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The effect of isotope on confinement in high‐recycling, L‐mode plasmas is studied on the Tokamak Fusion Test Reactor (TFTR) [see D. M. Meade, J. Fusion Energy 7, 107 (1988)] by comparing hydrogen and deuterium plasmas with the same magnetic field and similar electron densities and heating power, with both Ohmic and deuterium‐neutral‐beam heating. Following a long operational period in deuterium, nominally hydrogen plasmas were created through hydrogen glow discharge and hydrogen gas puffing in Ohmic plasmas, which saturated the exposed limiter surface with hydrogen and raised the H/(H+D) ratio from 10±3% to 65±5%. Ohmic deuterium discharges obtained higher stored energy and lower loop voltage than hydrogen discharges with similar limiter conditions. Neutral‐beam power scans were conducted in L‐mode plasmas at minor radii of 50 and 80 cm, with plasma currents of 0.7 and 1.4 MA. To minimize transport differences from the beam deposition profile and beam heating, deuterium neutral beams were used to heat the...

The TRANSP code [R. V. Budny et al., Nucl. Fusion 35, 1497 (1995)] is used to construct comprehensive, self‐consistent models for plasmas within the separatrix surface in the International Thermonuclear Experimental Reactor (ITER) [Technical Basis for the ITER Interim Design Report, Cost Review and Safety Analysis (International Atomic Energy Agency, Vienna, 1996)]. Steady state profiles of two plasmas from the ITER ‘‘Interim Design’’ database are used. Effects of 1 MeV 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 the alpha particle heating of electrons and thermal ions, and the thermalization rates. The modeling indicates that the deposition of the neutral beam ions will peak in the plasma center, and the average beam ion energy will be half the injected energy. Sawtooth mixing will broaden the fast alpha profile. The toroidal ripple loss rate of alpha energy...

An experiment was performed in TFTR DT plasmas in an attempt to destabilize the alpha particle driven toroidicity-induced Alfven eigenmodes (TAEs) by reducing their thermal ion Landau damping. The thermal ion Landau damping was reduced by transiently lowering the ion temperature using either helium (He) gas puffs or deuterium (D) or lithium (Li) pellet injection during neutral beam injection (NBI) into DT supershots. The ion temperature was successfully lowered from Ti(0) approximately=20 keV to Ti(0) approximately=10 keV in about 0.2 s; however, no alpha driven TAEs were observed. Theoretical analyses of the TAE instability of these DT discharges indicate that the alpha pressure required for TAE stability still remained greater than that actually obtained in this experiment, mainly because of the effects of beam ion Landau damping

Measurements of DT alpha particle loss to the outer midplane region of TFTR have been made using a radially movable scintillator detector. The conclusion from this data is that mechanisms determining the DT alpha loss to the outer midplane are not substantially different from those for DD fusion products. Some of these results are compared with a simplified theoretical model for TF ripple induced alpha loss, which is expected to be the dominant classical alpha loss mechanism near the outer midplane. An example is given of plasma driven MHD induced alpha particle loss to the outer midplane, but no sign of any `collective` alpha instability induced alpha particle loss has yet been observed at this location

Experiments with plasmas having nearly equal concentrations of deuterium and tritium have been carried out on TFTR. To date (September 1995), 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 P discharge following a current ramp-down. The fusion power density in the core of the plasma has reached 2.8 MW/m3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER). The energy confinement time tau E is observed to increase in DT, relative to D plasmas, by 20% and the n1(0).T1(0). tau 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 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 of 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 DT plasma, using the second harmonic of tritium, has been demonstrated. DT 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

Modelling of TF ripple loss of alphas in DT experiments on TFTR now includes neoclassical calculations of first orbit loss, stochastic ripple diffusion, ripple trapping and collisional effects. A rapid way to simulate experiment has been developed which uses a simple stochastic domain model for TF ripple loss within the TRANSP analysis code, with the ripple diffusion threshold evaluated by comparison with more accurate but computationally expensive Hamiltonian co-ordinate guiding centre code simulations. Typical TF collisional ripple loss predictions are 6-10% loss of alphas for TFTR DT experiments at Ip=1.0-2.0 MA and R=2.52 m

