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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.

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.

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

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.

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.

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...

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.

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 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 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.

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.

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.

A series of observations is presented concerning divertor heat flux, qdiv, in the DIII-D tokamak, and it is shown that many features can be accounted for by assuming that the heat flux flows preferentially along field lines because τ|| < τ⊥ in the scrape-off layer (SOL). Exceptions to this agreement are pointed out and the discrepancies explained by means of two dimensional (2-D) effects. About 80% of the discharge input power can be accounted for. The power deposited on the target plate due to enhanced losses during edge localized modes (ELMs) is less than 10% of the total target power in most cases. X point height scans for lower single null (LSN) diverted discharges show that the peak heat flux variation is primarily due to flux expansion and secondarily due to transport of energy across the magnetic field in the divertor. At the outer strike point qdiv,peak Pin(Ip - Ip,0)G(gin)(1/Bt)4/9(Bdiv/Bmp)f(Ldivχ⊥), where G is a linear function of the inner gap, gin, over a specified range and f describes cross-field energy transport in the divertor. Evidence of radial in-out asymmetries (comparing the outer strike point with the inner strike point or centre-post) and toroidal asymmetries in qdiv is presented and the heat flux peaking due to tile gaps and misalignment of tiles is examined. For magnetically balanced double null (DN) discharges with downward ∇B ion drift, it is found that qdiv is inherently higher in the lower divertor than in the upper divertor, having a 3:1 downward bias. Examples of heat flux reduction by gas puffing deuterium or neon in LSN and DN discharges are given. At least a threefold reduction of the peak heat flux in both the upper and lower divertors of a DN discharge, using D2 puffing, is reported.

In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma separates the first wall from the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma. Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma - wall interaction. In this paper we will describe a set of experiments on DIII-D designed to provide detailed two-dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. We will also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modelling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process.

Recent measurements of the two-dimensional (2-D) spatial profiles of divertor plasma density, temperature, and emissivity in the DIII-D tokamak [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] under highly radiating conditions are presented. Data are obtained using a divertor Thomson scattering system and other diagnostics optimized for measuring the high electron densities and low temperatures in these detached divertor plasmas (ne⩽1021 m−3, 0.5 eV⩽Te). D2 gas injection in the divertor increases the plasma radiation and lowers Te to less than 2 eV in most of the divertor volume. Modeling shows that this temperature is low enough to allow ion–neutral collisions, charge exchange, and volume recombination to play significant roles in reducing the plasma pressure along the magnetic separatrix by a factor of 3–5, consistent with the measurements. Absolutely calibrated vacuum ultravi...

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

Thomson scattering measurements of ne and Te in the divertor region of a Tokamak are reported. These data are used as input to boundary physics codes such as UEDGE and DEGAS and to benchmark the predictive capabilities of these codes. These measurements have also contributed to the characterization of tokamak disruptions. A Nd:YAG laser (20 Hz, 1 J, 15 ns, 1064 nm) is directed vertically through the lower divertor region of the DIII–D Tokamak. A custom, aspherical collection lens (f/6.8) images the laser beam from 1 to 21 cm above the target plates into eight spatial channels with 1.5 cm vertical and 0.3 cm radial resolution. Two-dimensional mapping of the divertor region is achieved by sweeping the divertor X-point location radially through the fixed laser beam location. Fiber optics carry the light to polychromators whose interference filters have been optimized for low Te measurements. Silicon avalanche photodiodes measure both the scattered and plasma background light. Temperatures and densities are typically in the range of 5–200 eV and 1–10×1019 m−3, respectively. Low temperatures, Te 8×1020 m−3 have been observed in detached plasmas. Background light levels have not been a significant problem. Reduction of the laser stray light permits Rayleigh calibration. Because of access difficulties, no in-vessel vacuum alignment target could be used. Instead, an in situ laser alignment monitor provides alignment information for each laser pulse. Results are compared with Langmuir probe measurements where good agreement is found except for regions of high ne and low Te as measured by Thomson scattering.

