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The mission of the low aspect ratio spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step steady-state tokamak fusion devices. NSTX-U will ultimately operate at up to 2 MA of plasma current and 1 T toroidal field on axis for 5 seconds, and has available up to 15 MW of Neutral Beam Injection (NBI) power at different tangency radii and 6 MW of High Harmonic Fast Wave (HHFW) heating. With these capabilities NSTX-U will develop the physics understanding and control tools to ramp-up and sustain high performance fully non-inductive plasmas with large bootstrap fraction and enhanced confinement enabled via the low aspect ratio, high beta configuration. With its unique capabilities, NSTX-U research also supports ITER and other critical fusion development needs. Super-Alfvénic ions in beam-heated NSTX-U plasmas access energetic particle parameter space that is relevant for both -heated conventional and low aspect ratio burning plasmas. NSTX-U can also generate very large target heat fluxes to test conventional and innovative plasma exhaust and plasma facing component (PFC) solutions. This paper summarizes recent analysis, theory and modelling progress to advance the tokamak physics basis in the areas of macrostability and 3D fields, energetic particle stability and fast ion transport, thermal transport and pedestal structure, boundary and plasma material interaction, RF heating, scenario optimization and real-time control.

We report on the development and application of coherent Rayleigh-Brillouin scattering for the in situ detection of large molecules and nanoparticles. This four wave mixing diagnostic technique relies on the creation of an electrostrictive optical lattice in a medium due to the interaction between polarized particles and the intense electric field gradient created by the optical interference of two intense pulsed laser beams. Though this interaction, we can detect the temperature, pressure, relative density, polarizability and speed of sound of a gas and gas mixture. This diagnostic was already successfully demonstrated in atomic and molecular gaseous environments, where the different gas polarizabilities and pressures were successfully measured. We are currently conducting in situ measurements with large molecules and nanoparticles produced in an arc discharge, the results of which will be presented in this meeting.

We report on the development and application of a new laser diagnostic for the in situ detection of large molecules and nanoparticles. This four wave mixing diagnostic technique relies on the creation of an optical lattice in a medium due to the interaction between polarized particles and intense laser fields. Though this interaction, we can detect the temperature, pressure, relative density, polarizability and speed of sound of a gas and gas mixture. This diagnostic was already successfully demonstrated in atomic and molecular gaseous environments, where the different gas polarizabilities and pressures were successfully measured. We are currently conducting measurements with large molecules and nanoparticles, the results of which will be presented in this meeting.

An overview of the research capabilities and the future plans on the MA-class National Spherical Torus Experiment (NSTX) at Princeton is presented. NSTX research is exploring the scientific benefits of modifying the field line structure from that in more conventional aspect ratio devices, such as the tokamak. The relevant scientific issues pursued on NSTX include energy confinement, MHD stability at high beta, non-inductive sustainment, solenoid-free start-up, and power and particle handling. In support of the NSTX research goal, research tools are being developed by the NSTX team. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a high beta Demo device based on the ST, are being considered. For these, it is essential to develop high performance (high beta and high confinement), steady-state (non-inductively driven) ST operational scenarios and an efficient solenoid-free start-up concept. We will also briefly describe the Next-Step-ST (NSST) device being designed to address these issues in fusion-relevant plasma conditions.

A new type of high-resolution x-ray imaging crystal spectrometer is being developed to measure ion and electron temperature profiles in tokamak plasmas. The instrument is particularly valuable for diagnosing plasmas with purely ohmic heating and rf heating, since it does not require the injection of a neutral beam—although it can also be used for the diagnosis of neutral-beam heated plasmas. The spectrometer consists of a spherically bent quartz crystal and a two-dimensional position-sensitive detector. It records spectra of helium-like argon (or krypton) from multiple sightlines through the plasma and projects a de-magnified image of a large plasma cross section onto the detector. The spatial resolution in the plasma is solely determined by the height of the crystal, its radius of curvature, and the Bragg angle. This new x-ray imaging crystal spectrometer may also be of interest for the diagnosis of ion temperature profiles in future large tokamaks, the Korea Superconducting Tokamak Advanced Research tok...

The multilevel physics, massively parallel plasma simulation code, M3D has been used to study spherical toris (STs) and tokamaks. The magnitude of outboard shift of density profiles relative to electron temperature profiles seen in NSTX under strong toroidal flow is explained. Internal reconnection events in ST discharges can be classified depending on the crash mechanism, just as in tokamak discharges; a sawtooth crash, disruption due to stochasticity, or high-β disruption. Toroidal shear flow can reduce linear growth of internal kink. It has a strong stabilizing effect nonlinearly and causes mode saturation if its profile is maintained, e.g. through a fast momentum source. Normally, however, the flow profile itself flattens during the reconnection process, allowing a complete reconnection to occur. In some cases, the maximum density and pressure spontaneously occur inside the island and cause mode saturation. Gyrokinetic hot particle/MHD hybrid studies of NSTX show the effects of fluid compression on a fast-ion driven n = 1 mode. MHD studies of recent tokamak experiments with a central current hole indicate that the current clamping is due to sawtooth-like crashes, but with n = 0.