A TFTR supershot with a plasma current of 2.5 MA, a neutral beam heating power of 33.7 MW and a peak DT fusion power of 7.5 MW is studied using the TRANSP plasma analysis code. Simulations of alpha parameters such as the alpha heating, pressure and distributions in energy and v1/v are given. The effects of toroidal ripple and mixing of the fast alpha particles during the sawteeth observed after the neutral beam injection phase are modelled. The distributions of alpha particles on the outer midplane are peaked near forward and backward v1/v. Ripple losses deplete the distributions in the vicinity of v1/v=-0.2. Sawtooth mixing of fast alpha particles is computed to reduce their central density and broaden their width in energy

Predictions for ripple loss of fast ions from TFTR are investigated with a guiding centre code including both collisional and ripple effects. A synergistic enhancement of fast ion diffusion is found for toroidal field ripple with collisions. The total loss is calculated to be roughly twice the sum of ripple and collisional losses calculated separately. Discrepancies between measurements and calculations of plasma beta at low current and large major radius are resolved when both effects are included for neutral beam ions. A 20 to 30% reduction in alpha particle heating is predicted for qa=6-14, R=2.6 m DT plasmas on TFTR owing to first orbit and collisional stochastic ripple diffusion

Alpha particle loss was measured during the TFTR DT experiments with a scintillator detector located at the vessel bottom in the ion Del B drift direction. The DT alpha particle loss to this detector was consistent with the calculated first orbit loss over the whole range of plasma current I=0.6-2.7 MA. In particular, the alpha particle loss rate per DT neutron at a given plasma current did not increase significantly with fusion power up to 10.7 MW, indicating the absence of any new 'collective' alpha particle loss processes in these experiments

High or enhanced confinement (H‐mode) plasmas have been obtained for the first time with nearly equal concentrations of deuterium and tritium in high‐temperature, high poloidal beta plasmas in the Tokamak Fusion Test Reactor (TFTR) [McGuire, Phys. Plasmas 2, 2176 (1995)]. Tritium fueling was provided mainly through high‐power neutral beam injection (NBI) with powers up to 31 MW and beam energies of 90–110 keV. A transition to a circular limiter H‐mode configuration has been obtained, following a programmed rapid decrease of the plasma current. Isotope effects, due to the presence of tritium, led to different behavior for deuterium–deuterium (DD) and deuterium–tritium (DT) H‐modes relative to confinement, edge localized magnetohydrodynamic modes (ELMs), and ELM effects on fusion products. However, the threshold power for the H‐mode transition was the same in DD and DT. Some of the highest values of the global energy confinement time, τE, have been achieved on TFTR during the ELM‐free phase of DT H‐mode plasmas. Enhancements of τE greater than four times the L‐mode have been attained.

The confinement and heating of supershot plasmas are significantly enhanced with tritium beam injection relative to deuterium injection in the Tokamak Fusion Test Reactor [Plasma Phys. Controlled Fusion 26, 11 (1984)]. The global energy confinement and local thermal transport are analyzed for deuterium and tritium fueled plasmas to quantify their dependence on the average mass of the hydrogenic ions. Radial profiles of the deuterium and tritium densities are determined from the D–T fusion neutron emission profile. The inferred scalings with average isotopic mass are quite strong, with τE∝〈A〉0.85±0.20, τEthermal∝〈A〉0.89±0.20, χitot∝〈A〉−2.6±0.5, and De∝〈A〉−1.4±0.2 at fixed Pinj. For fixed local plasma parameters χitot∝〈A〉−1.8±0.4 is obtained. The quoted 2σ uncertainties include contributions from both diagnostic errors and shot irreproducibility, and are conservatively constructed to attribute the entire scatter in the regressed parameters to uncertainties in the exponent on plasma mass.

TFTR has begun its campaign to study deuterium-tritium fusion under reactor-like conditions. Variable amounts of deuterium and tritium neutral beam power have been used to maximize fusion power, study alpha heating, investigate alpha particle confinement, and search for alpha driven plasma instabilities. Additional areas of study include energy and particle transport and confinement, ICRF heating schemes for DT plasmas, tritium retention, and fusion in high βp plasmas. The majority of this work is done in the TFTR supershot confinement regime. To obtain supershots, extensive limiter conditioning using helium fueled ohmic discharges and lithium pellet injection into ohmic and neutral beam heated plasmas is performed, resulting in a low recycling limiter. The relationship between recycling and core plasma confinement has been studied by using helium, deuterium and high-Z gas puffs to simulate high recycling limiter conditions. These studies show that confinement in TFTR supershots is very sensitive to the influx of neutral particles at the plasma edge.

