Session 7R - C-Mod and TdeV.
POSTER session, Thursday morning, November 14
Exhibit Hall - Concourse Level, Adam's Mark

Operational improvements implemented during the past year have increased the reliability of high power RF heating of L-mode, H-mode, and PEP mode plasmas. Extremely high power densities of P/V \le 5\rm\,MW/m^3 and P/S \le 0.6\rm\,MW/m^2 were achieved with up to 3.5\,MW of RF power at 80\,MHz. Heating efficiencies of low single-pass absorption scenarios, including D(^3He) minority heating at 8\,T, have improved substantially by reducing the high-Z impurity influx using boronization. H-mode confinement has also improved substantially after boronization. High power heating of H-mode plasmas has produced a quasi-steady-state plasma with H_ITER89-P=2.0 and \beta_N=1.5 while maintaining P_rad^\rm\,main/P_in\le 0.3. The combination of increased input power, reduced radiation from the main plasma, and short scrape-off length in H-mode has produced ITER-relevant parallel heat fluxes into the divertor region (q_\parallel \ge 0.5\rm\,GW/m^2). Off-axis electron heating by the mode-converted ion Bernstein wave was demonstrated in D-^3He plasmas at 8\,T. This mode of operation will provide the current profile control capability necessary for future advanced tokamak experiments. Current drive studies are scheduled to start in 1997 using the 40\,MHz RF power currently being installed in collaboration with PPPL. \$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.02] Confinement of H-mode Plasmas in C-Mod

A series of experiments, examining the confinement properties of ICRF heated H-mode plasmas, has been carried out on the C-Mod tokamak. For these plasmas, the plasma is essentially thermal with very little contribution to the stored energy from energetic ions and with T_i \simeq T_e. The data include those taken both before and after the molybdenum first wall surfaces were coated with boron. H-modes obtained with boronized walls typically had lower impurity content and radiated power and attained higher stored energy than those on bare molybdenum. Confinement enhancement relative to ITER89P for discharges with boronized walls, ranged from 1.6 to 2.4. The unique operating regime of the C-Mod device provided a means for extending the tests of global scaling laws to parameter ranges not previously accessible. For example, the C-Mod ELMfree data was found to be 1.4 times the ITER94 scaling and the ELMy data almost 2 times the ITER92 ELMy scaling law, suggesting that the size scaling in both may be too strong. A clear linear relationship between the edge temperature and the temperature gradient in the core plasma was observed; the discharges with the ``best'' transport barriers also showing the greatest improvement in core confinement. A summary of the data described in this paper are now available in the ITER H-mode database. \$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.03] Negative Central Shear Modes of Operation Near the \beta-limit in Alcator C-Mod

The ideal MHD stability properties of equilibria with negative central shear and high bootstrap current fraction (f_bs \simeq 0.75), that may be achieved in Alcator C-Mod have been examined. Without a conducting shell and for relatively broad pressure profiles (p(0)/p_avg \simeq 3.0), values of normalized beta [\beta_N=\beta_t/(I_p/aB)] up to 3.7 were found for highly shaped plasmas with \kappa_x \simeq 1.8,\ \delta_x \simeq 0.7,\ q_min \simeq 2.2,\ and\ r_min/a \simeq 0.75. For elliptical and circular plasmas with low triangularity (\delta_x < 0.1), the stability limits are significantly lower (\beta_N \simeq 2.2). The \beta-limits for more peaked pressure profiles (p(0)/p_avg \simeq 4.5) were also found to be lower than for the broader pressure profiles. In all these cases the stability limit is determined by the ideal n=1, external kink mode. For the highly shaped plasma cross-section with a broad pressure profile, a maximum \beta_N \simeq 5.5 was found with a conducting shell placed at r_w=1.3 a. In this case the \beta-limit is set by the n=3, external kink mode. In these studies self-consistent current density profiles and MHD equilibria were computed using the ACCOME code and the stability analysis was carried out using the JSOLVER and PEST-II codes. Details of the stability analyses and the current profile control studies will be discussed. \$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.04] Reversed Shear Mode in Alcator C-Mod with Early ICRF Heating, Current Ramp and Pellet Injection

Initial scoping experiments were performed to obtain the enhanced reversed shear (ERS) mode in Alcator C-Mod using current ramp and early ICRF heating in the D(H) minority heating regime. Typical parameters at the end of a 0.23\,sec sawtooth free period were øverlinen_e=1\times10^20\rm\,m^-3, B_T=5.3\rm\,T, P_rf=2\rm\,MW, T_e=5\rm\,keV, T_i=2.5\rm\,keV, and I_p=0.8\rm\,MA. The minority concentration was low, typically 2% and therefore we expect mostly electron heating. While good ICRF coupling and excellent heating were observed, there was no evidence of the ERS mode, presumably owing to insufficient power being coupled to ions. In the coming experimental campaign we plan to (a) increase the minority concentration to couple a greater fraction of the power into ions; (b) slow down the current ramp rate to allow more time for the formation of the RS mode; (c) use pellet injection during the initial sawtooth free period with the aim of combining the PEP mode and the RS mode, with the goal of achieving the ERS mode. We have undertaken a modeling effort using TRANSP, FELICE, and ACCOME to predict the optimal operating regimes.

