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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

We report on recent experiments on DIII-D that examined the effects that variations in the parallel connection length in the scrape-off layer (SOL), L||, and the radial location of the outer divertor target, RTAR, have on divertor plasma properties. Two-point modeling of the SOL plasma predicts that larger values of L|| and RTAR should lower temperature and raise density at the outer divertor target for fixed upstream separatrix density and temperature, i.e., nTAR ∝ [RTAR]2[L||]6/7 and TTAR ∝ [RTAR]−2[L||]−4/7. The dependence of nTAR and TTAR on L|| was consistent with our data, but the dependence of nTAR and TTAR on RTAR was not. The surprising result that the divertor plasma parameters did not depend on RTAR in the predicted way may be due to convected heat flux, driven by escaping neutrals, in the more open configuration of the larger RTAR cases. Modeling results using the SOLPS code support this postulate.

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.

The effect of 2D transport in the tokamak scrape-off layer (SOL) on relating measured divertor heat-flux profiles to midplane plasma profiles is explored with the UEDGE code for a range of transport assumptions. The relationship between the divertor heat-flux profile and the midplane plasma temperature profile as determined from UEDGE is compared to commonly used predictions from 1D and “2-point” analytic relation. Results show that the parametric variation in λ q , div with midplane separatrix values follows the conduction-limited 2-point model, though the actual value of q // on a given flux surface is much lower than implied by the midplane T e , sep , and λ q , div is ∼ twice as wide as expected from these models. Poloidal variations in χ result in minimal change to radial profiles and probably can’t be distinguished experimentally from cases with uniform χ having the same flux-surface average value.

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.

Data from a recently installed insertable magnetic probe array in the Sustained Spheromak Physics Experiment (SSPX) [E. B. Hooper et al., Nucl. Fusion 39, 863 (1999)] is compared against NIMROD [C. R. Sovinec et al., J. Comp. Phys. 195, 355 (2004)], a full 3D resistive magnetohydrodynamic code that is used to simulate SSPX plasmas. The experiment probe consists of a linear array of chip inductors arranged in clusters that are spaced every 2 cm, and spans the entire machine radius at the flux conserver midplane. Both the experiment and the numerical simulations show the appearance, shortly after breakdown, of a column with a hollow current profile that precedes magnetic reconnection, a process essential to the formation of closed magnetic flux surfaces. However, there are differences between the experiment and the simulation in how the column evolves after it is formed. These differences are studied to help identify the mechanisms that eventually lead to closed-flux surfaces (azimuthally averaged) and flux amplification, which occur in both the experiment and the simulation.

The effects of magnetic perturbation on the quality of confinement in the sustained spheromak plasma experiment (SSPX) (Hooper E B et al 1999 Nucl. Fusion 39 863) are studied. The magnetic field is taken as a sum of an unperturbed axisymmetric magnetic field, B 0 , calculated from a magneto-hydrodynamic equilibrium, and a non-axisymmetric part, δB, representing the effects of kinking perturbation of the central plasma column. Poincare maps show that flux surfaces are strongly affected by perturbations with δB/B 0 ? 5%, with islands and stochastic regions covering a substantial part of plasma volume. The effective heat transport in the perturbed magnetic field, calculated by a Monte-Carlo procedure, is found to be consistent with the experimentally measured level of confinement.

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)

In October 2003, Dr. Raymond Orbach, Director of the Department of Energy’s Office of Science, issued a charge to the Fusion Energy Sciences Advisory Committee (FESAC) “to identify the major science and technology issues that need to be addressed, recommend how to organize campaigns to address these issues, and recommend the priority order for these campaigns.” The sections in this report document the results of the Panel’s work. The first two sections describe the concepts of the overarching themes, topical scientific questions, and campaigns. The next six sections (Sections 3–8) describe in detail the six scientific campaigns. Section 9 describes some important enabling research activities necessary for the campaigns. Sections 10–12 describe the overarching themes, which provide a crosscutting perspective of the activities in the six campaigns. Finally, the Panel’s recommendations are set forth in Section 13. The charge letter to the panel is provided as Appendix A; the FESAC response letter is provided as Appendix D.

In this paper we report on the results of deuterium and hydrogen experiments in the Sustained Spheromak Physics Experiment (SSPX). We have compared ∼500 deuterium discharges with similar discharges in hydrogen. Typically, we produce H 2 plasmas with peak toroidal currents in the range of 0.6 MA, electron temperatures ( T e ) of ∼200 eV and energy confinement times ( τ E ) of ∼200 μs. The D 2 fueled discharges show similar results to those with H 2 fueling with no obvious differences in confinement time. Electron temperatures of ∼200 eV with similar electron densities were observed. Both the deuterium and hydrogen fueled discharges have a calculated thermal diffusivity below χ e 2 /s. However, the D 2 fueled discharges had a factor of ∼5 increase in high- Z (titanium) impurity content suggesting an increase of physical sputtering. We find no significant mass scaling effects.