Both global and thermal energy confinement improve in high-temperature supershot plasmas in the Tokamak Fusion Test Reactor (TFTR) when deuterium beam heating is partially or wholly replaced by tritium beam heating. For the same heating power, the tritium-rich plasmas obtain up to 22% higher total energy, 30% higher thermal ion energy, and 20-25% higher central ion temperature. Kinetic analysis of the temperature and density profiles indicates a favorable isotopic scaling of ion heat transport and electron particle transport, with τEi(a/2) ∝ A0.7-0.8 and τpe(a) ∝ A0.8.

An analytic model is developed to estimate the classical pitch angle scattering loss of energetic fusion product ions into prompt loss orbits in a tokamak geometry. The result is applied to alpha particles produced by deutrium–tritium fusion reactions in a plasma condition relevant to Tokamak Fusion Test Reactor (TFTR) [Proceedings of the 14th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Wurzburg, 1992 (International Atomic Energy Agency Vienna, 1992)]. A poloidal angular distribution of collisional fast ion loss at the first wall is obtained and the numerical result from the transp [‘‘A standard D–T supershot simulation’’ (to be published in Nucl. Fusion, 1994)] code is discussed. The present model includes the effect that the prompt loss boundary moves away from the slowing‐down path due to reduction in banana thickness, which enables one to understand, for the first time, the dependence of the collisional loss rate on Zeff.

Peak fusion power production of 6.2 ± 0.4 MW has been achieved in TFTR plasmas heated by deuterium and tritium neutral beams at a total power of 29.5 MW. These plasmas have an inferred central fusion alpha particle density of 1.2 x 1017 m ₋3 without the appearance of either disruptive magnetohydrodynamics events or detectable changes in Alfven wave activity. The measured loss rate of energetic alpha particles agreed with the approximately 5% losses expected from alpha particles which are born on unconfined orbits.

In the Tokamak Fusion Test Reactor (TFTR) [International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Wurzburg, Paper No. A‐2‐2 (International Atomic Energy Agency, Vienna, 1993)] there have been at least three types of anomalous loss of alpha‐like deuterium–deuterium (D–D) fusion products: (1) a magnetohydrodynamic (MHD)‐induced loss of D–D fusion products correlated with Mirnov and fishbone‐type oscillations and sawtooth crashes, (2) a slow ‘‘delayed’’ loss of partially thermalized D–D fusion products occurring without large MHD activity, and (3) ion cyclotron resonance heating (ICRH)‐induced loss of D–D fusion products ions observed during direct electron heating experiments, and possibly also during 3He minority heating. In this paper each of these will be reviewed, concentrating on those due to MHD activity, which are the largest of these anomalous losses. The experimental results are compared with numerical models of various fusion product transport mechanisms.

The increased core electron temperature produced by ICRF heating of TFTR, D-T neutral-beam-heated supershot plasmas is expected to extend the alpha-particle slowing down time and hence enhance the central alpha-particle pressure. In preparation for the TFTR D-T operational phase, which started in late 1993, a series of experiments were conducted on TFTR to explore the effect of ICRF heating on the performance and stability of low-recycling, deuterium supershot plasmas in the 3He minority heating regime. The coupling of up to 7.4 MW of 47 MHz ICRF power to full size (R approximately 2.62 m, a approximately 0.96 m), 3He minority, deuterium supershots heated with up to 30 MW of deuterium neutral beam injection has resulted in a significant increase in core electron temperature ( Delta Te=3-4 keV). Simulations of equivalent D-T supershots predict that such ICRF heating should result in approximately a 60% increase in the alpha-particle slowing down time and an enhancement of about 30% in the central alpha pressure. Future experiments to be conducted at ICRF powers up to 12.5 MW during the upcoming TFTR D-T campaign may result in even greater enhancements in core alpha parameters. This paper presents results from experiments performed at an axial toroidal magnetic field of about 4.8 T, where the minority resonance was within 0.1-0.15 m of the plasma core. Combined ICRF and neutral beam heating powers in these experiments reached TFTR record levels of over 37 MW, which allowed an exploration of the power loading limits on the carbon limiter tiles. The plasma current was operated at 1.85 and 2.2 MA and sawtooth suppression was observed at the higher plasma current.