\$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.05] Studies of ICRF Heating Experiments on Alcator C-Mod Using ECE Temperature Measurements

On Alcator C-Mod, the main heating scenarios are minority ion heating in D(H) and D(He^3) plasmas. The power from the launched fast magnetosonic wave (80 MHz) is absorbed by minority ions at their cyclotron resonance. The energetic minority ion population heats the bulk plasma by slowing down on electrons and majority ions. Another heating regime in which the magnetosonic wave is mode converted to an ion Bernstein wave (IBW) has been explored in both H-He^3 and D-He^3 plasmas. Because the IBW damps strongly on electrons, this regime can provide localized direct electron heating. Both on axis and off axis heating have been observed, with power profile widths (FWHM) of \sim 0.2 a . The heating location is observed to follow that of the ion hybrid resonance which is controlled by both magnetic field and ion species ratios. Grating polychromator measurements of electron temperature are used to study the behavior of the heating in both regimes. A minority concentration scan in D(He^3) plasmas allows an examination of the competition between minority heating and mode conversion mechanisms. Experimental measurements are compared with results from the modeling codes FPPRF amp; FELICE where possible. \$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.06] Sawteeth and Global Stability in Alcator C-Mod

The morphology and the physics of the sawtooth oscillations observed on Alcator C-Mod has been analyzed. Sawteeth are almost always present. They can vary from small, rapid oscillations, to the large sawteeth observed during RF H-modes, where the interval between crashes at the center can exceed 25 msec, several times the ohmic value, and the drop of central temperature at the crash can be 2 keV. The period is observed to be nearly independent of density, to increase with current and with ICRF power. A good correlation is found between the sawtooth period and the plasma stored energy over a range from 20 to 200 kJ. Modes with m=1, n=1 involving magnetic reconnection by finite electrical resistivity are considered to be responsible for the observed sawtooth crashes. The experimental equilibrium configurations have been found to be stable against ideal MHD modes, excluding those with long sawteeth. For these cases, reminiscent of the "monsters" observed at JET, the stabilizing effect of high energy particles \footnote B.Coppi, et al., Phys. Rev. Lett. 63, 2733 (1989) is considered.

\$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.07] Observation of High Energy Hydrogen Tail Ions during ICRF Heating in Alcator C-Mod

In Alcator C-Mod, the main heating scenario is based on the hydrogen minority heating at 80 MHz, with a toroidal field near 5.3 T. The energetic hydrogen population is studied via passive charge-exchange measurements, using two neutral particle analyzers (E\parallelB), one with a fixed tangential view, and one with a scannable poloidal and toroidal view. In low density plasmas (\le 1.5x10^20 m^-3), and low hydrogen concentration (\le5%), we observe tail ions with energies as high as 150keV (with 2.5 MW of RF power), whereas at higher densities, the hydrogen population is nearly Maxwellian. Preliminary comparisons of the ion velocity distribution with the FPPRF code (Fokker-Planck), indicate a reasonable agreement with the measured neutral particle energy spectrum. However, measurements of the sawtooth reheat rate, made after the RF is turned off, indicate that, compared to code predictions, a smaller amount of energy is stored in the tail, implying that the confinement of fast ions or the tail ion production is lower than anticipated. In addition, we will also explore the stabilizing effects of the tail ions on sawteeth. \$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.08] Observations of Ion Edge Heating during ICRH in Alcator C-Mod

Observations of edge ion heating during ICRF injection on Alcator C-Mod have been made using a toroidally and poloidally scanning charge-exchange neutral particle analyzer. The phenomenon is characterized by a large flux of charge-exchange neutrals (hydrogen and deuterium), at suprathermal energies, with a short rise time (\leq.2ms, the instrumental time resolution), but is not associated with impurity generation or loss of heating efficiency. Previous data showed that the RF power threshold for edge heating is decreased from 500kW to <10kW at certain values of the toroidal field. In this experiment, the energy spectra of escaping energetic neutrals were obtained for several RF power levels and particle pitch angles at three of these fields. It was found that the total energy in these suprathermal edge particles increases faster than linearly with applied RF power. Some conclusions are made about the structure of the damped field that generates the energetic ions.

