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Experiments in the Tokamak Fusion Test Reactor (TFTR) have explored several novel regimes of improved tokamak confinement in deuterium-tritium (D-T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high-l{sub i}). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in-situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q{sub a} {approx} 4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l{sub i} plasmas produced by rapid expansion of the minor cross-section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D-T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D-T plasmas with q{sub 0} > 1 and weak magnetic shear in the central region, a toroidal Alfven eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode-conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions.

Operation of the Tokamak Fusion Test Reactor (TFTR) with a mixture of deuterium and tritium fueling has permitted the opportunity to measure the retention of tritium in the graphite limiter and other internal hardware. The use of discharge cleaning techniques and venting to remove the tritium was investigated. The tritium was introduced into TFTR by neutral beam injection and by gas puffing. The graphite limiter is subject to erosion and codeposition. While short term retention was high, the retention averaged over the 1993-1995 D-T campaign was 52% {+-} 15%. The tritium removal techniques resulted in lowering the in-vessel inventory from 16.4 kCi at the end of 1995 operation to 7.2 kCi at the start of the 1996 experimental program. 14 refs., 3 figs., 1 tab.

The major contributions to this workshop came from TORE-SUPRA, TPX, LHD and W7-X (two tokamaks, one heliotron and a stellarator). All four devices are of similar size and designed for a similar range of parameters and, in particular, they address the same target - steady-state plasma operation. TORE-SUPRA is a circular cross section tokamak, LHD is designed with optimized continuous coils and provides a helical divertor, TPX and W7-X incorporate strongly-shaped geometries to improve confinement and stability and optimize the bootstrap current. In TPX, the system is optimized for a large bootstrap current, in W7-X for basically no bootstrap current. Also the heating systems are designed for the specific purpose of steady-state operation. TORE-SUPRA is in operation, LHD is near completion; W7-X has recently (after the workshop) been approved; TPX is still in the approval phase. Superconducting coil material is NbTi for TORE-SUPRA, LHD and W7-X and Nb3Sn for TPX. The TF-coils of TORE-SUPRA (1.8 K, superfluid He) and the helical coil of LHD (4.2 K for the first part of experiments with 3 T) are bath-cooled. Cable-in-conduit conductors are used in the other SC coils. Largely different solutions are selected for the SC cable composition. Except for TPX, which has a double-null poloidal field divertor matched to the strong plasma shaping, most of the steady-state devices have a very flexible plasma edge configuration: islands or ergodic boundaries (W7-X, TORE-SUPRA, LHD), limiter (TORE-SUPRA), helical divertor (LHD, W7-X). The status of plasma equilibrium control, high heat flux materials and activation problems were also discussed. Power plant studies based on different steady-state confinement concepts were presented and confronted with a summary on the PULSAR study based on pulsed systems.

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 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...

Heating of the first wall of TFTR due to disruptions is investigated experimentally using an extensive array of thermocouples. By comparing results from discharges with and without disruptions, we extract effects due to the disruption alone. Disruptions preferentially heat the same areas which are heated during discharges without disruptions. Hot areas are inward protrusions or regions unshielded by neighboring areas. Peaking factors in the toroidal direction, defined as peak temperature divided by average toroidal temperature, as a function of poloidal angle, are calculated. For nondisruptive discharges, the peaking factor varies between 2 and 4. For the disruptive portion of a discharge only, the peaking factor near the midplane, where most of the energy is deposited, ranges from 3 to 5. Further away from the midplane, the peaking factor can reach 28, although the heat load is less in that region.

The poloidal and toroidal spatial distributions of D α , He I and C II emission have been obtained in the vicinity of the TFTR bumper limiter and are compared with models of ion flow to the surface. The distributions are found not to agree with a model (the “cosine” model) which determines the incident flux density using only the parallel fluxes in the scrape-off layer and the projected area of the surface perpendicular to the field lines. In particular, the cosine model is not able to explain the significant fluxes observed at locations on the surface which are oblique to the magnetic field. It is further shown that these fluxes cannot be explained by the finite Larmor radii of impinging ions. Finally, it is demonstrated, with the use of Monte Carlo codes, that the distributions can be explained by including both parallel and cross-field transport onto the limiter surface.

