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Routine tritium operation in TFTR has permitted investigations of alpha particle physics in parameter ranges resembling those of a reactor core. ICRF wave physics in a DT plasma and the influence of isotopic mass on supershot confinement have also been studied. Continued progress has been made in optimizing fusion power production in TFTR, using extended machine capability and Li wall conditioning. Performance is currently limited by MHD stability. A new reversed magnetic shear regime is being investigated with reduced core transport and a higher predicted stability limit.

After many years of fusion research, the conditions needed for a D–T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D–T plasma are now being studied in a laboratory plasma.The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral‐beam heating, in a supershot discharge and 6.7 MW in a high‐βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m−3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter‐H‐mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high‐βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first‐orbit loss mechanism. No loss due to alpha‐particle‐driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approximately}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the P{sub FUS} {approximately}6 MW level. Instability in the TAE mode frequency range has been observed at P{sub FUS} > 7 MW and its effect on performance is under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fullymore » explored.« less

The interaction between a lower hybrid wave and a fusion alpha particle displaces the alpha particle simultaneously in space and energy. This results in coupled diffusion. Diffusion of alphas down the density gradient could lead to their transferring energy to the wave. This could, in turn, put energy into current drive. An initial analytic study was done by Fisch and Rax (1992). Here the authors calculate numerical solutions for the alpha energy transfer and study a range of conditions that are favourable for wave amplification from alpha energy. They find that it is possible for fusion alpha particles to transfer a large fraction of their energy to the lower hybrid wave. The numerical calculation shows that the net energy transfer is not sensitive to the value of the diffusion coefficient over a wide range of practical values. An extension of this idea, the use of a lossy boundary to enhance the energy transfer, is investigated. This technique is shown to offer a large potential benefit.

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a FluorinertTM system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium–deuterium (D–D) run to simulate expected deuterium–tritium (D–T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D–T experiments using D–D have been performed. The physics objectives of D–T operation are production of ≊10 MW of fusion power, evaluation of confinement, and heating in deuterium–tritium plasmas, evaluation of α‐particle heating of electrons, and collective effects driven by alpha particles and testing of diag...

Recent research on TFTR has emphasized optimization of performance in deuterium plasmas, transport studies and studies of energetic ion and fusion product physics in preparation for the D-T experiments that will commence in July of 1993. TFTR has achieved full hardware design parameters, and the best TFTR discharges in deuterium are projected to QDT of 0.3 to 0.5.The physics phenomena that will be studied during the D-T phase will include: tritium particle confinement and fueling, ICRF heating with tritium, species scaling with tritium, collective alpha-particle instabilities, alpha heating of the plasma and helium ash buildup. It is important for the fusion program that these physics issues be addressed to identify regimes of benign alpha behavior, and to develop techniques to actively stabilize or control instabilities driver by collective alpha effects.

Detached plasmas have been proposed as a means of distributing power losses uniformly on walls in ITER and other reactor scenarios. Ohmically and neutral beam injection (NBI) heated detached plasma studies are being carried out on TFTR in order to help determine the practicality of this concept. NBI heated detached plasmas with up to 8.5 MW of beam power have been maintained detached for up to 300 ms. A confinement related delay in the radiated power is observed at low beam power. H-mode-like events have been observed for several beam heated detached plasmas and radiative emissivities as high as 0.8 MW/m 3 have been realized. These results are especially of interest to ITER and other reactor design studies.

We have developed a code for following the radial diffusion and slowing down of alpha particles from DT fusion. The partial differential equation is solved by a finite‐difference method. Ion as well as electron drag is included. Any given source function of radius, time, and energy width can be handled correctly. The code has been used to examine time waveforms, radial distributions, and energy spectra. The results can be used to model signals expected from the alpha charge exchange recombination spectroscopy and collective Thomson (‘‘gyrotron’’) scattering diagnostics. The code has also been used to calculate, as a function of diffusion coefficient, the flux of escaping fast ions that enters a lost‐ion detector, and thus to deduce from measurements an upper limit for the diffusion of fast charged fusion products.