\$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.09] Spatial Structure of Pump Wave and Parametric Decay Instabililty during ICRF Injection on Alcator C-Mod

The spatial structure of RF waves in the edge of Alcator C-Mod has been explored. The radial structure of the pump wave and the Parametric Decay Instability (PDI) it generates has been outlined by a shot-to-shot scan of stationary Langmuir probes, and will be measured by a new reciprocating probe during the summer run campaign. The toroidal and poloidal extent of the RF waves has been inferred from stationary probe data from three different toroidal and two different poloidal locations. Preliminary data indicates that the pump wave is not toroidally localized near the antennas. The frequency of the PDI by-product ion quasi-mode detected by the probe appears to vary with the local cyclotron frequency, and the PDI amplitude scales linearly with input RF power for constant plasma conditions. Edge wave fields during different RF heating scenarios are compared. The likelihood that the PDI mediates the oft-observed increase in impurity radiation during ICRF injection is evaluated. \$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.10] Ion Mass Spectrometry on Alcator C-Mod: First Results

An Omegatron probe designed for Alcator C-Mod [1] has recently been installed for operation during the summer run campaign. This diagnostic combines an ion mass spectrometer (Omegatron) with a gridded energy analyzer. The Omegatron employs a radiofrequency cavity to selectively collect impurity and background plasma ions with cyclotron frequencies between 0.1 \leq f_c \mbox (MHz) \leq 100. In principle, this diagnostic can measure the relative charge state densities of impurity and background ions in the scrape-off layer (SOL) as well their separate temperatures. Knowledge of the impurity concentrations, charge state distributions, and temperatures in the divertor and SOL are critically important both for modelling impurity transport and for optimizing dissipative divertor regimes. We expect ion currents in the range of 1 to 100 nA, and so must carefully consider noise sources in the tokamak environment. We present results from first operation in a tokamak plasma, including signal strength, signal/noise ratio, Z/M resolution, and impurity ion spectra.

[1] Thomas, E. Jr., MIT masters thesis, PFC/RR-93-03

\$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.11] Initial Results from the C-Mod Divertor Thomson Scattering System

Thomson scattering system has been installed recently to diagnose the x-point and divertor plasma regions with a resolution of 2-3 mm over a 12 cm field. The light scattered from a 30 HZ Nd:YAG laser is viewed from below through a slot in the outer divertor plate with a reentrant, high throughput collection system. A compact laser dump is located inside the inner divertor plate. Laser alignment is maintained under feedback control to track vessel motion. A filter polychromator spectrally resolves the scattered light from 25 spatial positions onto four 25 element avalanche photodiode arrays. System performance is described in terms of both calibration results and initial measurements of divertor plasma parameters. \$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013,

DE-AC02-76-CHO-3073 and SBIR Grant No. 20431-92-II.

[7R.12] Phase Contrast Imaging in Alcator C-Mod

A Phase Contrast Imaging (PCI) diagnostic is being set up in the Alcator C-MOD tokamak. It uses 18\rm\,cm wide infrared beam generated by a 100\,Watt CO_2 laser. The beam passes through the tokamak plasma core vertically. Electron density fluctuations inside the plasma cause the beam wavefront to distort and these distortions can then be observed by PCI. The diagnostic uses a 12-channel photovoltaic detector array optimized for 10.6\rm\mu m wavelength. This allows 1-dimensional (along the major radius) measurements of density fluctuations with long wavelength (up to the beam width) and a good frequency response (up to 100\,MHz). The expected sensitivity is about \tilde n \geq 10^9cm^-3. The PCI diagnostic can be used for observation of ICRF wave (80\,MHz) propagation and turbulence studies. The diagnostic design and the first measured data will be presented at the session. \$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.13] Results from a Prototype Second Harmonic Tangential Array Interferometer

A tabletop tangential array interferometer operating at 1.06 and 0.53 mm has been built to test its sensitivity for use on C-Mod. This type of interferometer is insensitive to mechanical vibrations. The prototype uses a pulsed, 35 mJ, 10 Hz multimode, Nd:YAG laser, LiB3O5 non-critically phase matched doublers, a fan beam created by a cylindrical lens, several retro-reflector elements to define the array, and a CCD camera detector. This interferometer has beam diameters of a few millimeters and spatial resolution of a few centimeters as needed for C-Mod. Commercial lasers and CCD arrays are available which can scale this design to \sim20kHz. Tokamak applications typically require a limiting sensitivity of 10^-3 fringes and maximum changes of \sim1 fringe. The limiting sensitivity of the prototype system has been investigated along with technical limitations on elements of the optical system relevant to its use on tokamaks. In particular methods of normalizing shot-to-shot and spatial mode variations in the laser intensity are examined. The use of visible and near visible components allows a compact optical design and efficient use of port space. \$^*Supported by US DoE Contract Nos. DE-AC02-76-CH0-3073 and DE-AC02-78ET51013.

