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Boronization of NSTX has allowed access to higher density higher confinement plasmas. A glow discharge with 4 mtorr helium and 10% deuterated trimethyl boron deposited 1.7 g of boron on the plasma facing surfaces. Ion beam analysis of witness coupons showed a B + C areal density of 1018 cm-2 corresponding to a film thickness of 100 nm. Subsequent ohmic discharges showed oxygen emission lines reduced by a factor of 15, carbon emission reduced by a factor of two and copper reduced to undetectable levels. After boronization, the plasma current flat-top time increased by 70% enabling access to higher density higher confinement plasmas.

The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current Ip was successfully brought up to the design value of 1 MA on 14 December 1999. The planned plasma shaping parameters, elongation κ = 1.6-2.2 and triangularity δ = 0.2-0.4, were achieved in inner wall limited, and single null and double null diverted configurations. The coaxial helicity injection (CHI) and high harmonic fast wave (HHFW) experiments were also initiated. CHI current of 27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With the injection of 2.3 MW of HHFW power, using 12 antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5 × 1013 cm-3, increasing the plasma energy to 59 kJ and the toroidal β, βT, to 10%. The NBI system commenced operation in September 2000. The initial results with two ion sources (PNBI = 2.8 MW) show good heating, producing a total plasma stored energy of 90 kJ corresponding to βT ≈ 18% at a plasma current of 1.1 MA.

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.

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...

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.

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.

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.

The vacuum pumping requirements for a fusion device are dictated by a number of factors, including the materials used in construction of the device, the cleaning and conditioning techniques implemented, the operating conditions and device temperature, the plasma characteristics, and the fuel gases and impurities retained and released by the plasma-facing components (PFCs). In an attempt to derive guidelines for determining the vacuum pumping requirements of a generic fusion device, a study was undertaken to examine the vacuum pumping capabilities of existing large fusion devices, to review the cleaning and conditioning techniques now in use, and to catalog pertinent vacuum equipment that is now available or anticipated soon. In a survey of six large fusion devices [ASDEX, DIII-D, the Joint European Torus (JET), JT-60, the Tokamak Fusion Test Reactor (TFTR), Tore Supra], information was collected on PFC materials, cleaning and conditioning techniques, device operating temperatures, and vacuum pumping system characteristics and on the surface conditions and partial pressures necessary to achieve high-purity plasma discharges in each device. The results of the survey are reviewed to determine how the various factors affect vacuum pumping requirements. The use of tritium in future fusion devices is expected to create a new set of problems for the vacuum pumping systems of these devices. These problems are addressed, and some possible solutions for the pumping and gas handling systems are identified. New vacuum equipment is under development, and the fusion community is in a position to influence its direction and emphasis. Information gained in studies of this type should be brought to bear on the development process to ensure that vacuum equipment will be available to meet the needs of future fusion devices.

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.

In the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Fusion 26, 11 (1984)], the highest neutron source strength Sn and D–D fusion power gain QDD are realized in the neutral‐beam‐fueled and heated ‘‘supershot’’ regime that occurs after extensive wall conditioning to minimize recycling. For the best supershots, Sn increases approximately as P1.8b. The highest‐Q shots are characterized by high Te (up to 12 keV), Ti (up to 34 keV), and stored energy (up to 4.7 MJ), highly peaked density profiles, broad Te profiles, and lower Zeff. Replacement of critical areas of the graphite limiter tiles with carbon‐fiber composite tiles and improved alignment with the plasma have mitigated the ‘‘carbon bloom.’’ Wall conditioning by lithium pellet injection prior to the beam pulse reduces carbon influx and particle recycling. Empirically, QDD increases with decreasing pre‐injection carbon radiation, and increases strongly with density peakedness [ne(0)/〈ne〉] during the beam pulse. To date, the best fusion resu...

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.

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, ...

Boronization on the National Spherical Torus Experiment (NSTX) has proved to be quite beneficial with increases in confinement and density, and decreases in impurities observed in the plasma. The boron has been applied to the interior surfaces of NSTX, about every 2 to 3 weeks of plasma operation, by producing a glow discharge in the vacuum vessel using deuterated trimethylboron (TMB) in a 10% mixture with helium. Special NSTX requirements restricted the selection of the candidate boronization method to the use of deuterated boron compounds. Deuterated TMB met these requirements, but is a hazardous gas and special care in the execution of the boronization process is required. This paper describes the existing GDC, Gas Injection, and Torus Vacuum Pumping System hardware used for this process, the glow discharge process, and the automated control system that allows for remote operation to maximize both the safety and efficacy of applying the boron coating. The administrative requirements and the detailed procedure for the setup, operation and shutdown of the process are also described.