The TRANSP code is used to construct comprehensive, self-consistent models for ITER discharges. Plasma parameters are studied for two discharges from the ITER ``Interim Design`` database producing 1.5 GW fusion power with a plasma current of 21 MA and 20 toroidal field coils generating 5.7 T Steady state profiles for T{sub ion}, T{sub e}, n{sub e}, Z{sub eff}, and P{sub rad} from the database are specified. TRANSP models the full up/down asymmetric plasma boundary within the separatrix. Effects of high-energy neutral beam injection, sawteeth mixing, toroidal field ripple, and helium ash transport are included. Results are given for the fusion rate profiles, and parameters describing effects such as alpha particle slowing down, the heating of electrons and thermal ions, and the thermalization rates. The deposition of 1 MeV neutral beam ions is predicted to peak near the plasma center, and the average beam ion energy is predicted to be half the injected energy. Sawtooth mixing is predicted to broaden the fast alpha profile. The toroidal ripple losses rate of alpha energy is estimated to be 3% before sawtooth crashes and to increase by a factor of three to four immediately following sawtooth crashes. Assumptions for the thermal He transport and the He recycling coefficient at the boundary are discussed. If the ratio of helium and energy confinement times, {tau}*{sub He}/{tau}{sub E} is less than 15, the steady state fusion power is predicted to 1.5 GW or greater. The values of the transport coefficients required for this fusion power depend on the He recycling coefficient at the separatrix. If R{sub rec} is near 1, the required He diffusivity must be much larger than that measured in tokamaks. If R{sub rec} {le} 0.50, and if the inward pinch is small, values comparable to those measured are compatible with 1.5 GW.

The equation for the radial diffusion and slowing down of fast ions in a plasma is solved by a finite-difference technique. The terms included are ion source, radial diffusion, electron and ion drag. From the ion density at the radial boundary, the loss flux is calculated and used to model the signals in a lost-ion diagnostic. The code is also used to model the density of {alpha}-particles in future DT experiments. This information is used to predict the features to be seen by alpha diagnostics.

Results obtained from experiments utilizing high power ICRF (ion cyclotron range of frequency) heating to stabilize sawtooth oscillations on TFTR are reviewed. The key observations include existence of a minimum ICRF power required to achieve stabilization, a dependence of the stabilization threshold on the relative size of the ICRF power deposition profile to the q=1 volume, and a peaking of the equilibrium pressure and current profiles during sawtooth-free phases of the discharges. In addition, preliminary measurements of the poloidal magnetic field profile indicate that q on axis decreases to a value of 0.55{plus minus}0.15 after a sawtooth-stabilized period of {approximately}0.5 sec has transpired. The results are discussed in the context of theory, which suggests that the fast ions produced by the ICRF heating suppress sawteeth by stabilizing the m=1 MHD instabilities believed to be the trigger for the sawtooth oscillations. Though qualitative agreement is found between the observations and the theory, further refinement of the theory coupled with more accurate measurements of experimental profiles will be required in order to complete quantitative comparisons.

A transport code (TRANSP) is used to simulate future deuterium-tritium experiments (DT) in TFTR. The simulations are derived from 14 TFTR DD discharges, and the modeling of one supershot is discussed in detail to indicate the degree of accuracy of the TRANSP modeling. Fusion energy yields and {alpha}-particle parameters are calculated, including profiles of the {alpha} slowing down time, average energy, and of the Alfven speed and frequency. Two types of simulations are discussed. The main emphasis is on the DT equivalent, where an equal mix of D and T is substituted for the D in the initial target plasma, and for the D{sup O} in the neutral-beam injection, but the other measured beam and plasma parameters are unchanged. This simulation does not assume that {alpha} heating will enhance the plasma parameters, or that confinement will increase with T. The maximum relative fusion yield calculated for these simulations is Q{sub DT} {approx} 0.3, and the maximum {alpha} contribution to the central toroidal {beta} is {beta}{sub {alpha}}(0) {approx} 0.5%. The stability of toroidicity-induced Alfven eigenmodes (TAE) and kinetic ballooning modes (KBM) is discussed. The TAE mode is predicted to become unstable for some of the equivalent simulations, particularly after the terminationmore » of neutral beam injection. In the second type of simulation, empirical supershot scaling relations are used to project the performance at the maximum expected beam power. The MHD stability of the simulations is discussed.« less

At the beginning of the 1990 TFTR experimental run, after replacement of POCO-AXF-5Q graphite tiles on the midplane of the bumper limiter by carbon fiber composite (CFC) tiles and prior to any Pulse Discharge Cleaning (PDC), boronization was performed. Boronization is the deposition of a layer of boron and carbon on the vacuum vessel inner surface by a glow discharge in a diborane, methane and helium mixture. The amount of discharge cleaning required after boronization was substantially reduced compared to that which was needed after previous openings when boronization was not done. Previously, after a major shutdown, about 10{sup 5} low current ({approximately}20 kA) Taylor Discharge Cleaning (TDC) pulses were required before high current ({approximately}400 kA) aggressive Pulse Discharge Cleaning (PDC) pulses could be performed successfully. Aggressive PDC is used to heat the limiters from the vessel bakeout temperature of 150{degrees}C to 250{degrees}C for a period of several hours. Heating the limiters is important to increase the rate at which water is removed from the carbon limiter tiles. After boronization, the number of required TDC pulses was reduced to