[7R.14] Prototype Tangential Interferometer for Alcator C-Mod

A prototype single chord tangential interferometer using a frequency doubled, Nd:YAG, CW, diode pumped laser has been installed on Alcator C-Mod. Both the 1.064 and 0.532 \mum beams are used in a two color configuration such that vibrational effects can be subtracted. Since the beams are both harmonically related and phase locked, a Bragg cell can be driven at two frequencies to both deflect the beams to the same angle and provide the offset frequencies needed for phase measurements. Electronics designed to provide the plasma phase shift as a direct analog output will be discussed. An in-vessel retro-reflector will reduce the effects of alignment errors caused by vibrations and thermal drifts. Bench tests of the prototype and initial results from C-Mod will be presented.

\$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.15] X-Mode Reflectometry on Alcator C-Mod

The Alcator C-Mod reflectometer consists of five amplitude modulated, O-mode channels spanning 50 to 110 GHz and provides electron density profiles for densities up to 1.5 \times 10^20 m^-3. However, for most current C-Mod conditions, these densities are limited to the plasma edge. By switching to X-mode launch, we will be able to study electron densities up to 4.5 \times 10^20 m^-3 at 8 tesla. We are currently installing equipment to allow operation in the X-mode. Studies of profile and turbulence evolution during H and PEP modes are planned and will be presented. \$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.16] Measurements of Vacuum Vessel Strain During Disruptions in Alcator C-Mod

Disruptions in Alcator C-Mod induce large toroidal eddy currents and poloidal halo currents in the highly conducting vacuum vessel and plasma facing components. The halo currents tend to be highly asymmetric, typically having an n=1 toroidal structure, and this structure usually rotates toroidally at a few kHz. In order to quantify the resultant J\times B forces and the dynamic stresses they induce, solid-state strain gauges have been installed on the interior wall of the Alcator C-Mod vacuum vessel, behind the tiles of the inboard divertor nose. Previous measurements of the poloidal distribution of halo currents during disruptions indicate that this is the region of highest current density. Dynamic measurements of vacuum vessel strain during disruptions will be presented. Comparison to engineering calculations and implications for Alcator C-Mod performance limits will be discussed.

\vskip 2.0 truemm Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.17] Novel Surface Thermocouple Probes for Divertor Heat Flux Measurement

An array of novel surface thermocouple probes have been installed and tested in the outer divertor of Alcator C-Mod. These sensors can, in principle, record divertor surface temperatures with fast time response (\tau \ge 10 \mu sec), allowing a direct estimate of the plasma heat flux to be inferred. The design is an adaptation of a commercially available device(``The Self-Renewing Thermocouple,'' Nanmac Corp., Framingham, MA), employing a coaxial-like geometry with a single tungsten-rhenium ribbon wire embedded inside a 6.35 mm diameter molybdenum rod. Various prototypes were tested, including probes with flush and 5^\circ angles with respect to the divertor surface, and probes with and without protective surface coatings. Typical surface temperature rises are \sim 300-700 ^\circC, corresponding to signals of \sim 3-9 mV. RC filters with 10 ms time constants are used to reduce noise introduced by the plasma environment. The surface temperature corresponding to typical RMS noise levels is \sim 25 ^\circC. Using a one-dimensional, semi-infinite slab model, parallel heat fluxes in the range of 50-500 MW/m^2 are estimated. A comparison with heat flux estimates from Langmuir probes located adjacent to the thermocouple array will be presented. \$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.18] A Diagnostic Neutral Beam for Turbulence and Transport Experiments on Alcator C-Mod

We will install a 50 kV diagnostic neutral beam on C-Mod for measurement of \tilden_\i, n_Z, T_Z, v_\theta, v_\phi, j. Charge exchange recombination spectroscopy (CHERS) will be used for measurement of n_Z and T_Z. Applications will include \chi_i for any heating scheme and minority ion content in ICRF heating experiments. Beam emission spectroscopy will be applied for measurement of \tilden_i. The turbulence and profile diagnostics will contribute to an important benchmark for the IFS/PPPL model for tokamak transport dominated by ITG and TE mode turbulence. E_r, inferred from CHERS measurements of v_\theta and v_\phi, will be especially useful for studies of H-mode discharges when complemented with turbulence measurements in the same region. Capabilities for the intended measurements will be described using simulations based on measured C-Mod profiles. ^\astSupported by USDOE under Grant DE-FG03-94ER-54241 at UT-FRC and by USDOE Contract DE-AC02-78ET51013 at MIT.

[7R.19] High Resolution Measurements of Zeeman Patterns in Visible/Near Ultraviolet Spectra from the Alcator C-Mod Tokamak

Plasmas in Alcator C-Mod have been observed using the University of Maryland high resolution spectrograph in the near ultraviolet and visible range (200-700 nm). Zeeman splittings of emission lines are routinely measured in spectra recorded from light gathered via spatially-resolving fiber optic arrays. With magnetic field reconstructions along selected viewing chords, analysis of the Zeeman patterns of the spectra can aid in localizing the observed emission within the machine. The line shifts and normalized intensities for the lines of interest have been calculated for arbitrary magnetic fields and have been used to create simulated spectra. The distribution of emitters and the ion temperature are adjusted to achieve good agreement with the measured spectra. The analysis from several different chords for several ions will be presented, and the use of these analyses in the determination of Doppler shifts resulting from bulk flow will be discussed.