Boronization of the National Spherical Torus Experiment (NSTX) has enabled access to higher density, higher confinement plasmas. A glow discharge with 4 mTorr helium and 10% deuterated trimethyl boron deposited 1.7 g of boron on the plasma facing surfaces. Ion beam analysis of witness coupons showed a B+C areal density of 10 to the 18 (B+C) cm to the -2 corresponding to a film thickness of 100 nm. Subsequent ohmic discharges showed oxygen emission lines reduced by x15, carbon emission reduced by two and copper reduced to undetectable levels. After boronization, the plasma current flattop time increased by 70% enabling access to higher density, higher confinement plasmas.

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...

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.

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

The plasma contacting surfaces of the TFTR movable limiter consist of an array of titanium carbide coated graphite tiles. The surface characteristics of the graphite tiles were analyzed prior to installation to establish the initial limiter conditions, and after exposure to 800 Ohmically heated discharges in TFTR to provide data on the limiter conditioning process. Some tile surfaces showed nearly complete removal of the TiC surface as a result of plasma exposure. 3D and 2D elemental mapping of the tile near surface regions show deposition of vacuum vessel metals (primarily Ti and Ni) in the form of oxides.

The conditioning techniques applied to the TFTR first-wall configuration that will be in place for the DT experiments in 1990–1991 are reviewed. Of primary interest is the helium conditioning procedure that was developed to control hydrogenic recycling from the graphite, inner-wall bumper limiter. Operation of TFTR over the plasma density range for gas-fueled ohmic plasmas, ne = (2–5) × 1019m−3, typically results in hydrogenic recycling coefficients near unity. The use of the helium conditioning procedure produced recycling coefficients as low as 0.5, and decreased the minimum ohmic plasma density to ne = 0.5 × 1019m−3 at IP = 0.8 MA. Low density ohmic target plasmas with low recycling conditions are prerequisite conditions for the enhanced confinement (e.g., “supershot”), neutral-beam-heated discharges observed in TFTR during 1986–1987, which is the primary mode being considered for study in the DT experiments. The recycling changes induced by the helium conditioning procedure are believed to be the result of a plasma pumping effect in the graphite induced by He and C ion desorption of hydrogenic species from the near-surface (< 20 nm) layer of the limiter. The capacity of the conditioned limiter to pump gas-fueled, pellet-fueled, and neutral-beam-fueled discharges is compared. The helium conditioning technique is also beneficial for isotopic exchange and for minimizing the in-vessel tritium inventory.

A tritium compatible system has been designed to handle the exhaust gas produced by neutral beam operations in the Tokamak Fusion Test Reactor (TFTR). The system transfers gas highly concentrated in hydrogen isotopes to intermediate storage and final shipping containers using a combination of dry pumping and gettering. Details are given for functional and design requirements, component characteristics, and estimated system performance. Process flow and equipment arrangements are diagrammed, and various modes of operation are illustrated. The system is designed to provide safe handling of the tritiated exhaust and the rapid tritium turnaround necessary for operation with TFTR's small on-site inventory.

During the 1987 run, TFTR reached record values of QDD, neutron source strength, and Ti(0). Good confinement together with intense auxiliary heating has resulted in a plasma pressure greater than 3*105 Pascals on axis, which is at the ballooning stability boundary. At the same time improved diagnostics, especially ion temperature profile measurements, have led to increased understanding of tokamak confinement physics. Ion temperature profiles are much more peaked than previously thought, implying that ion thermal diffusivity, even in high ion temperature supershot plasmas, is greater than electron thermal diffusivity. Based on studies of the effect of beam orientation on plasma performance, one of the four neutral beamlines has been re-oriented from injecting co-parallel to counter parallel, which will increase the available balanced neutral injection power from 14 MW to 27 MW. With this increase in balanced beam power, and the addition of 7 MW of ICRF power it is planned to increase the present equivalant QDT of 0.25 to close to break-even conditions in the coming run.

Recent experiments on TFTR have extended the operating regime of TFTR in both ohmic- and neutral-beam -heated discharges. The TFTR tokamak has reached its original machine-design specifications ( I p = 2.5 MA and B T = 5.2 T). Initial neutral-beam -heating experiments used up to 6.3 MW of deuterium beams. With the recent installation of two additional beamlines, the power has been increased up to 11 MW. A deuterium pellet injector was used to increase the central density to 2.5 x 10 20 m -3 in high-current discharges. At the opposite extreme, by operating at low plasm a current ( I p ~ 0.8 MA) and low density ( n e~ 1 x 10 19 m -3 ), high ion temperatures (9 + 2 keV) and rotation speeds (7 x 10 5 m s -1 ) have been achieved during injection. In addition, plasma-compression experiments have demonstrated acceleration of beam ions from 82 to 150 keV, in accord with expectations. The wide operating range of TFTR, together with an extensive set of diagnostics and a flexible control system, has facilitated transport and scaling studies of both ohmic- and neutral-beam -heated discharges. The result of these confinement studies are presented.