A new method for obtaining a transient ('pulse') electron heat diffusivity (χep) in the radial region 0.38 < r/a < 0.56 in TFTR L mode discharges is presented. Small electron temperature perturbations were caused by single bursts of injected impurities which radiated and cooled the plasma edge. A case of iron injection by laser ablation was found to be more definitive than a supporting helium gas puff case. In this new 'cold pulse' method, the authors concentrate on modelling just the electron temperature perturbations, tracked with electron cyclotron emission diagnostics, and on being able to justify separation of the perturbations in space and time from the cooling source. This χep is obtained for these two cases to be χep = (6.0 m2/s ± 35%) ~ 4χe (power balance), which is consistent with but more definitive than results from other studies that are more susceptible to ambiguities in the source profile

Using perturbations in electron density and temperature induced by small helium gas puffs in TFTR (Tokamak Fusion Test Reactor) [Plasma Phys. Controlled Nucl. Fusion Res. 1, 51 (1986)], the dominance of convective heat transport in the core (r/a

Krypton has been injected into ohmically-heated TFTR plasmas with peak electron temperatures of 6 key to study the effects of krypton on the plasma performance and to investigate the emitted krypton line radiation, which is of interest for future-generation tokamaks such as ITER, both as a diagnostic of the central ion temperature and for the control of energy release from the plasma by radiative cooling. The emitted radiation was monitored with a bolometer array, an X-ray pulse height analysis system, and a high-resolution Johann-type crystal spectrometer; and it was found to depend very sensitively on the electron temperature profile. Satellite spectra of heliumlike krypton, KrXXXV, near 0.95 [Angstrom] including lithiumlike, berylliumlike and boronlike features were recorded in second order Bragg reflection. Radiative cooling and reduced particle recycling at the plasma edge region were observed as a result of the krypton injection for all investigated discharges. The observations are in reasonable agreement with modeling calculations of the krypton ion charge state distribution including radial transport.

The complete ion cyclotron range of frequency (ICRF) heating system for the Tokamak Fusion Test Reactor (TFTR) [Fusion Tech. 21, 1324 (1992)], consisting of four antennas and six generators designed to deliver 12.5 MW to the TFTR plasma, has now been installed. Recently a series of experiments has been conducted to explore the effect of ICRF heating on the performance of low recycling, supershot plasmas in minority and nonresonant electron heating regimes. The addition of up to 7.4 MW of ICRF power to full size (R∼2.6 m, a∼0.95 m), helium‐3 minority, deuterium supershots heated with up to 30 MW of deuterium neutral‐beam injection has resulted in a significant increase in core electron temperature (ΔTe=3–4 keV). Simulations of equivalent deuterium–tritium (D–T) supershots predict that such ICRF heating should result in an increase in βα(0)∼30%. Direct electron heating has been observed and has been found to be in agreement with theory. The ICRF heating has also been coupled to neutral‐beam heated plasmas f...

General plasma physics principles state that power flow Q(r) through a magnetic surface in a tokamak should scale as Q(r)= {32π2Rr3Te2c nea/[eB (a2−r2)2]} F(ρ*,β,ν*,r/a,q,s,r/R,...) where the arguments of F are local, nondimensional plasma parameters and nondimensional gradients. This paper reports an experimental determination of how F varies with normalized gyroradius ρ*≡(2TeMi)1/2c/eBa and collisionality ν*≡(R/r)3/2qRνe(me/ 2Te)1/2 for discharges prepared so that other nondimensional parameters remain close to constant. Tokamak Fusion Test Reactor (TFTR) [D. M. Meade et al., in Plasma Physics and Controlled Nuclear Fusion Research, 1990, Proceedings of the 13th International Conference, Washington (International Atomic Energy Agency, Vienna, 1991), Vol. 1, p. 9] L‐mode data show F to be independent of ρ* and numerically small, corresponding to Bohm scaling with a small multiplicative constant. By contrast, most theories predict gyro‐Bohm scaling: F∝ρ*. Bohm scaling implies that the largest scale size f...