\$^*DOE Contract No. DE-AC02-78ET51013 and DE-FG02-95ER54307

[7R.20] High-Resolution Visible/Ultraviolet Measurements from the Alcator C-Mod Tokamak

High-resolution visible and ultraviolet light spectra have been taken of the emission from the edge and divertor regions of the Alcator C-Mod tokamak. These spectra indicate that transitions from the Balmer series of deuterium and transitions from boron I-IV are the predominant lines. The boron is present as a result of the boronization of the vessel walls. The lineshapes of the D_\alpha and boron transitions indicate Zeeman splitting (due to the 3-9 T magnetic field) and Doppler broadening (due to the neutral or ion temperature).

Spatially resolved views with chords approximately tangent to the field lines near the last closed flux surface in the main chamber and in the scrape-off layer in the divertor allow measurements of parallel flows. Initial observations from the divertor line of sight indicate a net shift of D_\alpha emission at times during the discharge. These shifts imply a toroidal flow of the neutral deuterium in the divertor. These results, as well as spectra of boron, will be presented.

\$^*Supported by U.S. DOE Contract Nos. DE-AC02-78ET51013 and DE-FG02-95ER54307.

[7R.21] Mo \footnotesizeXIV to Mo \footnotesizeXXIX

Spectra of low and intermediate charge states of molybdenum (Mo \footnotesize XVI to Mo \footnotesizeXXIX) have been measured between 65 and 80Å\ using a multilayer mirror based polychromator and a grazing incidence spectrometer. The spectral content of the observed `quasi-continuum' has been analyzed using high resolution spectra from TFR, TEXT and Alcator C-Mod tokamaks and a detailed collisional radiative model for each of the above charge states. The fractional abundance of each ion has been determined using the MIST transport code with updated ionization physics. The analysis focuses on the relative contribution of the low charge states (Mo \footnotesizeXVI to Mo \footnotesizeXXIII) versus that of the intermediate charge states (Mo \footnotesizeXXIV to Mo \footnotesizeXXIX) to the observed signal. Theoretical investigation of atomic quantities such as the mean transition wavelength and total array strength of the \Deltan=0 transition arrays in this region is presented. The utility and limitations of the `quasi-continuum' emission as a plasma diagnostic are discussed. \$^*Supported by U.S. DOE Grant DE-FG02-86ER53214 at JHU, Contract No. DE-AC02-78ET51013 at MIT, and No. W-7405-ENG-48 at LLNL.

[7R.22] Measurement of Neutron Production from Alcator C-Mod

Neutron rate measurements in excess of 1\times 10^14 neutrons/second lasting for several hundred milliseconds have been obtained routinely from Alcator C-Mod plasmas. These high neutron rate plasmas result from high power (3 MW) ICRF injection, following boronization of the vacuum vessel walls. These data are obtained from a set of 18 neutron detectors of differing sensitivity (12 U^235 fission chambers, 4 BF_3 and 2 He^3) divided among moderator stations located at four different sites in the experimental cell. Neutron measurements with fast time response are obtained from a bank of 14 He^3 detectors ganged together, which allow study of neutron sawteeth. The addition of a collimated neutron detector array to obtain spatially resolved measurement of the neutron source is planned. Zinc Sulfide detectors will be used to minimize the sensitivity to neutrons below 1 MeV in energy, and thus reduce the collimator size. The design of the collimator has been optimized using three dimensional neutron transport calculations (MCNP). The details of recent measurements from the existing neutron detectors will be presented. The final collimator design and the results obtained from the collimator prototype will be discussed as well.

\$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.23] The Alcator C-Mod Boronization System

A system for boronization of the Alcator C-Mod vacuum vessel and plasma facing components was installed and began routine operation in January of 1996. Significant suppression of impurity levels and record stored energy were achieved during ICRF heating following boronization. Electron cyclotron discharge cleaning in dilute diborane gas (10% D_2B_6, 90% helium) was selected as the simplest and most cost efficient method to achieve boronization in a short period of time. Safety of the diborane storage, delivery, and exhaust systems was the prime consideration in the overall design. In addition, complete remote operation of the system was required in order to minimize personnel exposure to potential gas leaks. The delivery system and its performance will be detailed, as well as discussion of unanticipated post operative clean up problems. Upgrades to the system utilizing multipoint diborane delivery and glow discharge paddles are currently being installed. \$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.24] Boronization on Alcator C-Mod