Central ion temperatures up to 30 keV and rotation speeds up to 8/times/10/sup 5/ m/sec have been confirmed with new diagnostic measurements in the TFTR hot-ion enhanced-confinement regime, and the ion thermal diffusivity is found to be non-neoclassical and comparable to the anomalous electron thermal diffusivity. The dominant effect of strong rotation is the down-shifting of the neutral beam energies in the plasma frame, which results in reduced ion and electron heating on axis, and the presence of off-axis ion heating from viscous damping of the plasma rotation.

A specialized method has been developed to maximize the two‐dimensional detail obtained from two perpendicular bolometer arrays. The technique relies on the assumption that poloidal variation exists only near the plasma surface. The cross section is divided into appropriate zones and the emittance is reconstructed by a numerical method. The position, intensity, and width of large features are clearly displayed. A marfe is tracked as it drifts around the plasma and evolves into a radiating shell detached from the limiter. A central peak, plus inner‐wall radiating layer, plus a marfe appear in a high‐density case reached by pellet injection.

Neutral-beam heating of plasmas in the Tokamak Fusion Test Reactor at low preinjection densities (n/sub e/(0)approx. =10/sup 19/ m/sup -3/) were characterized by T/sub e/(0) = 6.5 keV, T/sub i/(0) = 20 keV, n/sub e/(0) = 7 x 10/sup 19/ m/sup -3/, tau/sub E/ = 170 msec, ..beta../sub t//sub h//sub e//sub t//sub a/ = 2, and a d(d,n)/sup 3/He neutron emission rate of 10/sup 16/ sec/sup -1/. The ion temperature and the deuterium-fusion neutron yields were significantly higher than for previous tokamak experiments. The low initial densities were achieved by operation of the Tokamak Fusion Test Reactor with low plasma currents (less than or equal to1 MA) and by extensive limiter conditioning.

TSTR (Tokamak Fusion Test Reactor) has operated since December 1982 with ohmically heated plasmas. Routine operation with feedback control of plasma current, position and density has been obtained for plasmas with Ip800 kA, a = 68 cm, R = 250 cm, and Bt=27 kG. A maximum plasma current of 1 MA was achieved with q2.5. Energy confinement times of ~150 msec were measured for hydrogen and deuterium plasmas with e = 2 x 1013 cm-3, Te(0) 21.5 keV, Ti(0) = 1.5 keV and Zeff1 3. The preliminary results suqgest a size-cubed scaling from PLT, and are consistent with Alcator C scaling where T ~ nR2a. Initial measurements of plasma disruption characteristics indicate current decay rates of ~ 800 kA in 8 ms which is within the TFTR design requirement of 3 MA in 3 ms.

Quantitative spectroscopic measurements of Zeff, impurity densities and radiated power losses have been made for ohmically heated and neutral beam heated TFTR discharges at a plasma current of 2.2 MA and a toroidal field of 4.7 T. Variations in these quantities with line average plasma density (e) and beam power up to 5.6 MW are presented for discharges on a movable graphite limiter. A detailed discussion of the use of an impurity transport model to infer absolute impurity densities and radiative losses from line intensity and visible continuum measurements is given. These discharges were dominated by low-Z impurities, with carbon having a considerably higher density than oxygen, except in high e Ohmic discharges where the densities of carbon and oxygen were comparable. Metallic impurity concentrations and radiative losses were small, resulting in hollow radiated power profiles and fractions of the input power radiated being 30–50% for Ohmic heating and 30% or less for beam heating. Spectroscopic estimates of the radiated power were in good agreement with bolometrically measured values. Because of an increase in the carbon density, Zeff rose from 2.0 to 2.8 as the beam power increased from 0 to 5.6 MW, pointing to a potentially serious dilution of the neutron producing plasma ions with increasing beam power. Both the low-Z and the metallic impurity concentrations were approximately constant with minor radius, indicating no central impurity accumulation in these discharges.

Methods are discussed for reconstructing spatial distributions from two types of projection-in a direction tangential to a torus and in a cross-sectional plane. For a tangential projection the reconstruction is unambiguous. For cross-sectional projections one usually does not have enough information for an unambiguous reconstruction, unless strong simplifying assumptions are made. A method is given for fitting a spatial distribution to such projections. Tests of both methods are discussed. The fitting method for cross-sectional projections is compared with a Fourier-series method.

