheath expansion. Changes to the collection area are modeled using the Child-Langmuir law and effective expansion perimeter lp, and the model is thus referred to as the “perimeter sheath expansion method.” lp is determined solely from electrode geometry, so the method may be employed without prior measurement of the magnitude of the sheath expansion effects for a given Langmuir probe and can be used for electrodes of different geometries. This method correctly predicts the non-saturating ΔI/ΔV slope for cold, low-density plasmas where sheath-expansion effects are strong, as...

Flow velocities of the order 10-20 km/s in the direction of the high-field side divertor have been measured for deuterons and low charge-state carbon ions in the scrape-off layer at the crown of low-density L-mode plasmas, suggesting that these carbon ions at the crown move with the background plasma flow. Simulations with the multi-fluid edge code UEDGE including cross-field drifts due to E x B and B x del B yield calculated divertor conditions which are more consistent with the measurements, but flows at the crown that are stagnant or in the opposite direction than observed. The simulations indicate that both the ion temperature gradient force and deuteron frictional drag play a role in determining the flow direction and magnitude of low charge-state carbon ions. The effect of the assumed radial transport model, toroidal core rotation, and neutral pumping at the divertor plates on the flow at the crown is investigated. (C) 2009 Elsevier B.V. All rights reserved.

13 C methane puffing experiments were conducted on DIII-D in both L- and ELMy H-mode conditions. The puffing was toroidally symmetric into the crown of a series of well-characterized lower single-null discharges. The hydrocarbon breakup, carbon transport and deposition were modeled using the OEDGE interpretive code. Three separate hypotheses were tested using OEDGE to try to reproduce the experimental deposition: Radial variation of fast parallel flow, erosion of the puffed 13 C deposited in the divertor, and a pinch in the inner scrape off layer (SOL) towards the separatrix. A fast parallel flow was imposed for all hypotheses. The magnitude and the distribution of the 13 C deposition resulting from each hypothesis are compared. A fast parallel flow in the SOL toward the inner divertor combined with a pinch/drift of 10–30 m/s in the inner SOL towards the separatrix roughly reproduces the deposition in both the L- and H-mode experiments.

Erosion of the main chamber plasma-facing components is of concern for ITER. Plasma interaction with the outboard chamber wall is studied in DIII-D using Langmuir probes and optical diagnostics. Fast camera data shows that edge localized modes (ELMs) feature helical filamentary structures propagating towards the outboard wall. Upon reaching the wall, filaments result in regions of local intense plasma–material interaction (PMI) where peak incident particle and heat fluxes are up to two orders of magnitude higher than those between ELMs. In low density/collisionality H-mode discharges, PMI at the outboard wall is almost entirely due to ELMs. A moderate change of the gap between the separatrix and the outer wall strongly affects PMI intensity at the wall. Material samples exposed near the outboard wall showed net carbon deposition in high-density discharges (near the Greenwald limit) and tendency towards net erosion in lower density discharges (∼0.45 of the Greenwald limit).

The dependence of various SOL widths on the line-averaged density ({ovr n}{sub e}) and plasma current (l{sub p}) for the quiescent H-mode plasmas with Type-V ELMs in the National Spherical Torus Experiment (NSTX) was investigated. It is found that the heat flux SOL width ({lambda}{sub q}), measured by the IR camera, is virtually insensitive to {ovr n}{sub e} and has a strong negative dependence on l{sub p}. This insensitivity of {lambda}{sub q} to {ovr n}{sub e} is consistent with the scaling law from JET H-mode plasmas that shows a very weak dependence on the upstream density. The electron temperature, ion saturation current density, electron density, and electron pressure decay lengths ({lambda}{sub Te}, {lambda}{sub jsat}, {lambda}{sub ne}, and {lambda}{sub pe}, respectively) measured by the probe showed that {lambda}{sub Te} and {lambda}{sub jsat} have strong negative dependence on l{sub p}, whereas {lambda}{sub ne} and {lambda}{sub pe} revealed only a little or no dependence. The dependence of {lambda}{sub Te} on l{sub p} is consistent with the scaling law in the literature while {lambda}{sub ne} and {lambda}{sub pe} dependence shows a different trend.