Experiments to investigate the effects of boronization on Alcator C-Mod were begun in the winter campaign. The boron layer was deposited onto the Mo plasma facing components using a low temperature electron cyclotron discharge cleaning plasma with deuterated diborane/helium gas (10%/90%). Wall coverage is roughly estimated at about 1000 Åof boron for each 12 hour application. Plasma performance was profoundly impacted. In 0.8MA, 5.3 tesla, ohmic fiducial plasmas, Mo levels were reduced by more than a factor of 10, and C, which had been the dominant low Z impurity (pre-boronization n_c/n_e\sim .5%), was reduced by about a factor of 3; oxygen levels were also reduced by about an order of magnitude; total core radiated power was reduced from 0.2\times P_in to \sim0.08\times P_in. Operational limits were extended, allowing access to quasi-steady-state H-modes with high power ICRF heating (P_RF\le 3.6 MW). For the highest power H-mode cases, core radiation was reduced by up to a factor of 3, and H-factors were increased up to 2.4. Net power flowing into the divertor was more than doubled, with q_\|\le 500 MW/m^2 available to challenge the power handling capabilities of the divertor. \$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.25] Observations of Parallel and Perpendicular Impurity Transport from Alcator C-Mod Plasmas

Spatially scanning x-ray and VUV spectrometer systems have been used to observe emission from injected impurities in Alcator C-Mod plasmas. Cross-field impurity transport coefficients have been determined by comparison between observed brightness profile time histories following injection of scandium and predicted profiles from MIST. During L-mode operation, the impurity diffusion coefficient is anomalously high, while in H-mode plasmas it approaches the neo-classical value, and a large inward convection velocity is seen near the plasma edge, also qualitatively similar to neo-classical predictions. Large up-down argon density asymmetries near the plasma edge are infered from x-ray observations of the forbidden line in Ar^16+, in qualitative agreement with the predictions of neo-classical parallel impurity transport. The edge impurity drift is in the direction opposite to the ion B\times\nablaB drift direction, and switches direction when the toroidal magnetic field direction is reversed. Central toroidal impurity rotation during ohmic discharges has been measured from the doppler shift of argon and molybdenum x-ray lines. The impurities in the center rotate in the same direction as the electrons and these observations are also in qualitative agreement with the predictions of neo-classical theory. \$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013

[7R.26] Kinetic Effects in the Alcator C-Mod Scrape-off-layer

We present recent results from a kinetic simulation [1] of parallel electron transport in the scrape-off-layer (SOL) of Alcator C-Mod. We show that non-local effects result in strong deviation of the electron distribution function from equilibrium (elevated or depleted tail) for experimentally observed detached and attached SOL plasma profiles [2]. This leads to a significant change in plasma parallel heat conductivity, and modifies hydrogen excitation and ionization near the divertor plate. We compare our results with data from reciprocating and divertor Langmuir probes. We estimate, what will be kinetic effect on probe and Thompson scattering electron temperature measurements. We give a comparison between numerical predictions and spectroscopic measurements of impurity line radiation. \vskip .05 cm \hskip -1.55 cm [1] A.A.Batishcheva, et al., Bull. APS, 40, No.11, 1702 (1995). \vskip .0 cm \hskip -1.55 cm [2] I.H.Hutchinson, et al., Physics of Plasma 1, 1511 (1994). \vskip .0cm \hskip -1.55 cm ^*Supported by U.S. DOE Contracts No. DE-FG02-91-ER-54109 and DE-AC02-78ET51013 at MIT, and DE-FG02-88-ER-53263 at Lodestar.

[7R.27] Simulation of Divertor Detachment in Alcator C-Mod

The non-orthogonal 2-D edge plasma fluid code UEDGE has recently been upgraded with a Navier-Stokes fluid model for the neutral particles. Using this model we have successfully simulated detachment in ITER and Alcator C-Mod. Detachment in C-Mod is partial, extending to the divertor nose but not beyond, in agreement with the experimental observations. In this work we extend the C-Mod simulations to other discharge conditions and we investigate the influence of several parameters on partial detachment, such as H-mode and L-mode operation, divertor geometry, and impurity characteristics. The simulated edge plasma conditions are matched to divertor and scrape-off-layer probe measurements and comparisons are made with spatially resolved H-alpha measurements.

\$^*Supported by U.S. DOE Contract No. DE-AC02-78ET51013 and by the European Community under an association contract between EURATOM and Sweden

[7R.28] Research Program for TdeV-96.

A major upgrade of the TdeV tokamak has been completed recently. The upgrade consists of a complete redesign of the divertor hardware, and of the addition of several power supplies. In the new configuration, TdeV-96, the effect of varying the divertor geometry will be examined. Major points include helium pumping from the private flux region and variation of the baffle distance at the divertor entrance. The results of these experiments will be compared with the predictions of the B2-EIRENE edge code. The additional flexibility of the magnetic configuration, and the increased heating and current drive power (both lower hybrid and electron cyclotron) is ideal for the investigation of advanced tokamak scenarios which constitutes the other major program point of TdeV-96.