High poloidal beta (βp approximately 5) discharges produced with nearly balanced tangential neutral beam injection into low current discharges in the TFTR tokamak are described. As βp exceeds approximately 1.2 times the aspect ratio, a separatrix with an inside poloidal field null is observed to limit the outer boundary of the plasma. Since the curvature of TFTR's applied vertical field is constant, the appearance of the poloidal field null corresponds to the equilibrium poloidal beta limit

Under some conditions, a very large influx of carbon into TFTR occurs during neutral beam injection into low recycling plasmas (the supershot regime). These carbon 'blooms' result in serious degradation of plasma parameters. The sources of this carbon have been identified as hot spots on the TFTR bumper limiter at or near the last closed flux surface. Two separate temperature thresholds have been identified. One threshold, at about 1650°C, is consistent with radiation enhanced sublimation (RES). The other, at about 2300°C, appears to be thermal sublimation of carbon from the limiter. The carbon influx can be quantitatively accounted for by taking laboratory values for RES rates, making reasonable assumptions about the extent of the blooming area and assuming unity carbon recycling at the limiter. Such high carbon recycling is expected, and it is shown that, in target plasmas at least, it is observed on TFTR. The sources of the carbon blooms are sites which have either loosely attached fragments of limiter material (caused by damage) or surfaces that are nearly perpendicular to the magnetic field lines. Such surfaces may have local power depositions two orders of magnitude higher than usual. The TFTR team modified the limiter during the opening of winter 1989–1990. The modifications greatly reduced the number and magnitude of the blooms, so that they are no longer a problem.

Deuterium neutral beams with energies up to 110 keV were injected into TFTR (Tokamak Fusion Test Reactor) plasmas at low magnetic field such that the beam injection velocities were comparable to the Alfven velocity. Excitation of toroidal Alfven eigenmodes was observed by Mirnov coils and beam emission spectroscopy. 10 refs., 4 figs.

This paper addresses the processes during disruptions of plasmas in the Tokamak Fusion Test Reactor (TFTR), including precursor phenomena, and the resulting effects which are of concern in future devices, such as heat loads and runaways. The pre-disruptive behavior varies for disruptions due to operation at different operating limits such as (a) high density, (b) low q a , and (c) high poloidal beta β p . Power loading and runaway electron impact on the first wall of a tokamak are important issues in present research devices and in the design of future devices such as CIT and ITER. In particular, disruptions can deposit large amounts of power in localized areas which can then damage the wall or adversely affect operation. The time scales and spatial profiles of heat deposition on the first wall have been studied with a poloidal array of infrared detectors and an extensive array of thermocouples providing coverage of the entire axi-symmetric, primary power-handling limiter in TFTR — the bumper limiter. While there were often common hot spots, different types of disruptions caused different heating patterns and heat loads on the bumper limiter.

A plasma-enhanced CVD process was used to deposit thin (40–100 nm), amorphous carbon/boron films onto the first wall of TFTR. A series of plasma experiments was performed to test the effect of this first-wall modification on low-Z impurity (carbon, oxygen) behavior, suppression of metallic impurities, post-disruption recovery, hydrogenic recycling, and plasma density limits. The analysis of the plasma performance following the first few boronization attempts on TFTR shows: (1) suppression by a factor of 3-1.5 of carbon impurities immediately after boronization, although this suppression is short-lived (~10 shots); (2) long term suppression of oxygen and metallic impurities by a factor of two; (3) significant suppression of oxygen impurity emission following disruptions and during pulse discharge cleaning; and (4) no significant change in hydrogenic recycling properties in comparison to the pre-existing carbon first wall.

The TFTR (Tokamak Fusion Test Reactor) bumper limiter (BL) is an axisymmetric toroidal limiter mounted on the inner wall of the vacuum vessel. It subtends 120 degrees poloidally and has a surface area of 22 m/sup 2/. The plasma-facing surface consists of 1000 kg of graphite tiles mounted on water-cooled Inconel backing plates. During the initial installation in the Spring of 1985, the limiter surface was aligned to the toroidal magnetic field by mechanical and magnetic measurements to an estimated accuracy of +or-2 mm. The methods used to mechanically align the limiter are described, and evidence, based on machine operation with a plasma, that the limiter is reasonably well aligned with the toroidal field is presented. Future work addressing the alignment of the total limiter to the toroidal field using mechanical and magnetic measurements and the replacement of a subset of the carbon times with carbon-carbon composite material is also discussed. >