Studies of high density disruptions on TFTR, including a comparison of minor and major disruptions at high density, provide important new information regarding the nature of the disruption mechanism. Further, for the first time, an (m,n)=(1,1) 'cold bubble' precursor to high density disruptions has been experimentally observed in the electron temperature profile. The precursor to major disruptions resembles the 'vacuum bubble' model of disruptions first proposed by B.B. Kadomtsev and O.P. Pogutse (Sov. Phys. JETP 38 (1974) 283)

Propagation of the heat pulse arising from a sawtooth crash was studied in Ohmically heated plasmas as a function of density, and in neutral beam heated plasmas with centrally peaked and hollow power deposition profiles at two densities. The sawtooth period became longer with centre heating than with edge heating and with co-injection than with counter-injection. The lengthening of the sawtooth period was consistent with the reduction of the core loop voltage due to heating and noninductive current drive. In Ohmic plasmas, the electron thermal diffusivity χeHPP determined from a time-to-peak analysis decreased inversely proportionally to density for ne 2 × 1019 m-3

Sequential injection of multiple deuterium pellets combined with high power neutral beam heating in TFTR has produced stable plasmas with very high central densities, up to 5×1020 m-3. The line averaged densities achieved in these experiments correspond to Murakami parameters neR/BT up to 12×1019 m-2.T-1. This value, which does not appear to represent an intrinsic density limit, surpasses the values previously obtained in TFTR and exceeds, by more than a factor of two, the density limit predicted by the scaling of Greenwald et al. (Nuclear Fusion 28 (1988) 2199). The highest densities obtained with both pellet and gas fuelling have been achieved since boronization was applied to the first wall in TFTR. The characteristics and energy balance of the highest density plasma are discussed

Results obtained from experiments utilizing high‐power ion cyclotron range of frequencies (ICRF) heating to stabilize sawtooth oscillations on Tokamak Fusion Test Reactor (TFTR) [Hawryluk et al., Plasma Phys. Controlled Fusion 33, 1509 (1991)] 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±0.15 after a sawtooth‐stabilized period of ∼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 magnetohydrodynamic (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.

The role of the boundary plasma in determining the power and particle balance of tokamak discharges is discussed. Detailed boundary plasma measurements of edge density, edge temperature, deuterium influx and carbon influx are reported from ohmically heated TFTR discharges over a range of plasma densities. The experimental results are compared with predictions from a simple zero-dimensional model based on power and particle balance. Reasonable agreement is obtained. More detailed impurity modelling is performed with the LIM impurity production and transport code. The complementary modelling approaches reveal, amongst other results, the important role of the sputtering yield at the limiter in determining the central effective charge of the discharge, the ability of a densified boundary plasma to screen impurities from the central plasma and the importance of cross-field particle transport to the TFTR limiter. The authors demonstrate that, under the conditions of this experiment, the behaviour of the boundary plasma and its effect on the central plasma appear to be explicable using rather simple considerations of power and particle balance

Plasma-size scans have been performed in the Tokamak Fusion Test Reactor to investigate the effect of major and minor radius upon energy confinement in high-recycling plasmas heated by neutral-beam injection. The major radius, minor radius, and aspect ratio were varied over a large range ({ital a}=0.4--0.9 m, {ital R}=2.08--3.2 m, {ital R}/{ital a}=2.8--8.0). The energy confinement measured with diamagnetic and kinetic diagnostics scales strongly with major radius, and weakly or negatively with minor radius.

MHD activity at the plasma edge is studied during limiter H-modes. The ELMs are observed as large drops in the edge density and temperature, and are also clearly observed on the Mirnov coils and soft X-ray signals, during the H-mode phase. The ELMs are found to have a precursor instability dominated by an m/n = 1/0 mode at ~100–500 kHz. Simultaneously, a burst of hot electrons is observed to originate about 20 cm inside the plasma and propagate to the plasma edge. The burst of Dα emission may be triggered by the arrival of the hot electrons at the limiter surface.