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.

The current density on the open field lines of the Livermore spheromak (SSPX) typically exceeds the saturation current density of the bulk plasma. We assume that the mechanism that provides conditions for that is associated with the formation of a thin layer near the cathode surface, where both the plasma and the neutral density are higher than in the bulk plasma and where intense ionization occurs. The ions formed in this layer fall back onto the cathode, whereas electrons contribute to the high current density in the bulk plasma. The particle balance in the ionizing layer is determined by the recycling coefficient, which, in turn, depends on the cathode temperature and the sheath voltage. As it turns out, these dependences give rise to an instability that leads to the current filamentation and the formation of hot spots on the cathode surface. The instability can be characterized in a phenomenological manner without going into the details of the structure of the ionizing layer, whose effect on the instability shows up in the form of a couple of numerical coefficients of the order of one. We predict the characteristic size and the shape of the filaments (and the hot spots), which are in a general agreement with discoloration patterns on the surface of the cathode in the SSPX. If the magnetic field is tilted to the surface, the footpoints of the filaments move with a significant velocity, whose direction depends on the ratio of the ion gyroradius and the thickness of the ionizing layer. This instability, although primarily considered in conjunction with the SSPX experiment, may play a role in spherical tokamaks and other systems with coaxial helicity injection.

The magnetic helicity balance between the helicity input injected by a magnetized coaxial gun, the rate-of-change in plasma helicity content, and helicity dissipation in electrode sheaths and Ohmic losses have been examined in the Sustained Spheromak Plasma Experiment (SSPX) [E. B. Hooper, L. D. Pearlstein, and R. H. Bulmer, Nucl. Fusion 39, 863 (1999)]. Helicity is treated as a flux function in the mean-field approximation, allowing separation of helicity drive and losses between closed and open field volumes. For nearly sustained spheromak plasmas with low fluctuations, helicity balance analysis implies a decreasing transport of helicity from the gun input into the spheromak core at higher spheromak electron temperature. Long pulse discharges with continuously increasing helicity and larger fluctuations show higher helicity coupling from the edge to the spheromak core. The magnitude of the sheath voltage drop, inferred from cathode heating and a current threshold dependence of the gun voltage, shows that sheath losses are important and reduce the helicity injection efficiency in SSPX.

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.

We are in the process of testing a technique for measuring the magnetic-field line topology in magnetically confined plasmas. The basic idea is to use a high-powerful short-pulse laser to launch a burst of energetic (∼100 keV) electrons from a target passing through the plasma of interest; these electrons then generally follow field lines until they strike a solid surface, where a burst of x rays is produced and then detected. The field line connection length can be determined from the time delay between the laser pulse and the burst of x rays. The topology of the field lines can be inferred by measuring the connection length as a function of initial target location inside the plasma. Measuring the spatial distribution of the x-ray production will provide further information on the field topology, including the effects of magnetic-field fluctuations and stochasticity. The work will eventually include testing the appropriate x-ray detectors, measuring the background x-ray emission in a spheromak plasma, me...

New types of toroidally rotating fluctuations (toroidal mode numbers n = 1 and n = 2) of line-integrated electron density and Hα emission, with frequencies ranging from 10 to 100 kHz, are observed in the sustained spheromak physics experiment (SSPX). The rotating directions of these fluctuations are the same as the direction determined by E×B, while the E and B directions are determined by the gun voltage and gun magnetic flux polarities, respectively. These results take advantage of one distinctive signature of spheromaks, i.e. it is possible to observe toroidal MHD activity during decay and sustainment at any toroidal angle. A theoretical constraint on line-integrated measurement is proposed and is found to be consistent with experimental observations. Fluctuation analysis in the time and frequency domains indicates that the observed density and Hα fluctuations correlate with magnetic modes. Observation of Hα fluctuations correlating with magnetic fluctuations indicates that, at least in some cases, MHD n = 1 modes are due to the so-called `dough-hook' current paths that connect the coaxial gun to the flux conserver, rather than internal kink instabilities. These results also show that electron density and Hα emission diagnostics complement other tools for spheromak mode study.

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.