Some key topics in tokamak edge plasma transport and turbulence are reviewed. Multi-device results reveal a new paradigm of scrape-off layer (SOL) transport. Radial transport is driven by intermittency throughout the SOL, in between edge localized modes (ELMs) in H-mode, and comprised of plasma filaments that are generated near the last closed flux surface likely by interchange instability. The filaments travel radially at speeds of ∼1 km/s into the SOL and have a poloidal size of 1–3 cm in most devices. The radial transport in the SOL is poloidally asymmetric, by factors of 2–5, causing a pressure peak in the low field side. This asymmetry and other neo-classical terms, such as Pfirsch–Schluter currents, are found to drive strong SOL flows. The intermittent particle flux, is 20% of the total, including ELMs, at low collisionality, becoming 70% of total at high collisionality. Numerical and analytical models can reproduce the scaling of intermittency with collisionality as well as many details of the filament dynamics in the SOL.

An improved, self-contained gas injection system for the divertor material evaluation system (DIMES) on DIII-D has been employed for in situ study of chemical erosion in the tokamak divertor environment. To minimize perturbation to local plasma, the Mark II porous plug injector (PPI) releases methane through a porous graphite surface at the outer strike point at a rate precisely controlled by a micro-orifice flow restrictor to be approximately equal as that predicted for intrinsic chemical sputtering. Effective photon efficiencies resulting from CH(4) are found to be 58 +/- 12 in an attached divertor (n(e) similar to 1.5 x 10(13)/cm(3), T(e) similar to 25 eV, T(surf)similar to 450 K), and 94 +/- 20 in a semi-detached cold divertor (n(e) similar to 6.0 x 10(13)/cm(3), T(e) similar to 2-3 eV, T(surf) similar to 350 K). These values are significantly more than previous measurements in similar plasma conditions, indicating the importance of the injection rate and local re-erosion for the integrity of this analysis. The contribution of chemical versus physical sputtering to the source of C(+) at the target is assessed through simultaneous measurement of CII line, and CD plus CH-band emissions during release of CH(4) from the Pill, then compared with thatmore » seen in intrinsic sputtering. (C) 2009 Elsevier B.V. All rights reserved.« less

In this chapter we review numerous diagnostics capable of measurements at or near the first wall, many of which contribute information useful for safe operation of a tokamak. Infrared cameras, visible and vacuum ultraviolet cameras, pressure gauges and residual gas analyzers, thermocouples, and erosion and deposition measurements by insertable probes, quartz microbalances, and a rather extensive review of Langmuir probes are discussed. Also discussed are dust measurements by electrostatic detectors, laser scattering, visible and infrared cameras, and manual collection of samples after machine opening. In each case the diagnostic is discussed with a view toward application to a burning plasma machine such as ITER.

Mirrors will be used in ITER in all optical diagnostic systems observing the plasma radiation in the ultraviolet, visible and infrared ranges. Diagnostic mirrors in ITER will suffer from electromagnetic radiation, energetic particles and neutron irradiation. Erosion due to impact of fast neutrals from plasma and deposition of plasma impurities may significantly degrade optical and polarization characteristics of mirrors influencing the overall performance of the respective diagnostics. Therefore, maintaining the best possible performance of mirrors is of the crucial importance for the ITER optical diagnostics. Mirrors in ITER divertor are expected to suffer from deposition of impurities. The dedicated experiment in a tokamak divertor was needed to address this issue. Investigations with molybdenum diagnostic mirrors were made in DIII-D divertor. Mirror samples were exposed at different temperatures in the private flux region to a series of ELMy H-mode discharges with partially detached divertor plasmas. An increase of temperature of mirrors during the exposure generally led to the mitigation of carbon deposition, primarily due to temperature-enhanced chemical erosion of carbon layers by D atoms. Finally, for the mirrors exposed at the temperature of ∼160 °C neither carbon deposition nor degradation of optical properties was detected.

High-pressure noble gas jet injection is a mitigation technique which potentially satisfies the requirements of fast response time and reliability, without degrading subsequent discharges. Previously reported gas jet experiments on DIII-D showed good success at reducing deleterious disruption effects. In this paper, results of recent gas jet disruption mitigation experiments on Alcator C-Mod and DIII-D are reported. Jointly, these experiments have greatly improved the understanding of gas jet dynamics and the processes involved in mitigating disruption effects. In both machines, the sequence of events following gas injection is observed to be quite similar: the jet neutrals stop near the plasma edge, the edge temperature collapses and large MHD modes are quickly destabilized, mixing the hot plasma core with the edge impurity ions and radiating away the plasma thermal energy. High radiated power fractions are achieved, thus reducing the conducted heat loads to the chamber walls and divertor. A significant (2 × or more) reduction in halo current is also observed. Runaway electron generation is small or absent. These similar results in two quite different tokamaks are encouraging for the applicability of this disruption mitigation technique to ITER.

