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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 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...

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%.

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.