A newly installed density diagnostic using CO2 laser interferometry and an Hα diagnostic using interference filters and photomultiplier tubes for the recently constructed sustained spheromak physics experiment (SSPX) are described. First diagnostic results of the Hα diagnostic were useful to understand the breakdown physics in the new SSPX experiments. Low-noise density data validates techniques to reduce vibration and electronic pickup. The data-processing electronics of the new interferometer can yield unambiguous density data that is equivalent to 16 fringe shifts. Density data is also critical to understand the particle source, and the J/ne parameter for SSPX.

In this article we discuss the measurement of the field profile in the sustained spheromak physics experiment (SSPX). We have built a transient internal probe (TIP) diagnostic to measure the internal field profile in a SSPX plasma sustained by dc coaxial helicity injection. TIP is a diagnostic that makes a spatially resolved (i.e., not chord averaged) measurement of the local magnetic field using Faraday rotation. A 1-cm by 4-mm-diameter verdet probe is fired through the plasma at about 2 km/s by a two-stage light gas gun. The probe is illuminated by an argon ion laser throughout the traverse of the plasma—the retro-reflected light is then analyzed with an ellipsometer to determine the field at each location. The speed, small size of the probe, and the probe cladding make this measurement possible even in hot plasmas (100 s of eV). The measurement is accurate enough (1 MHz, ±7 G, 1-cm spatial resolution) to map out magneto hydrodynamic (MHD) mode amplitudes from the edge to the magnetic axis.

In this article we present an overview of the plasma diagnostics operating or planned for the sustained spheromak physics experiment device now operating at Lawrence Livermore National Laboratory. A set of 46 wall-mounted magnetic probes provide the essential data necessary for magnetic reconstruction of the Taylor relaxed state. Rogowski coils measure currents induced in the flux conserver. A CO2 laser interferometer is used to measure electron line density. Spectroscopic measurements include an absolutely-calibrated spectrometer recording extended domain spectrometer for obtaining time-integrated visible ultraviolet spectra and two time-resolved vacuum monochrometers for studying the time evolution of two separate emission lines. Another time-integrated spectrometer records spectra in the visible range. Filtered silicon photodiode bolometers provide total power measurements, and a 16 channel photodiode spatial array gives radial emission profiles. Two-dimensional imaging of the plasma and helicity injec...

The phenomena that lead to the observed power and particle flux detachment have been studied with the code B2-Eirene for DIII-D Helium and Deuterium experiments. Contrary to the usual Deuterium experiments, a significant reduction of the power load to the divertor is observed in Helium discharges while the ion flux remains high. Modelling indicates that this is due to the longer ionisation mean free path of Helium, which can penetrate from the divertor into the bulk plasma with the consequent power loss. At this stage of Helium detachment, most of the outer divertor plasma remains at an electron temperature of ∼5-10 eV, for which Helium recombination does not occur.

Recent progress towards an increased understanding of the physical processes in the divertor and scrape-off layer (SOL) plasmas in DIII-D has been made possible by a combination of new diagnostics, improved computational models and changes in divertor geometry. The work focused primarily on ELMing H mode discharges. The physics of partially detached divertor plasmas, with divertor heat flux reduction by divertor radiation enhancement using D2 puffing, was studied in two dimensions, and a model of the heat and particle transport was developed that includes conduction, convection, ionization, recombination and flows. Plasma and impurity particle flows were measured with Mach probes and spectroscopy and compared with the UEDGE model. The model now includes self-consistent calculations of carbon impurities. Impurity radiation was increased in the divertor and SOL with `puff and pump' techniques using SOL D2 puffing, divertor cryopumping and argon puffing. The important physical processes in plasma-wall interactions were examined with a DiMES (divertor material evaluation system) probe, plasma characterization near the divertor plate and the REDEP code. Experiments comparing single null plasma operation in baffled and open divertors demonstrated a change in the edge plasma profiles. These results are consistent with a reduction in the core ionization source calculated with UEDGE. Divertor particle control in ELMing H mode with pumping and baffling resulted in a reduction in H mode core densities to ne/nGr ≈ 0.25 (with nGr the Greenwald density). Divertor particle exhaust and heat flux were studied as the plasma shape was varied from a lower single null to a balanced double null, and finally to an upper single null.