Measurements of the toroidal rotation speed vφ(r) driven by neutral beam injection in tokamak plasmas and, in particular, simultaneous profile measurements of vφ, Ti, Te, and ne, have provided new insights into the nature of anomalous transport in tokamaks. Low‐recycling plasmas heated with unidirectional neutral beam injection exhibit a strong correlation among the local diffusivities, χφ≊χi>χe. Recent measurements have confirmed similar behavior in broad‐density L‐mode plasmas. These results are consistent with the conjecture that electrostatic turbulence is the dominant transport mechanism in the tokamak fusion test reactor tokamak (TFTR) [Phys. Rev. Lett. 58, 1004 (1987)], and are inconsistent with predictions both from test‐particle models of strong magnetic turbulence and from ripple transport. Toroidal rotation speed measurements in peaked‐density TFTR ‘‘supershots’’ with partially unbalanced beam injection indicate that momentum transport decreases as the density profile becomes more peaked. In hi...

Magnetohydrodynamic (MHD) activity within three zones (core, half‐radius, and edge) of TFTR [Plasma Physics and Controlled Nuclear Fusion Research 1986 (IAEA, Vienna, 1987), Vol. 1, p. 51] tokamak plasmas are discussed. Near the core of the plasma column, sawteeth are often observed. Two types of sawteeth are studied in detail; one with complete, and the other with incomplete, magnetic reconnection. Their characteristics are determined by the shape of the q profile. Near the half‐radius the m/n=3/2 and 2/1 resistive ballooning modes are found to correlate with a beta collapse. The pressure and the pressure gradient at the mode rational surface are found to play an important role in stability. MHD activity is also studied at the plasma edge during limiter H modes. The edge localized modes (ELM’s) are found to have a precursor mode with a frequency between 50–200 kHz and a mode number m/n=1/0. The mode does not show a ballooning structure. While these instabilities have been studied on many other machines, ...

At the beginning of the 1990 Tokamak Fusion Test Reactor (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 105 low current (∼20 kA) Taylor discharge cleaning (TDC) pulses were required before high current (∼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 to 250 °C for a period of several hours. Heating the limiters is important to in...

(1) For shots 18528 and 20345 the limiter is believed to have had a high recycling coefficient; shot 21790 occurred after modest limiter 'conditioning', which produced some reduction in recycling, though much less than for 'supershots'; shot 22871 is a 'supershot'. These plasma regimes have been described previously [1, 2]. The older work does not include the 1987 corrections to the TVTS [2] electron temperatures (and Te dependent quantities) and so those data cannot be directly compared with the corrected data in the present tables.

Measurements of DT alpha particle loss to the outer midplane region of TFTR have been made using a radially movable scintillator detector. The conclusion from this data is that mechanisms determining the DT alpha loss to the outer midplane are not substantially different from those for DD fusion products. Some of these results are compared with a simplified theoretical model for TF ripple-induced alpha loss, which is expected to be the dominant classical alpha loss mechanism near the outer midplane. An example of plasma-driven MHD-induced alpha particle loss is shown, but no signs of any ``collective`` alpha instability-induced alpha loss have yet been observed.

Predictions for ripple loss of fast ions from TFTR are investigated with a guiding center code including both collisional and ripple effects. A synergistic enhancement of fast ion diffusion is found for toroidal field ripple with collisions. The total loss is calculated to be roughly twice the sum of ripple and collisional losses calculated separately. Discrepancies between measurements and calculations of plasma beta at low current and large major radius are resolved when both effects are included for neutral beam ions. A 20--30% reduction in alpha particle heating is predicted for q{sub a} = 6--14, R = 2.6 m DT plasmas on TFTR due to first orbit and collisional stochastic ripple diffusion.

A peak fusion power production of 9.3±0.7 MW has been achieved on the Tokamak Fusion Test Reactor (TFTR) in deuterium plasmas heated by co and counter injected deuterium and tritium neutral beams with a total power of 33.7 MW. The ratio of fusion power output to heating power input is 0.27. At the time of the highest neutron flux the plasma conditions are: Te(0)=11.5 keV, Ti(0)=44 keV, ne(0)=8.5×1019 m−3, and 〈Zeff〉=2.2 giving τE=0.24 s. These conditions are similar to those found in the highest confinement deuterium plasmas. The measured D‐T neutron yield is within 7% of computer code estimates based on profile measurements and within experimental uncertainties. These plasmas have an inferred central fusion alpha fraction of 0.2% and central fusion power density of 2 MW/m3 similar to that expected in a fusion reactor. Even though the alpha velocity exceeds the Alfven velocity throughout the time of high neutron output in most high power plasmas, MHD activity is not substantially different from that in co...