A project undertaken to measure the alignment of the TFTR (Tokamak Fusion Test Reactor) bumper limiter in relation to the toroidal magnetic field axis is discussed. The process involved the measurement of the toroidal magnetic field and the positions of the tiles that make up the bumper limiter. The basis for the instrument control and data acquisition system was LabVIEW. LabVIEW is a graphical programming system for developing scientific and engineering applications on a Macintosh. With LabVIEW, users create graphical software modules called virtual instruments instead of writing conventional text-based code. To measure the magnetic field, the control system acquired data from two nuclear magnetic resonance magnetometers while the toroidal field coils were pulsed. To measure the position of the tiles on the limiter, an instrumented mechanical arm was used inside the vessel. LabVIEW's graphical programming environment provided an easy transition from engineering-design block diagram to a VI block diagram, which made for speedy development. It also provided for quick and painless online systems integration, which proved to be essential to the timely completion of the project. >

To optimize power loading on the first wall components in the TFTR (Tokamak Fusion Test Reactor), parts must be carefully aligned with the toroidal magnetic field (TF) surfaces. First, the location of a template with six fixed positions was established with respect to the TF using measurements of magnetic field strength at several positions on the template. Next, a measuring arm was installed at each of these locations and manually positioned in order to resolve relative and absolute coordinates of the in-vessel components with respect to the TF. The measuring arm is a flexible linkage consisting of six links and five joints instrumented with absolute optical encoders. Prior to use in the vacuum vessel, the arm was calibrated to determine the best fit for the 23 unknowns involving link lengths, encoder zero positions, and shaft-to-link angles. The position of the indicator point with respect to the measuring arm base is known to +or-0.75 mm after this calibration process. During operation, the shaft encoder positions are processed by an algorithm to determine the relative and global coordinates of the indicator point for real-time use and stored for later detailed analysis. Data obtained with this device are shown. >

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.

Toroidal Alfven Eigenmodes (TAE) were excited by the energetic neutral beam ions tangentially injected into TFTR plasmas at low magnetic field such that the injection velocities were comparable to the Alfven speed. The modes were identified by measurements from Mirnov coils and beam emission spectroscopy (BES). TAE modes appear in bursts whose repetition rate increases with beam power. The neutron emission rate exhibits sawtooth-like behavior and the crashes always coincide with TAE bursts. This indicates ejection of fast ions from the plasma until these modes are stabilized. The dynamics of growth and stabilization was investigated at various plasma current and magnetic field. The results indicate that the instability can effectively clamp the number of energetic ions in the plasma. The observed instability threshold is discussed in the light of recent theories. In addition to these TAE modes, intermittent oscillations at three times the fundamental TAE frequency were observed by Mirnov coils, but no corresponding signal was found in BES. It appears that these high frequency oscillations do not have direct effect on the plasma neutron source strength.

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

A progress report on the experimental studies of vacuum vessel conditioning during the first year of TFTR operation is presented. A previous paper described the efforts expended to condition the TFTR vessel prior to and during the initial plasma start-up experiments. During the start-up phase, discharge cleaning was performed with the vessel at room temperature. For the second phase of TFTR operations, which was directed towards the optimization of ohmically-heated plasmas, the vacuum vessel could be heated to 150°C. The internal configuration of the TFTR vessel was more complex during the second phase with the addition of a TiC/C moveable limiter array, inconel bellows cover plates, and ZrAl getter pumps. A quantitative comparison is given on the effectiveness of vessel bakeout, glow discharge cleaning, and pulse discharge cleaning in terms of total quantity of removed carbon and oxygen, residual gas base pressures and the resulting plasma impurity levels as measured by visible, UV and soft X-ray spectroscopy. The initial experience with hydrogen isotope changeover in TFTR is presented including the results of the attempt to hasten the changeover-time by using a glow discharge to precondition the vessel with the new isotope.

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.