Isotopic exchange experiments performed in the TokamakFusion Test Reactor (TFTR) are used to examine the outgassing and diffusive properties of graphite used as the plasmalimiter. Changeover from hydrogen to deuterium for different periods ranges from ∼600 to 60 plasma discharges, which appears to be correlated in the limitertemperature. We present a simple analytical model that predicts a fast transient (∼10 plasma discharges) changeover, where the deuterium fueling dilutes the adsorbed and near surface hydrogen, and a slowly changing term where bulk hydrogen diffuses to the surface. Using this model we can extract an activation energy for diffusion of 0.15±0.02 eV. We hypothesize that interpore diffusion for this porous (∼15%) material is consistent with our observations.

The principal heating system for the Tokamak Fusion Test Reactor (TFTR) consists of four beamlines to inject beams of neutral deuterium into the tokamak plasma. We have recently injected beams at total powers up to about 30 MW, and we have achieved ion temperatures in the plasma core of 30 keV and more.

This paper presents a summary of plasma-material interactions which influence the operation of TFTR with high current (≤ 2.2 MA) ohmically heated, and high-power (∼ 10 MW) neutral-beam heated plasmas. The conditioning procedures which are applied routinely to the first-wall hardware are reviewed. Fueling characteristics during gas, pellet, and neutral-beam fueling are described. Recycling coefficients near unity are observed for most gas fueled discharges. Gas fueled discharges after helium discharge conditioning of the toroidal bumper limiter, and discharges fueled by neutral beams and pellets, show R

Following a two‐year operational period the Tokamak Fusion Test Reactor (TFTR) graphite fixed bumper limiter has been examined by a variety of methods. The areal density of metals was mapped in situ by beta backscattering. Several tiles were examined in detail by nuclear‐reaction analysis, Rutherford backscattering, and proton‐induced x‐ray emission to measure areal densities of deuterium and impurities. Some areas of the limiter were found to be covered by deposited material several microns thick. Other areas where the incident plasma flux is higher were much cleaner. Long‐term collection coupons were also examined to characterize deposition on the wall. From these results the total amount of deuterium on the limiter and wall in TFTR is estimated to be ∼1.3×1024 atoms. The implications of this for wall‐pumping effects and future tritium inventory are discussed.

The neutral beam injection experiment on TFTR has now been underway for over two years, during which time we have injected as much as about 20 MW of beam power at energies up to 110 keV. We have obtained qualitative evidence of direct current drive by the injected beam ions. Recently we have produced plasmas in which the energy confinement time, βp, neutron emission, stored energy (both total and electron), and ion temperature exhibit a pronounced enhancement over previous behavior. Under these conditions we have obtained preliminary indications that significant bootstrap current is being driven. Operation will continue with new sources with which we should be able to increase the beam pulse length fourfold.

A strong pumping effect was observed with plasma operation on the toroidal graphite bumper limiter in TFTR. The pumping effect was induced by conditioning the limiter with a short series (10-20) of low density deuterium or helium initiated discharges. The decay constant for gas fuelled Ohmic discharges was reduced from before conditioning to a minimum of after conditioning, which corresponds to a reduction in the global recycling coefficient from about 100% to less than 50%. Coincident with the low recycling conditions, low current neutral beam fuelled discharges showed global energy confinement times which were enhanced by a factor of two over the values obtained with an unconditioned limiter. Two models are proposed for the observed pumping effects: (1) a depletion model, based on pumping of hydrogenic species in the near-surface region of the limiter after depletion of the normally saturated surface layer by (carbon and helium) ion induced desorption; and (2) a co-deposition model, based on adsorption of hydrogenic species in carbon films formed by material sputtered from the limiter during conditioning.

TFTR pellet injection results reported by Schmidt (1985) have been extended to higher density and n tau in plasmas limited by a graphite inner-wall belt limiter. Increased pellet penetration and larger density increases were achieved by operation at reduced plasma current (1.6 MA) and minor and major radius (70 cm and 235 cm). Under these conditions, beam heating results have been extended to 7 MW.