Three campaigns, prior to July 1994, attempted to increase the fusion power in DT plasmas on the Tokamak Fusion Test Reactor (TFTR). The first campaign was dedicated to obtaining >5 MW of fusion power while avoiding MHD events similar to the JET X-event. The second was aimed at producing maximum fusion power irrespective of proximity to MHD limits, and achieved 9 MW limited by a disruption. The third campaign increased the energy confinement time using lithium pellet conditioning while raising the ratio of alpha heating to beam heating.

The Tokamak Fusion Test Reactor (TFTR) has performed initial high-power experiments with the plasma fueled by deuterium and tritium to nominally equal densities. Compared to pure deuterium plasmas, the energy stored in the electron and ions increased by ~20%. These increases indicate improvements in confinement associated with the use of tritium and possibly heating of electrons by α-particles.

A review of measurements of ion heat transport in Ohmic and beam‐heated plasmas is given. A study of ion heat transport in Ohmic TFTR plasma is also reported. (AIP)

This paper reviews the recent progress made toward understanding the MHD-induced loss of D-D fusion products which has been seen on TFTR since 1988. These measurements have been made using the ``lost alpha`` diagnostic, which is described briefly. The largest MHD- induced loss occurs with coherent 3/2 or 2/1 MHD activity (kink/tearing modes), which can cause up to {approx}3--5 times the first-orbit loss at I{approx}1.6--1.8 MA, roughly a {approx}20--30% global los of D-D fusion products. Modeling of these MHD-induced losses has progressed to the point where the basic loss mechanism can be accounted for qualitatively, but the experimental results can not yet be understood quantitatively. Several alpha loss codes are being developed to improve the quantitative comparison between experiment and theory.

A review of TFTR plasma transport studies is presented. Parallel transport and the confinement of suprathermal ions are found to be relatively well described by theory. Cross-field transport of the thermal plasma, however, is anomalous with the momentum diffusivity being comparable to the ion thermal diffusivity and larger than the electron thermal diffusivity in neutral beam heated discharges. Perturbative experiments have studied nonlinear dependencies in the transport coefficients and examined the role of possible nonlocal phenomena. The underlying turbulence has been studied using microwave scattering, beam emission spectroscopy and microwave reflectometry over a much broader range in k perpendicular to than previously possible. Results indicate the existence of large-wavelength fluctuations correlated with enhanced transport.

Limiter H-modes with centrally peaked density profiles have been obtained in TFTR using a highly conditioned graphite limiter. The transition to these centrally peaked H-modes takes place from the supershot to the H-mode rather than the usual L- to H-mode transition observed on other tokamaks. Bidirectional beam heating is required to induce the transition. Density peaking factors, n{sub e}(0)/, greater than 2.3 are obtained and at the same time the H-mode characteristics are similar to those of limiter H-modes on other tokamaks, while the global confinement, {tau}{sub E}, can be >2.5 times L-mode scaling. The transport analysis of the data shows that transport in these H-modes is similar to that of supershots within the inner 0.6 m core of the plasma, but the stored electron energy (calculated using measured values of T{sub e} and n{sub e}) is higher for the H-mode at the plasma edge. Microwave scattering data for the edge plasma shows broad spectra at k = 5.5 cm{sup {minus}1} which begin at the drop in D{sub {alpha}} radiation and are strongly shifted in the electron diamagnetic drift direction. At the same time, beam emission spectroscopy (BES) shows a coherent mode near the boundary which propagates in the ionmore » direction with m = 15--20 at 20--30 kHz. During the ELM event these apparent rotations cease and Mirnov fluctuations in the frequency range of 50--500 kHz increase in intensity. 16 refs., 8 figs.« less

Circular limiter H-modes are obtained on TFTR during high power neutral beam heating. The transition is usually from the supershot to the H-mode rather than the usual L- to H- transition, and thus is obtained in a low recycling environment with core fueling mainly from the heating beams. As a result, the density and pressure profiles are highly peaked at the center. Global confinement time, {tau}{sub E}, is enhanced over L-mode scaling by up to {approx} 2.5 times. The onset of ELMs shortly after the H-mode transition appears to limit {tau}{sub E}. Limiter H-modes of up to 1.5 sec duration have been realized. 18 refs., 4 figs.