Edge localized mode (ELM) measurements in many tokamaks, including ASDEX-Upgrade, DIII-D, JET, JT-60U and MAST, are reviewed, which includes progress in experimental observations at the plasma edge region by means of fast-time resolved diagnostics with high precision, such as scanning probe, radial interferometer chord, BES and tangentially viewing fast-gated camera at the midplane. ELM dynamics data show that the majority of the ELM particle and energy transport should be dominated by ion convection physics and associated timescales. Furthermore, recent diagnostic upgrades on many tokamaks reveal the ELM filament structure and their complex motion towards radial, poloidal and toroidal directions. Approaches to control the Type-I ELMs, in addition to the alternative scenarios to Type-I ELMy H-mode operation (so-called, small/no ELM regimes) are also a key area of research for current tokamaks, which demonstrated a high confinement (being comparable to that of Type-I ELMy H-mode plasmas at similar parameters) in the absence of large, ELM induced, transient heat/particle fluxes to the divertor targets. Although tolerable ELM regimes are obtained in existing devices, their application to ITER is uncertain. Issues of these regimes towards further experiments and power deposition on divertor targets and main chamber wall are discussed.

Migration of pre-characterized carbon dust in a tokamak environment was studied by introducing about 30 milligrams of dust flakes 5-10 {micro}m in diameter in the lower divertor of DIII-D using the DiMES sample holder. The dust was exposed to high power ELMing Hmode discharges in lower-single-null magnetic configuration with the strike points swept across the divertor floor. When the outer strike point (OSP) passed over the dust holder exposing it to high particle and heat fluxes, part of the dust was injected into the plasma. In about 0.1 sec following the OSP pass over the dust, 1-2% of the total dust carbon content (2-4 x 10{sup 19} carbon atoms, equivalent to a few million dust particles) penetrated the core plasma, raising the core carbon density by a factor of 2-3. When the OSP was inboard of the dust holder, the dust injection continued at a lower rate. Individual dust particles were observed moving at velocities of 10-100 m/s, predominantly in the toroidal direction for deuteron flow to the outer divertor target, consistent with the ion drag force. The observed behavior of the dust is in qualitative agreement with modeling by the 3D DustT code.

The radial particle flux in the scrape-off-layer (SOL) during ELMing H-mode is examined in DIII-D as a function of density. The global radial particle flux in the outboard far SOL is determined by a window frame technique. Between ELMs, the outboard far SOL particle flux increases strongly with density and remains similar to the particle flux across the separatrix as estimated by the pedestal density and temperature gradients. At low density the steep density gradient of the pedestal extends up to 2 cm outside the separatrix, while at high density the density gradient flattens just outside the separatrix. During ELMs the far SOL particle flux becomes localized to the outboard midplane and the assumptions for the window frame analysis break down. Implications for scaling of outboard main chamber wall particle flux and pedestal fueling are explored.

Use of carbon in tokamaks leads to a serious tritium retention issue due to co-deposition. To further investigate the processes involved, a detached ELMy H-mode (6.5 MW NBI) experiment was performed on DIII-D in which {sup 13}CH{sub 4} was puffed into the main vessel through the toroidally symmetric pumping plenum at the top of lower single-null discharges. Subsequently, the {sup 13}C content of tiles taken from the vessel wall was measured. The interpretive OEDGE code was used to model the results. The {sup 13}C deposition pattern could be reproduced, in general shape and magnitude, by assuming in the code the existence of a parallel flow and a radial pinch in the scrape-off layer. Parallel flows of Mach {approx} 0.3 toward the inner divertor and a radial pinch {approx}10 to 20 m/s (+ R-direction) were found to yield {sup 13}C deposition comparable to the experiment.

Gas puff imaging (GPI) experiments are designed to provide high time and space resolution data on the structure of plasma turbulence in the plane perpendicular to the magnetic field. We first examine the temporal behavior of the helium atoms used as the emitting species for GPI and show that for the time scales of interest (≳1 μs), the atomic physics model underlying the conventional interpretation of GPI is valid. Second, we continue the Monte Carlo neutral transport simulations of the GPI diagnostic begun in [D.P. Stotler et al., Contrib. Plasma Phys. 44 (2004) 294]. The radial characteristics of the simulated emission clouds match observations to within the estimated errors. The upshot of these two results is that the technique for unfolding the 2-D, time-dependent plasma density and temperature data from helium GPI emission, relying on this atomic physics model and utilizing the simulated neutral density data from DEGAS 2, is valid.