\altaffiltext*Supported by the Government of Canada, Hydro-Québec and INRS

[7R.29] Fast Calculation of MHD Equilibrium Parameters for TdeV.

A method for calculating equilibrium plasma parameters, based on a multivariate statistical analysis of a data base generated by either MHD Lackner code or real plasma shots, will be presented. A reduced set of flux loop and coil current measurements are used in a regression to determine plasma parameters such as plasma major and minor radius, saddle point position, separatrix strike point on the divertor plates, safety factor, elongation and \beta_p + l_i/2. The main advantage of this method is that it has the option to calculate these parameters and their time evolution for real-time plasma control. Using two INTEL860 processors, this approach will be employed for the TdeV-96 control system, which is now in development. Different scenarios will be presented.

\altaffiltext*Supported by the Government of Canada, Hydro-Québec and INRS

[7R.30] Study of LHCD physics through suprathermal electron diagnostics on TdeV.

A set of diagnostics based on X-ray emission and electron cyclotron absorption is used to characterize the LH generated suprathermal population. The propagation and absorption of LH waves, the spatial diffusion of fast electrons and the possible existence of a velocity space instability are analysed in order to obtain consistency between the experimental results and the predictions of a combination of ray-tracing, 2D Fokker-Planck and X-ray bremsstrahlung emission codes. Partial results are that: the data sets an upper bound of \sim 1 m^2/sec for the spatial diffusion coefficient; there is evidence that the LH power deposition profile is broader than what the codes predict; and the LH plateau electron distribution extends beyond what is calculated by the codes, even when the residual V_loop is taken into account. The level of current profile control that can be expected with LH waves for the latest TdeV upgrade is assessed.

\footnotemark[6]\footnotetext[6]P. Brooker et al, Proceedings of the...

\altaffiltext*Supported by the Government of Canada, Hydro-Québec and INRS \addtocounterfootnote-1

[7R.31] Acceleration of Electrons in the Vicinity of a Lower Hybrid Waveguide Array.

The interaction of tokamak plasma edge electrons with the electric near-field generated by a lower hybrid slow wave antenna is studied. Waveguide array power spectra of interest for non-inductive current drive and/or plasma heating have lobes at high n_\parallel values (n_\parallel > 30) due to small scale-length structures of the antenna. It is shown that near the antenna, the electric field associated with the high n_\parallel components is large enough for resonant acceleration of the relatively cold (\sim 25 eV) edge electrons. For waveguide electric fields, typically around 3 kV/cm, the higher-order modes overlap in the Chirikov sense [Phys. Reports 52, 263 (1979)], so that electron global stochasticity is induced, allowing the electrons to gain energy up to the stochastic limit. For TdeV conditions and for 90\arcdeg waveguide array phasing, the stochastic limit is about 2 keV, determined by the 5th harmonic - the last overlapping mode. The progress of electrons through accessible phase space is very efficient: the TdeV 32 waveguide array can accelerate the electrons to the possible limit. An area-preserving map is derived to study the electron dynamics. \vspace-.2cm

\altaffiltext*Supported by the Government of Canada, Hydro-Québec and INRS

[7R.32] The 110 GHz ECRH System for TdeV.

A new microwave system operating at the second harmonic of the Electron Cyclotron Resonance is being installed on the TdeV tokamak. It is planned to be operational in the spring of 1997. Its main objectives are to provide efficient and well localised heating and current drive up to the X2-mode cutoff density (n_e=7.4 \times 10^19 m^-3). The generator will consist of two new GLGF-110/1.0 gyrotrons developed by Gycom. The output power of each tube after conversion to the HE_1,1 waveguide mode is 700 kW at \hboxf_0 = 110 GHz for pulse lengths of 2 s. The rf transmission lines which are made of both quasi-optical and waveguide components will be described. Their transmission losses are evaluated to be less than 12% leading to a total injected power of 1.2 MW. Launching structures with mirrors steerable in the poloidal and toroidal directions are used to control the power deposition. The launching position was optimized using the TREC ray tracing code which includes weakly relativistic approximations. Simulations of ECCD experiments in TdeV with the BANDIT-3D code will also be presented. \vspace-.2cm

\altaffiltext*Supported by the Government of Canada, Hydro-Québec and INRS

[7R.33] Kinetic Modelling of Parallel Electron Transport in TdeV.

The 1D2V Fokker-Planck code ALLA [1] is used to model parallel electron transport in the SOL of TdeV. Our model uses detached and attached experimental data [2] given by Langmuir probes, Li ablation and He spectroscopy.

We obtain the electron distribution function on a precise 257x65x60 non-uniform grid. Strong deviations of hydrogen and carbon excitation rates, and heat conduction coefficient from their Maxwellian values are shown. We compare the calculated variation of the effective temperature at the reciprocating probe position with experimental measurements. We also explain by non-local effects why different experimental techniques show differences in the electron temperature.