Simultaneous profile measurements of the toroidal rotation speed and ion temperature during unbalanced neutral beam injection in the Tokamak Fusion Test Reactor show that the ion momentum and thermal diffusivities are comparable in magnitude (${\mathrm{\ensuremath{\chi}}}_{\mathrm{\ensuremath{\varphi}}}$\ensuremath{\approxeq}1.5${\mathrm{\ensuremath{\chi}}}_{\mathrm{i}}$) and vary similarly with plasma current and minor radius. The correlation of ${\mathrm{\ensuremath{\chi}}}_{\mathrm{\ensuremath{\varphi}}}$ and ${\mathrm{\ensuremath{\chi}}}_{\mathrm{i}}$ is consistent with anomalous transport driven by collisionless electrostatic microinstabilities including ion-temperature-gradient-driven modes (${\mathrm{\ensuremath{\eta}}}_{\mathrm{i}}$ modes) and collisionless trapped-electron modes.

It has been found necessary to perform a series of first-wall conditioning steps prior to successful high power plasma operation in the Tokamak Fusion Test Reactor (TFTR). This series begins with glow discharge cleaning (GDC) and is followed by pulse discharge cleaning (PDC). During machine conditioning, the production of impurities is monitored by a Residual Gas Analyzer (RGA). PDC is made in two distinct modes: (1) Taylor discharge cleaning (TDC), where the plasma current is kept low (15--50 kA) and of short duration (50 ms) by means of a relatively high prefill pressure, and (2) aggressive PDC, where lower prefill pressure and higher toroidal field result in higher current (200--400 kA) limited by disruptions at q(a){approx}3 at {approx}250 ms. At a constant repetition rate of 12 discharges/minute, the production rate of H{sub 2}O, CO, or other impurities has been found to be an unreliable measure of progress in cleaning. However, the ability to produce aggressive PDC with substantial limiter heating, but without the production of X-rays from runaway electrons, is an indication that TDC is no longer necessary after {approx}10{sup 5} pulses. During aggressive PDC, the uncooled limiters are heated by the plasma from the bakeout temperature of 150 {degree}Cmore » to about 250 {degree}C over a period of three to eight hours. This limiter heating is important to enhance the rate at which H{sub 2}O is removed from the graphite limiter.« less

We have investigated trends in the conditioning of the Tokamak Fusion Test Reactor (TFTR) vacuum vessel during the May 1984 to April 1985 run period. The initial conditioning of the vessel, consisting of glow discharge cleaning (GDC) and pulse discharge cleaning (PDC) in concert with a 150°C vessel bakeout, is necessary to assure plasma operation after atmospheric venting. A long term conditioning process, ascribed to limiter conditioning, effectively improves operational conditions during the course of the run. Over several thousand high power plasma discharges, the improvement was documented by using standard parameter (fiducial) plasma discharges. Several techniques demonstrated short term improvements in vessel conditioning during this time period, including: Cr gettering and programming the plasma position relative to the limiter contact area.

Impurity transport has been studied in ohmic- and neutral-beam-heated TFTR discharges by observation and numerical modeling of time evolutions of emissions from germanium injected using the laser-blowoff method. VUV spectral lines, the total radiated power, and soft x-ray intensities are compared with code predictions in which the impurity flux is assumed to have diffusive and convective terms. In plasmas with I/sub p/=1.4 MA and anti n/sub e/approx. = 2.5 x 10/sup 13/ cm/sup -3/, little difference in transport between ohmic and beam-heated discharges is observed. At I/sub p/=0.7 MA and anti n/sub e/approx. = 1.0 x 10/sup 13/ cm/sup -3/, a three-fold increase in the diffusion coefficient is found during beam heating.