Large Type-I edge localized modes (ELMs) were completely suppressed by applying edge resonant magnetic perturbations (RMPs) in DIII-D H-mode plasmas with the same low pedestal collisionality (v* e ∼ 0.1) as in ITER. The RMP suppressed ELMs by reducing the edge pressure gradient through a reduction in the pedestal density gradient. Divertor emission profiles showed increases in toroidally localized D ∝ and carbon ion emission and carbon emission throughout both divertor legs. The lowest density ELM-suppressed discharges showed elevated levels of carbon in the core plasma. The data and boundary simulations indicate that the divertor strikepoints make a transition from a high recycling regime during the ELMing phase to a sheath-limited regime during ELM suppression in the lowest density plasmas. ELM suppression during scans of RMP amplitude, injected power and density suggests a possible optimum combination of these actuators for ELM control, without elevated impurity levels, that is extrapolatable to ITER.

The effect of electron cyclotron heating (ECH) on the intrinsic rotation profile in DIII-D is shown experimentally. Former DIII-D experiments have shown that ECH tends to cause an interior reduction in the normally co-Ip directed intrinsic rotation profile, and this core rotation can be fully reversed to the opposite direction. This effect is due to a turbulent rearrangement of the interior rotation profile. Here, we show results that there is more than one mechanism causing this. We compare two low density L-mode discharges where the only operational difference is the location of the ECH deposition. At low ECH power, comparable to the Ohmic power, the primary change is in the q-profile accompanied by a reversal of the core intrinsic rotation direction for the more off-axis deposition. The change in the shear of the q-profile fits well with a recent theoretical prediction for this rotation reversal. At higher ECH power, the primary change is in the core electron temperature, Te, accompanied by a hollowing...

The National Spherical Torus Experiment (NSTX) is currently being upgraded to operate at twice the toroidal field and plasma current (up to 1 T and 2 MA), with a second, more tangentially aimed neutral beam (NB) for current and rotation control, allowing for pulse lengths up to 5 s. Recent NSTX physics analyses have addressed topics that will allow NSTX-Upgrade to achieve the research goals critical to a Fusion Nuclear Science Facility. These include producing stable, 100% non-inductive operation in high-performance plasmas, assessing plasma–material interface (PMI) solutions to handle the high heat loads expected in the next-step devices and exploring the unique spherical torus (ST) parameter regimes to advance predictive capability. Non-inductive operation and current profile control in NSTX-U will be facilitated by co-axial helicity injection (CHI) as well as radio frequency (RF) and NB heating. CHI studies using NIMROD indicate that the reconnection process is consistent with the 2D Sweet–Parker theory. Full-wave AORSA simulations show that RF power losses in the scrape-off layer (SOL) increase significantly for both NSTX and NSTX-U when the launched waves propagate in the SOL. Toroidal Alfven eigenmode avalanches and higher frequency Alfven eigenmodes can affect NB-driven current through energy loss and redistribution of fast ions. The inclusion of rotation and kinetic resonances, which depend on collisionality, is necessary for predicting experimental stability thresholds of fast growing ideal wall and resistive wall modes. Neutral beams and neoclassical toroidal viscosity generated from applied 3D fields can be used as actuators to produce rotation profiles optimized for global stability. DEGAS-2 has been used to study the dependence of gas penetration on SOL temperatures and densities for the MGI system being implemented on the Upgrade for disruption mitigation. PMI studies have focused on the effect of ELMs and 3D fields on plasma detachment and heat flux handling. Simulations indicate that snowflake and impurity seeded radiative divertors are candidates for heat flux mitigation in NSTX-U. Studies of lithium evaporation on graphite surfaces indicate that lithium increases oxygen surface concentrations on graphite, and deuterium–oxygen affinity, which increases deuterium pumping and reduces recycling. In situ and test-stand experiments of lithiated graphite and molybdenum indicate temperature-enhanced sputtering, although that test-stand studies also show the potential for heat flux reduction through lithium vapour shielding. Non-linear gyro kinetic simulations have indicated that ion transport can be enhanced by a shear-flow instability, and that non-local effects are necessary to explain the observed rapid changes in plasma turbulence. Predictive simulations have shown agreement between a microtearing-based reduced transport model and the measured electron temperatures in a microtearing unstable regime. Two Alfven eigenmode-driven fast ion transport models have been developed and successfully benchmarked against NSTX data. Upgrade construction is moving on schedule with initial physics research operation of NSTX-U planned for mid-2015.