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

This paper describes divertor density and temperature measurements using both a new reciprocating Langmuir probe (XPT-RCP) which plunges vertically above the divertor floor up to the X-point height and swept, single, Langmuir probes fixed horizontally across the divertor floor. These types of measurements are important for testing models of the SOL and divertor which then are used to determine engineering design criteria for plasma facing components in reactor size tokamaks. The 6 mm diameter fixed single probes (19 domed and 2 flat at radially equivalent locations) are incorporated into the lower divertor floor at 19 radial locations and swept at 250 Hz. These probes are critical for determining plasma detachment from the floor during operation with high density, highly radiating divertors. By sweeping the divertor strike point across the fixed probes, different regions of the target plate incident flux profile can be sampled and a high resolution spatial profile can be obtained from each probe tip as the strike point moves past. The X-point reciprocating probe (XPT-RCP) provides ne and Te profiles with high spatial (2 mm) and temporal (0.5 ms) resolution from the target plate to the X-point along a single vertical chord at the same radial location as a fixed probe tip at a different azimuthal location. The probe ne and Te are compared to the divertor Thomson scattering (DTS) ne and Te (eight vertical points at 20 Hz, RThomson = RX-point-rcp). Recent observations have also shown divertor densities from 3 × 1019 to 4 × 1020 m−3 near the target plate with the highest densities observed with D2 gas puffing. Electron temperature is typically of the order of 15–25 eV at the target rising to about 70 eV near the X-point. Lower temperature, higher density plasmas are observed along the inner leg. Generally good agreement among the XPT-RCP, the fixed floor probes, and the DTS is observed. Differences between these diagnostic measurements will also be discussed with respect to different operating regimes and conditions.

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

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

The physics capabilities of an ignition-and-moderate-burn tokamak to explore the physics of burning plasmas and burn control on ash accumulation time scales are described. The machine provides physics capabilities comparable to those of the International Thermonuclear Experimental Reactor (ITER) for pulse lengths up to 120 s, but lacks the nuclear component testing, superconducting magnet technology, and long-pulse aspects of ITER`s mission. Strong plasma shaping is adopted to reduce the cost relative of ITER. Using ITER guidelines to evaluate the physics performance, this machine has the same ignition margin as ITER`s, and operates within the limits on beta, density (i.e., the Greenwald density limit), and safety-factor specified in the ITER physics guidelines. Acceptable peak heat fluxes to divertor target surfaces are maintained with an attached, high-recycling divertor operating scenario typical of present-day machines. A range of ignited and driven operating modes is available, including advanced modes prototypical of steady-state tokamak operation. 10 refs., 3 figs., 3 tabs.

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

Sufficient neutral atom and molecular throughput is essential for the steady-state operation of the proposed Tokamak Physics Experiment tokamak. To predict the throughput, the B2 edge-plasma fluid code and the DEGAS Monte Carlo neutral transport code were coupled globally. For the day 1 low-power (17.5-MW) operation condition, the recycling coefficient for both codes matched at 0.985, implying that for every 1000 ions striking the divertor plate, 15 are ultimately removed down the pump duct. The neutral molecular density was 2.52{+-}0.15x10{sup 19}m{sup 3}, giving a throughput of 92.6{+-}5.6 Torr.l/s. Varying the scrape-off length for the plasma extending into the gap between the baffle and the plate from 0 to 2 cm decreased the throughput by a factor of >2. Moving the strike point away from the gap at first increases the throughput by lessening the pumping efficiency of the plasma in the gap. As the plasma is moved even farther away, the throughput drops due to a lack of source term for neutrals entering the pumped region. Illustrating the importance of moving the source term, moving the strike point away from the gap but retaining the original plasma in the gap lowers the throughput by a factor of 10. Altering themore » curvature of the baffle has little effect on the neutral solution. 7 refs., 9 figs.« less

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

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

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

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

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

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

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

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

A 48‐channel platinum‐foil bolometer system on DIII‐D was installed to achieve better spatial and temporal resolution of the radiated power in diverted discharges. Two 24‐channel arrays provide complete plasma coverage with optimized views of the divertor. The divertor radiation profile was measured for a series of radiative divertor and power balance experiments. A significant change in the magnitude and distribution of divertor radiation with heavy gas puffing was observed. Unfolding the radiation profile with only two views requires one to treat the core and divertor radiation separately. The core radiation is fitted to a function of magnetic flux and is then subtracted from the divertor viewing chords. The divertor profile is then fit to a 2D spline as a function of magnetic flux and distance from divertor floor.

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

Experimental and theoretical results from around the world point to the possibility of high confinement, high- beta , and high-bootstrap-fraction steady-state tokamak operating modes. These modes of operation, if fully developed and extended to steady-state, could lead to much less expensive tokamak demonstration power reactors and to a significantly reduced cost-of-electricity from fusion, as compared to projections based on low- beta N, pulsed operating modes. Present results have clear implications in the areas of particle control, plasma shaping, and current-profile control. Thus they have strongly influenced the design of the steady-state advanced tokamak TPX, which has the mission to combine the best results from present experiments and extend them to steady-state. These results also have important implications for follow-up tests in ITER, which have the goal of studying advanced-tokamak operation in an ignited plasma, as well as for the eventual configuration of an advanced-tokamak fusion reactor.