We describe measurements of tokamak edge plasma perturbations caused by rapid injection of impurities such as Sc, Fe, and Mo into Ohmically heated PLT discharges. The temporal evolution of the radiated power, convected power, charge‐exchange neutral flux, electron temperature, density, and floating potentials were monitored using recently developed fast bolometers, directional calorimeters, and other diagnostics. The central radiated power can be doubled without plasma disruptions. Radiation from the edge then increases tenfold for ∼2 ms. Neutral efflux at the limiter decreases up to two orders of magnitude for approximately 20 ms after injection. The electrical potential of the limiters increases, approaching the potential of the vacuum vessel. MHD activity in the m=1 (sawtooth) mode tends to increase while that in the m=2 and m=3 modes decreases. The power flowing along field lines in the limiter shadow region sometimes increases or sometimes decreases by more than a factor of 4. The density and electron temperature in the limiter shadow region, generally, does not change more than 50%.

A bolometer has been developed based on a thin, die‐cut platinum grid. It can survive high temperatures and the neutron and gamma radiation expected in the Toroidal Fusion Test Reactor (TFTR). The platinum resistance is measured with a square‐wave carrier system to minimize sensitivity to ambient electromagnetic interference. Electrical power fed back to the sensor holds its temperature constant and provides an output directly proportional to absorbed radiation power. With a bandwidth of 50 Hz the noise is equivalent to 100 μW/cm2. Methods are described for dealing with the background effects expected to contribute to bolometer heating.

The efflux of neutral hydrogen from PLT during discharge cleaning has been measured using a time‐of‐flight spectrometer. During high ionization pulsed discharge cleaning (PDC), the flux in the energy range of 5 to 1000 eV varies from 1014 H0/cm2⋅s to 1016 H0/cm2⋅s and the average energy from 10 to 80 eV. The energy distributions are nearly single temperature Maxwellians. Low ionization PDC (Taylor‐type) produces a 1000 times lower fluence in the same energy range; however, a flux of 1016 H°/cm2⋅s at energies less than 5 eV is inferred. The detailed submillisecond time variation of these parameters with the fill gas pressure and state of cleanliness of the machine is presented. Comparisons with UV spectroscopy, bolometric measurements, and residual gas analysis are made.

Initial operation of the tokamak fusion test reactor has concentrated upon confinement studies of Ohmically heated hydrogen and deuterium plasmas. Total energy confinement times (${\ensuremath{\tau}}_{E}$) are 0.1-0.2 s for a line-average density range (${\overline{n}}_{e}$) of (1-2.5)\ifmmode\times\else\texttimes\fi{}${10}^{19}$ ${\mathrm{m}}^{\ensuremath{-}3}$ with electron temperatures of ${T}_{e}(0)\ensuremath{\sim}1.2\ensuremath{-}2.2$ keV, ion temperatures of ${T}_{i}(0)\ensuremath{\sim}0.9\ensuremath{-}1.5$ keV, and ${Z}_{\mathrm{eff}}\ensuremath{\sim}3$. A comparison of Princeton large torus, poloidal divertor experiment, and tokamak fusion test reactor plasma confinement supports a dimension-cubed scaling law.

For the past year we have been making use of a horizontally viewing 19‐channel array and a bolometer which views a narrow cross‐sectional slice of the plasma. More recently, we have also obtained results from a second, vertically viewing array. Software has been developed to translate the data from general plasma and array locations to plasma minor radius and to do the Abel inversion with an antisymmetrical term included. Experience has been obtained on the noise and response‐time characteristics, as well as the accuracy of total radiated power and radial profiles. Representative cases of radiated power profiles and local power balance are presented, as well as comparisons with other measurements of impurity concentration and trends with electron density and limiter coating. Although most of the ohmic‐heating input power leaves by radiation, most of this loss occurs near the outer part of the plasma. Also, the behavior of power profiles during neutral beam injection is discussed briefly.

Designing bolometers for Compact Ignition Tokamak presents difficult, perhaps insurmountable, challenges. Detector environment will be quite harsh, viewing access will be severely limited, and, most importantly, the heating of detectors by neutron and gamma radiation will be larger than that by the radiation losses to be measured. Furthermore, the radiation losses themselves will no longer be dominated by impurity radiation as in earlier machines, but will be a combination of impurity radiation, bremsstrahlung, and synchrotron radiation. Possible approaches to a design have been developed, based on optimizing the response to VUV‐ and x‐radiation, possible local shielding, subtraction of the n and γ heating, constant in situ calibration, and deployment of the detectors in a tangentially viewing array. The array makes it possible to cover the plasma cross section, as well as to see into the divertors. Estimates indicate that it should be possible to make successful measurements of the two‐dimensional radiat...