In order to reduce recirculating power fraction to acceptable levels, the spherical torus concept relies on the simultaneous achievement of high toroidal β and high bootstrap fraction in steady state. In the last year, as a result of plasma control system improvements, the achievable plasma elongation on NSTX has been raised from κ ∼ 2.1 to κ ∼ 2.6 - approximately a 25% increase. This increase in elongation has led to a substantial increase in the toroidal β for long pulse discharges. The increase in β is associated with an increase in plasma current at nearly fixed poloidal β, which enables higher βtwith nearly constant bootstrap fraction. As a result, for the first time in a spherical torus, a discharge with a plasma current of 1 MA has been sustained for 1 s (0.8 s current flat-top). Data are presented from NSTX correlating the increase in performance with increased plasma shaping capability. In addition to improved shaping, H-modes induced during the current ramp phase of the plasma discharge have been used to reduce flux consumption and to delay the onset of MHD instabilities. Based on these results, a modelled integrated scenario, which has 100% non-inductive current drive with very high toroidal β, will also be discussed. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity (δ ∼ 0.8) at elevated elongation (κ ∼ 2.5). The other main requirement of steady state on NSTX is the ability to drive a fraction of the total plasma current with RF waves. The results of high harmonic fast wave heating and current drive studies as well as electron Bernstein wave emission studies will be presented. © 2006 IAEA, Vienna.

Far Scrape-Off Layer (SOL) and near-wall plasma parameters in DIII-D depend strongly on the discharge parameters and confinement regime. In L-mode discharges cross-field transport increases with the average discharge density and flattens far SOL profiles, thus increasing plasma contact with the low field side (LFS) main chamber wall. In H-mode between edge localized modes (ELMs) the plasma?wall contact is weaker than in L-mode. During ELM fluxes of particles and heat to the LFS wall increase transiently above the L-mode values. Depending on the discharge conditions, ELMs are responsible for 30?90% of the net ion flux to the outboard chamber wall. ELMs in high density discharges feature intermittent transport events similar to those observed in L-mode and attributed to blobs of dense hot plasma formed inside the separatrix and propagating radially outwards. Though the blobs decay with radius, some of them survive long enough to reach the outer wall and possibly cause sputtering. In lower density H-modes, ELMs can feature blobs of pedestal density propagating all the way to the outer wall.

Experiments are described that have increased our understanding of the transport and stability physics that set the H-mode edge pedestal width and height, determine the onset of Type-I edge localized modes (ELMs) and produce the nonlinear dynamics of the ELM perturbation in the pedestal and scrape-off layer (SOL). Models now exist for the ne pedestal profile and the pe height at the onset of Type-I ELMs, and progress has been made towards models of the Te pedestal width and nonlinear ELM evolution. Similarity experiments between DIII-D and JET suggested that neutral penetration physics plays an important role in the relationship between the width and height of the ne pedestal. Plasma physics appears to dominate in setting the Te pedestal width. Measured pedestal conditions including edge current at ELM onset agree with intermediate-n peeling?ballooning (P?B) stability predictions. Midplane ELM dynamics data show the predicted (P?B) structure at the ELM onset, large rapid variations of the SOL parameters and fast radial propagation in later phases, similar to features in nonlinear ELM simulations.

The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order to advance the spherical torus (ST) concept. In order to do this, NSTX utilizes up to 7.5 MW of neutral beam injection, up to 6 MW of high harmonic fast waves (HHFWs), and it operates with plasma currents up to 1.5 MA and elongations of up to 2.6 at a toroidal field up to 0.45 T. New facility, and diagnostic and modelling capabilities developed over the past two years have enabled the NSTX research team to make significant progress towards establishing this physics basis for future ST devices. Improvements in plasma control have led to more routine operation at high elongation and high βT (up to ~40%) lasting for many energy confinement times. βT can be limited by either internal or external modes. The installation of an active error field (EF) correction coil pair has expanded the operating regime at low density and has allowed for initial resonant EF amplification experiments. The determination of the confinement and transport properties of NSTX plasmas has benefitted greatly from the implementation of higher spatial resolution kinetic diagnostics. The parametric variation of confinement is similar to that at conventional aspect ratio but with values enhanced relative to those determined from conventional aspect ratio scalings and with a BT dependence. The transport is highly dependent on details of both the flow and magnetic shear. Core turbulence was measured for the first time in an ST through correlation reflectometry. Non-inductive start-up has been explored using PF-only and transient co-axial helicity injection techniques, resulting in up to 140 kA of toroidal current generated by the latter technique. Calculated bootstrap and beam-driven currents have sustained up to 60% of the flat-top plasma current in NBI discharges. Studies of HHFW absorption have indicated parametric decay of the wave and associated edge thermal ion heating. Energetic particle modes, most notably toroidal Alfven eigenmodes and fishbone-like modes result in fast particle losses, and these instabilities may affect fast ion confinement on devices such as ITER. Finally, a variety of techniques has been developed for fuelling and power and particle control.