We report new experimental results with the Radiative Divertor Project-outer baCe (RDP-OB) and cryopump in both upper single-null (USN) and double-null (DN) ELMing H-mode discharges. The baCed divertor reduced the core ionization (2‐2.5·), in reasonable agreement with predictions from UEDGE/DEGAS modeling (3.75·). The upper cryopump achieved density control of ne/ngw 0.22 (line density/Greenwald density) with Zeff 2 in high-d plasmas. The measured exhaust is comparable to the lower pump, except at lower core electron densities (ne < 5 · 10 19 m ˇ3 ). EAcient impurity exhaust was obtained with deuterium SOL flow. Preliminary experiments with DN operation has shown that the particle exhaust to the upper pump depends on the up/down magnetic balance. Preliminary experiments indicate that the DN exhaust is roughly 40‐50% of the USN exhaust at ne 4 · 10 19 m ˇ3 . ” 1999 Elsevier Science B.V. All rights reserved.

The ITER divertor power deposition database is described and analysed in this paper. The database contains experimental measurements from the major divertor experiments in L-mode and H-mode regimes. These measurements are used to derive multi-machine scaling laws for the peak divertor heat flux and width, particularly of their machine size dependence. The physical basis for these scalings is discussed and the laws obtained are used to extrapolate from existing experiments to the parameters expected in the ITER-EDA device.

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 ).

This paper presents the detailed evolution of conditions along both the inner and outer divertor legs during the transition from attached ELMing H-mode to partially detached divertor (PDD) operation in DIII-D. Visible emission profiles in a poloidal plane show that in ELMing H-mode prior to deuterium gas injection, CIII emission peaks in the inner SOL near the X-point and deuterium emission (from ionization and recombination) peaks at the inner target plate near the inner strike point (ISP). The spatial profiles of the recombination and ionization zones, determined by forming images of the ratio of intensities from simultaneous images of D α and D γ emission, show that recombination dominates the inner leg emission near the target; ionization dominates in a poloidally narrow zone upstream in the inner leg. After deuterium injection, when the PDD transition begins, the profiles of carbon visible emission show first an increase in the inner SOL near the X-point, followed by increases in emission in the lower regions of the outer leg. Deuterium emission at the transition onset decreases at the ISP and increases across the private flux region below the X-point. As the transition to PDD conditions proceeds the deuterium emission increases in the private flux region; recombination dominates near the floor and ionization higher near the X-point. Carbon emission appears along both divertor legs and at the X-point. In the final quasi-steady PDD state, the recombination emission in the outer leg is near the separatrix and along the target plate; emission from collisional excitation dominates in the upper part of the outer leg just below the X-point, and carbon emission is localized at the X-point. These results suggest that transport of neutral deuterium between the inner and outer divertor legs through the private flux region plays an important role in the initiation of outer leg detachment in DIII-D.

This paper summarizes some of the main conclusions and highlights of the Emerging Concepts Working Group (EC Group) of the Snowmass 1999 Fusion Summer Study. The EC Group considered technical challenges and opportunities related to emerging concept issues.

This paper summarizes results from a two-dimensional (2D) physics analysis of the transition to and stable operation of the partially detached divertor (PDD) regime induced by deuterium injection in DIII-D. The analysis [1] shows that PDD operation is characterized by a radiation zone near the X-point at -15 eV which reduces the energy flux into the divertor and thereby also reduces the target plate heat flux, an ionization zone below the X-point which provides a deuterium ion source to fuel parallel flow down the outer divertor leg, an ion-neutral interaction zone in the outer leg which removes momentum and energy from the flow and finally a volume recombination zone above the target plate which reduces the particle flux to the low levels measured on the plates and thereby also contributes to reduction in target plate heat flux.

This is the July 1996 report of a subpanel of the US Department of Energy Fusion Energy Sciences Advisory Committee (FESAC), charged with reviewing the present status of fusion alternative concept development and the prospects for alternative concepts not only as fusion power systems but also the scientific contributions of alternative concept research to the fusion energy sciences program and to plasma science in general.

A series of controlled experiments has been carried out in DIII-D to induce a bulk ion flow in the SOL and evaluate its effect on the localization of impurities in the divertor. This induced SOL flow was created by simultaneous deuterium puffing and divertor exhaust using a divertor cryopump, and the impurity enrichment was measured directly. The experiments were designed to compare enrichment in discharges with and without induced flow having otherwise similiar divertor parameters. Significant increases in impurity compression and enrichment are observed when flow is induced, and the degree of impurity enrichment in the divertor is found to be dependent on the impurity of interest. Detailed particle measurements made possible by the direct measurement of impurity densities in several reservoirs indicate reasonable particle balance for helium throughout the duration of the discharge. Conversely, while the total input of neon is balanced by the total exhaust by the end of a discharge, particle balance is not observed during the course of the discharge. A significant wall inventory with a short release time (~10 ms) is surmised.