This paper reviews the experience with first-wall materials in TFTR over a 20 month period of operatiori during 1985–1987. Experience with the axisymmetric inner wall limiter, constructed of graphite tiles, is described, including the necessary conditioning procedures needed for impurity and particle control of high power (≤ 20 MW) neutral injection experiments. The thermal effects in disruptions have been quantified and no significant damage to the bumper limiter has occurred as a result of disruptions. Carbon and metal impurity redeposition effects have been quantified through surface analysis of wall samples. Estimates of the tritium retention in the graphite limiter tiles and redeposited carbon films have been made, based on analysis of deuterium retention in removed graphite tiles and wall samples. New limiter structures have been designed using a 2D carbon/carbon ( C C ) composite material for RF antenna protection. Laboratory tests of the important thermal, mechanical, and vacuum properties of C C materials are described. Finally, the last series of experiments in TFTR with in-situ Zr Al surface pumps are discussed. Problems with Zr Al embrittlement have led to the removal of the getter material from the in-torus environment.

A suprathermal electron beam is injected parallel to vector B in a low-..beta.. plasma. The sharp space-potential drop across the surface of the beam generates a strong cross-field current which in turn drives a modified two-stream instability at the lower hybrid frequency. Intense stochastic ion heating is observed with the onset of the instability, and the heating rate is found to be proportional to the wave energy.

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.

Cellulose films having radically different physical properties were prepared by the standard viscose process and by the deacetylation of commercial cellulose acetate film. These films were characterized by x-ray, acid hydrolysis, differential swelling in caustic, and light and electron microscopy of freshly torn edges. Measurements by x-ray and acid hydrolysis reveal that the film from cellulose acetate has less crystalline material and smaller crystallite size than the viscose film. By drfferential swelling in caustic the film from cellulose acetate was shown to have less highly ordered material than the viscose film. Light and electron microscopy of freshly torn edges of the two films reveal a com mon structural element. Fibrils were seen, the size of which did not vary appreciably between the two films. The arrangement of the fibrils in the internal structure of the films was markedly different. By considering the data obtained from these different sources, an explanation of the property difference and its relation to fine structure can be postulated.

The QED-1 device at PPL has provided gettering efficiency data for neutralized hydrogen plasma on titanium. The hollow-anode arcjet produces a plasma column 1 cm in diameter with 10/sup 12/ < n/sub e/ < 10/sup 15/ cm/sup -3/ and T/sub i/ approx.< T/sub e/ = 3-10 eV, confined by an axial magnetic field of 1-6 kG. The gettering measurements are based on monitoring neutral gas density with respect to time in the divertor simulation chamber of QED-1. The present results indicate that the plasma particles lose their charge and most of their energy when they strike the neutralizer plate.

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.

The first wall of the Tokamak Fusion Test Reactor (TFTR) consists of a water cooled toroidal belt limiter, a cooled moveable limiter, and cooled protective plates to shield the vacuum vessel from neutral beam shinethrough. Each of these systems consists of Inconel support plates covered with graphite tiles. In addition, there are Inconel and stainless steel bellows cover plates to protect the bellows and the surface pumping system which provides enhanced pumping in the torus and also serves to protect the bellows. These systems are described and the design requirements, simulations and actual thermal and mechanical loads reviewed. The normal and off-normal operating conditions which were considered in the design of the TFTR components include thermal loading during normal and disruptive plasma operation, eddy-current induced mechanical forces and arcing. The failures which have occurred are generally associated with, thermal stress rather than mechanical failure due to disruption induced eddy currents.

Experiments with plasmas having nearly equal concentrations of deuterium and tritium have been carried out on TFTR. To date, the maximum fusion power has been 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-{beta}{sub p} discharge following a current ramp-down. The fusion power density in the core of the plasma has reached 2.8 MWm{sup {minus}3}, exceeding that expected in the International Thermonuclear Experimental Reactor (ITTER). The energy confinement time, {tau}{sub E}, is observed to increase in D-T, relative to D plasmas, by 20% and the n{sub i}(O){center_dot}{tau}{sub E} 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-{beta}{sub p} discharges. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP simulations assuming classical confinement. Measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from helium gas puffing experiments. The loss of energetic alpha particles to a detector at the bottom ofmore » the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. ICRF heating of a D-T plasma, using the second harmonic of tritium, has been demonstrated. D-T experiments on TFTR will continue both to explore the physics underlying the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.« less