Data from magnetic pick-up loops (Mirnov coils), located on the wall inside the vacuum vessel of TFTR, are used for studying the edge magnetic fluctuations. Experiments, such as impurity injection, gas puffing, and plasma motion, dramatically affect the fluctuations measured by the coils. A quantitative study of the fluctuation levels during these experiments has been made. Results show that there are simple relations between the amount of impurities or gas injected and changes in the fluctuation levels. Spatial locations of the fluctuation modes have also been tentatively identified. Finally, different models were studied in order to explain the behavior and dependence of the fluctuations on the relevant parameters of the plasma. 7 refs., 12 figs.

The Tokamak Fusion Test Reactor (TFTR) was designed to explore plasma confinement and heating at reactor-like parameters. Since the first plasma in December 1982, considerable progress has been made in this endeavor. During the run period ending April 13, 1985, operation of both the toroidal field and plasma current at full design parameters was achieved. The plasma parameters achieved are summarized in Table 1.

For the past year we have been making use of a horizontally viewing 19-channel array and a bolometer which views a narrow cross-sectional slice of the plasma. More recently, we have also obtained results from a second, vertically viewing array. Software has been developed to translate the data from general plasma and array locations to plasma minor radius and to do the Abel inversion with an antisymmetrical term included. Experience has been obtained on the noise and response-time characteristics, as well as the accuracy of total radiated power and radial profiles. Representative cases of radiated power profiles and local power balance are presented, as well as comparisons with other measurements of impurity concentration and trends with electron density and limiter coating. Although most of the ohmic-heating input power leaves by radiation, most of this loss occurs near the outer part of the plasma. Also, the behavior of power profiles during neutral beam injection and disruptions is discussed briefly.

Although the total radiated power in TFTR is often as high as 70% of the heating power, most of the radiation is concentrated near the surface of the plasma, and the interior loss is almost negligible. Fractional radiation loss declines during neutral beam heating. As the high-density limit is approached, under practically any conditions, a bright band of radiation (''marfe'') appears on the inner side of the plasma column. Marfe evolution has been tracked for various conditions of heating, magnetic field direction, and fueling. In enhanced-confinement discharges a peculiar bright band, distinct from a marfe, appears in the lower, outside part of the vacuum vessel, outside of the limiter radius. With pellet fueling, marfe evolution occurs at a higher average density, and an intense peak often appears in the plasma center, consisting of impurity radiation plus working-gas bremsstrahlung. In high density ohmically heated cases with pellet fueling radiative power loss exceeds heating power in the center. 34 refs., 19 figs.

Detached plasmas on TFTR have been heated with neutral beam auxiliary power for the first time. At beam powers above 2 MW the detached plasmas in TFTR expand and reattach to the limiters. Deuterium and/or impurity gas puffing can be used to maintain plasmas in the detached state at powers of over 5 MW. Transient events were observed in a number of these plasmas, including a confinement-related delay in evolution of the edge emissivity and some phenomena which appear similar to those seen in the H-mode. 16 refs., 5 figs.

Neutral beam injection experiments or PLT and PDX have been conducted over a wider range in parameter space than previously. On PLT H O beams have been injected into well-confined high toroidal field, high density Ohmic plasmas, giving n e (0) T Ee products during injection of up to 5 × 10 12 sec cm −3 . T Ee is found to rise slowly with increasing density in these experiments. Comparing these results with earlier (1979) discharges, which showed much lower heating efficiency, the importance of starting with a hot Ohmic plasma and a peaked density profile is striking. On PDX high power injection experiments over a range in plasma current have shown a significant variation with current of both ion heating and total stored plasma energy. Transport analysis of these results indicates that global confinement drops little when I p is varied from 480 to 320 kA, but as I p falls to 200 kA, T E deteriorates significantly. Attention has recently focused on low q operation, for the purpose of obtaining high beta plasmas. PLT and PDX have both been operated with cylindrical q in the range of 2.0 or below, with absorbed beam power of 1.8 MW (D O ). Operating in this regime, the two machines have obtained similar results, with PDX reaching a toroidal β of 1.3 − 1.4%, at B t = 12.6 kG. These results suggest that parallel and perpendicular neutral beam injection are about equally well suited tor low q operation. Strong sawtooth oscillations have been observed to travel to the plasma edge in PDX, but confinement is not substantially degraded relative to the higher toroidal field PDX results reported here.