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

The open divertor in DIII-D has been modified by the installation of a baffled plenum with a narrow entrance aperture. Neutral particle pressures in the range of 2-20 mtorr have been measured in this new plenum during experiments in L and H mode plasmas in which the neutral beam power was varied and the divertor target location was scanned with respect to the entrance aperture. These pressures exceed the predicted minimum required to provide adequate particle exhaust for controlling the density in H mode discharges when pumping experiments are conducted in the future. The pressure measurements presented here, which were carried out in the absence of a pump, were modelled with the aid of two transport codes: an edge plasma code and a neutrals code. It was found in this modelling process that the divertor flux amplification factor required to explain the measured pressures is of the order of 30 to 40, much higher than the factors used in the modelling of data in the absence of the baffle. This higher flux amplification factor results from increased recycling at the divertor owing to the presence of the baffle

Reduction of the divertor heat load is an important issue for future tokamaks, particularly during the technology phase of ITER. We discuss a conceptual design for one type of advanced divertor: the radiative slot divertor. The goal of this divertor configuration is to enhance the radiation in the divertor region and thereby reduce the heat load at the strike points. At the same time, any effects on the core plasma must be minimized. Proof-of-principle experiments to enhance the radiation in the DIII-D divertor have been performed both with deuterium and impurity injection. We compare several computer models with results from these experiments to predict performance and thereby guide designs of radiative divertors for future machines. We have estimated impurity radiation using calculations of the background plasma with a two-dimensional fluid code (B2 or LEDGE) coupled with models of impurity radiation. The DEGAS code has been used to estimate hydrogenic transport, charge exchange and radiation losses. Estimates of impurity transport are provided by 11/1-dimensional models and calculations of impurity frictional-force terms. These model, results are in qualitative agreement with the ∼1 MW reduction of measured divertor power in DIII-D during divertor impurity puffing experiments. Specific designs, including engineering details, for applications to DIII-D and ITER will be discussed.

We show that charge-exchange and other ion-neutral collisions can reduce the power transmission factor of the plasma sheath, thereby lowering the ion impact energy and target plate sputtering. The power transmission factor relates the heat flux reaching the divertor target to the plasma density and temperature just in front of the surface: δ=Qsurf/Jewk Te. Experimental data from the DIII-D tokamak suggests that δ could be as low as 2–3 near the region of peak divertor particle flux, instead of the 7–8 expected from usual sheath theory. Several effects combine to allow ion-neutral interactions to be important in the divertor plasma sheath. The shallow angle of incidence of the magnetic field (1–3° in DIII-D) leads to the spatial extension of the sheath from approximately ρi∼1 mm normal to the plate [R. Chodura, Phys. Fluids 25 (1982) 1628] to several centimeters along the field lines. Ionization reduces the sheath potential, and collisions reduce the ion impact energy.

We use a new multispecies 1D fluid code, NEWT-ID, to model DIII-D scrape-off layer (SOL) behavior during radiative divertor and high heat flux experiments. The separatrix location and the width of the SOL are uncertain, and affect the comparison of the data in important ways. The model agrees with many of the experimental measurements for a particular prescription for the separatrix location. The model cannot explain the recent data on the separatrix T i with a conventional picture of ion and electron power flows across the separatrix. Radial transport of particles and heat in some form is required to explain the peak heat flux data before and after gas puffing. For argon puffing in the private flux region, entrainment is poor in the steady state. The calculations suggest that strike point argon puffing in a slot divertor geometry results in substantially better entrainment. Self-consistent, steady-state solutions with radiated powers up to 80% of the SOL power input are obtained in 1D. We discuss significant radial effects which warrant the development of a code which can treat strongly radiating impurities in 2D geometries.

An array of fixed Langmuir probes spanning the inner and outer divertor strike points has been used to measure the electron temperature, plasma density, and particle/heat flux profiles of the divertor plasma in DIII-D. We report here on observations made during the steady state phase of high-power neutral-beam heated H-mode plasmas. These data are important for predicting the expected impurity sources and target plate material lifetimes in future high power tokamaks such as ITER. H-mode plasmas formed with a high divertor neutral pressure in the absence of external gas feed in DIII-D were found to have beneficial consequences in reducing the peak values of divertor electron temperature without loss of core energy confinement. We have observed a steady state, high density ( n pl,d ≈10 14 cm −3 ), moderate temperature ( 2 . A better understanding of the importance of collisional effects may be needed to interpret these results.