The characteristics of plasma operation on the axisymmetric inner toroidal limiter in TFTR are described. After conditioning, plasmas with low metal content and low zeff are obtained with this limiter. There is no substantial increase in zeff with total input power during neutral beam injection. Compared to operation on the outer blade limiter, additional gas is required to fuel plasmas on the inner limiter. Injection of D pellets increased the plasma density substantially and produced energy confinement times up to 0.5 s in ohmically heated plasmas. The four neutral beam lines have injected up to 13.5 MW total power into the plasma for 0.5 s with up to 90 kV accelerating voltage. The scaling of the plasma stored energy was studied as a function of the input power, plasma current and plasma density. In the range 1.4 to 2.2 MA, the overall and incremental confinement times for both the total and thermal stored energies increase with plasma current at fixed density. There appears to be a weak negative scaling of the total stored energy with density at high injection power.

New long-pulse ion sources have been employed to extend the neutral beam pulse on TFTR from 0.5 sec to 2.0 sec. This made it possible to study the long-term evolution of supershots at constant current and to perform experiments in which the plasma current was ramped up during the heating pulse. Experiments were conducted with co and counter injection as well as with nearly balanced injection of deuterium beams up to a total power of 20 MW. The best results, i.e., central ion temperatures Tio > 25 keV and neo τE Tio values of 3 × 1020 keV sec m-3, were obtained with nearly balanced injection. The central toroidal plasma rotation velocity scales in a linear-offset fashion with beam power and density. The scaling of the inferred global momentum confinement time with plasma parameters is inconsistent with the predictions of the neoclassical theory of gyroviscous damping. An interesting plasma regime with properties similar to the H-mode has been observed for limiter plasmas with edge qa just above 3 and 2.5.

We have studied the effects of the deposition of thin films (∼0.1 μm) of chromium onto ∼70% of the torus area of the Tokamak Fusion Test Reactor (TFTR). The purpose of these experiments was to test the difference between high surface coverage and high pumping speed gettering schemes with respect to minimizing oxygen impurity generation in high power tokamak discharges. The initial Cr deposition had significant effects on vessel outgassing and subsequent plasma performance: the outgassing of H2O, CO, and CO2 decreased by a factor of 10, oxygen impurity radiation decreased by a factor of 2, the plasma Zeff decreased from 1.3 to 1.1, and the plasma density limit increased by 20%. This improvement correlates with a significant reduction of the edge radiation as the density limit is approached. The effects of the initial and subsequent Cr depositions were relatively long lasting, exhibiting time constants of the order of weeks. We attribute the observed impurity reduction to a modification of the oxide surface...

Bare POCO AXF‐5Q graphite tiles were installed as the plasma‐facing surface of the Tokamak Fusion Test Reactor (TFTR) movable limiter for a series of ∼2700 high power discharges (600 with up to 6 MW neutral beams). During this operating phase, erosion and deposition processes modified the surface of the limiter. In the regions of the most intense plasma contact, which reached temperatures over 2400 °C, only small amounts of metallic impurities (

Plasma particle pumping effects with carbon limiters have been seen in Tokamak Fusion Test Reactor (TFTR), Joint European Torus (JET), and Torus Experiment for Technology Oriented Research (TEXTOR). The pumping effects in TFTR were induced by conditioning the large‐area axisymmetric bumper limiter with low‐density, helium initiated discharges. As a result of the conditioning ∼150 Torr l of gas are removed from the carbon limiter tiles. The conditioning has reduced the particle recycling coefficient from near unity to values as low as 0.5. The low recycling conditions are not permanent. A pumping capacity of ∼100 Torr l has been determined from wall loading experiments. The total particle input has been compared to the total outgassing after a discharge for Ohmic plasmas in both helium and deuterium, and for deuterium neutral beam heated discharges. Ohmic deuterium plasmas have less outgassing than input. Ohmic helium plasmas have equal or greater outgassing compared to gas input. Neutral beam heated disch...

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.

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/sup 0/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.

The Tokamak Fusion Test Reactor (TFTR) is a large tokamak which has performed experiments with 50:50 deuterium - tritium fuelled plasmas. Since 1993, TFTR has produced about 1090 D - T plasmas using about 100 grams of tritium and producing about 1.6 GJ of D - T fusion energy. These plasmas have significant populations of 3.5 MeV alphas (the charged D - T fusion product). TFTR research has focused on alpha particle confinement, alpha driven modes, and alpha heating studies. Maximum D - T fusion power production has aided these studies, requiring simultaneously operation at high input heating power and large energy confinement time (to produce the highest temperature and density), while maintaining low impurity content. The principal limitation to the TFTR fusion power production was the disruptive stability limit. Secondary limitations were the confinement time, and limiter power handling capability.