\altaffiltext [1] A.A.Batishcheva et al., Physics of Plasmas 3 (1996) 1634 \altaffiltext [2] B.L.Stansfield et al., Proc. 22 Eur.Conf., Bornemouth, 19C pIII-101.

\altaffiltext*Supported by Government of Canada, Hydro-Québec and INRS. \altaffiltext**Work performed under USDoE contracts DE-FG02-91-ER-54109 at MIT and DE-FG02-88-ER-53263 at Lodestar. \vspace-.2cm

[7R.34] Finite Element Modelling of Transport and Drift Effects in Tokamak Divertor and SOL.

A finite element code is used to simulate transport of a single-species plasma in the edge and divertor of a tokamak. The physical model is based on Braginskii's fluid equations for the conservation of particles, parallel momentum, ion and electron energy. In modelling recycling, transport of neutral density and energy is treated in the diffusion approximation. The electrostatic potential is obtained self-consistently from the charge conservation equation and from the generalized Ohm's law. In the transport equations, particle drifts (both E\timesB and diamagnetic) are included. Transport also accounts for a current flowing in the edge. Simulations with different types of boundary conditions, recently proposed in the literature, are considered and assessed. Comparisons are made between simulation and experimental results from TdeV. Particular attention is given to density and temperature profiles at the divertor plates, and to the plasma parallel velocity in the SOL with and without divertor plate biasing.

\altaffiltext*Supported by the Government of Canada, Hydro-Québec and INRS

[7R.35] Optimization of the TdeV high energy diagnostic neutral beam injector.

The development of a 40 kV, 1A diagnostic neutral beam injector for TdeV is near completion. The plasma source for this injector is a duoPIGatron source, which is maintained at ground potential. The neutralizer tube is at -40 kV. An electrostatic trap is used to retain the electrons in the neutralizer tube. Several diagnostics have been used to characterize the beam performance. For a single beamlet, the beam divergence was optimized using a scanning multi-pin open probe array. Up to 600 mA of ion current, at 35 kV, has been extracted from seven beamlets. Beam emission spectroscopy has been used to optimize the full energy component of the beam. A diamagnetic loop on the plasma source has been used for optimization and monitoring of the beam current fluctuations. We are in the process of installing an infrared imaging camera to study the beam power profile. Results will be presented at the meeting.

\altaffiltext*Supported by the Government of Canada, Hydro-Québec and INRS

[7R.36] Properties of LHCD Suprathermal Electron Distributions in TdeV*

Suprathermal electron distribution functions have been measured in TdeV using absorption of an extraordinary mode electron cyclotron wave on two chords through the plasma. The results were obtained during the steady-state portion of LHCD discharges. Distributions fluctuate on timescales of a fraction to a few milliseconds. Sometimes on-axis chord distributions are symmetric in the parallel direction to several tenUs of keV. The off-axis chord always shows a high energy plateau extending to several times the energy associated with the injected LH wave Nll. Simultaneous measurements on both chords show that at high energies, the spatial distribution is hollow

[7R.37] Thallium Heavy Neutral Beam Probe on TdeV: Final Results and Future Plans

The Thallium Neutral Beam Probe was operational on the Tokamak de Varennes for the experimental campaign of 1995. The diagnostic performed measurements of the edge potential profile and plasma turbulence in ohmically heated plasmas with floating and biased divertor plates. The results of these measurements will be summarized. TdeV has recently finished a significant upgrade which alters the operational parameters required for a beam probe diagnostic. A new system has been conceived to provide measurements of the TdeV plasma from the edge to the core plasma. The system will be designed to allow for current density profile measurements to be made as well as potential profiles and plasma turbulence measurements. The preliminary design criteria will be reviewed and design options presented.

[7R.38] The Compact Toroid Fueller - II Device.

In our previous experiments with the first generation CTF device we were able to experimentally demonstrate compact toroid fuelling of TdeV (at B-toroidal of 1.4 T) without any adverse perturbation to the tokamak discharge [1]. These favorable results led to the fabrication of a second generation injector, CTF-II, for CT injection into a 2 T field. Initial operation of this injector has begun. CTF-II is about 40 percent larger in radius than CTF. In addition, all plasma facing surfaces are coated with dense tungsten and maintained at 150 degrees C even during injector operation. CT acceleration measurements and tokamak injection results will be presented. \altaffiltext[1] R. Raman, F. Martin, M. St-Onge et al., Enhanced Plasma Confinement after Compact Toroid Fuelling, Proceedings of the 23rd EPS conference on plasma physics and controlled fusion, 24 - 28 June 1996 Kiev, Ukraine

\altaffiltext*Supported by the Government of Canada, Hydro-Québec and INRS

Part 7 of program listing