The National Spherical Torus Experiment (NSTX) at Princeton University, Princeton, NJ, is designed to assess the potential of the low-aspect-ratio spherical torus concept for magnetic plasma confinement. The plasma has been heated by up to 7 MW of neutral beam injection (NBI) at an injection energy of 100 keV and up to 6 MW of high harmonic fast wave (HHFW) at 30 MHz. NSTX has achieved /spl beta//sub T/ of 32%. A variety of MHD phenomena have been observed to limit /spl beta/. NSTX has now begun addressing /spl tau//sub E/ scaling, /spl beta/ limits, and current drive issues. During the NBI heating experiments, a broad T/sub i/ profile with T/sub i/ up to 2 keV, T/sub i/>T/sub e/ and a large toroidal rotation were observed. Transport analysis suggests that the impurity ions have diffusivities approaching neoclassical. For L-Mode plasmas, /spl tau//sub E/ is up to two times the ITER97L L-Mode scaling and exceeds the ITER98pby2 H-Mode scaling in some cases. Transitions to H-Mode have been observed which result in an approximate doubling of /spl tau//sub E/ after the transition in some conditions. During HHFW heating, T/sub e/>T/sub i/ and T/sub e/ up to 3.5 keV were observed. Current drive has been studied using both coaxial helicity injection with up to 390 kA of toroidal current and HHFW. HHFW has produced H-modes with significant bootstrap current fraction at low I/sub p/, high q, and high /spl beta//sub p/.

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

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.

Results from helium, iron, and electron transport studies on the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Nucl. Fusion Res. 26, 11 (1984)] in L‐mode and supershot deuterium plasmas with the same toroidal field, plasma current, and neutral beam heating power are presented. They are compared to results from thermal transport analysis based on power balance. Particle diffusivities and thermal conductivities are radially hollow and larger than neoclassical values, except possibly near the magnetic axis. The ion channel dominates over the electron channel in both particle and thermal diffusion. A peaked helium profile, supported by inward convection that is stronger than predicted by neoclassical theory, is measured in the supershot. The helium profile shape is consistent with predictions from quasilinear electrostatic drift‐wave theory. While the perturbative particle diffusion coefficients of all three species are similar in the supershot, differences are found in the L mode. Quasilinear theory calculations of the ratios of impurity diffusivities are in good accord with measurements. Theory estimates indicate that the ion heat flux should be larger than the electron heat flux, consistent with power balance analysis. However, theoretical values of the ratio of the ion to electron heat flux can be more than a factor of 3 larger than experimental values. A correlation between helium diffusion and ion thermal transport is observed and has favorable implications for sustained ignition of a tokamak fusion reactor.

A high throughput, shielded, multichannel, extreme ultraviolet (XUV) spectrograph to be used for plasma diagnostics was built. The instrument uses seven hollow fibers to transport the radiation by multiple grazing incidence reflections from the tokamak port through the instrument’s shield to the spectrometer entrance slit. In this way, the ratio of background radiation streaming through the entrance slit to XUV flux is reduced by a factor of at least 2500. The seven hollow fibers provide seven channels corresponding to seven different lines of sight through the plasma. The spectrograph uses a toroidal mirror together with a plane, aberration‐corrected grating to disperse radiation in the 50–200 A region. Up to seven spectra are recorded simultaneously, at rates of 30–90 Hz, using a microchannel plate assembly optically coupled to a charge‐coupled device camera. This will permit, during a single discharge, measurement of spatial emissivity profiles from various ionization states of an injected impurity, pr...

Solid lithium and boron pellets have been injected into TFTR plasmas to improve plasma performance by coating the graphite inner wall bumper limiter with a small amount of lower Z pellet material, which reduces the influx of carbon from the walls and reduces the edge electron density. This new wall conditioning technique has been applied successfully when continued He conditioning discharges, which are normally used for wall conditioning, no longer significantly reduce the carbon and deuterium influxes. The results show that both Li and B pellets significantly improve wall conditioning and lead to 15–20% improvements in supershot plasma performance when injected ≥1 s prior to neutral beam injection in supershot target plasmas. Neutral beam penetration calculations indicate that the lower edge densities resulting from Li or B pellet wall conditioning lead to improved beam penetration. Sputtering yield calculations confirm that the addition of small amounts of Li on a graphite target can significantly reduce the C sputtering yield.

Steady‐state and perturbative transport analysis are complementary techniques for the study of transport in tokamaks. These techniques are applied to the investigation of auxiliary‐heated L‐mode and supershot plasmas in the tokamak fusion test reactor (TFTR) [R. J. Hawryluk et al., Plasma Physics and Controlled Nuclear Fusion Research, Proceedings of the 11th International Conference, Kyoto, 1986 (IAEA, Vienna, 1987), Vol. 1, p. 51.]. In the L mode, both steady‐state and perturbative transport measurements reveal a strong temperature dependence that is consistent with electrostatic microinstability theory and the degradation of confinement with neutral beam power. Steady‐state analysis of the ion heat and momentum balance in supershots indicates a reduction and a significant weakening of the power‐law dependence on the transport in the center of the discharge.

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.

Transport of iron in TFTR discharges heated by 6.7 MW of co-injected neutral beam power has been studied by spatially resolved charge exchange recombination spectroscopy in the visible region of the spectrum. The time evolutions of the densities of Fe24+ and Fe23+ ions following injection of iron are modelled by the neoclassical flux plus a moderately hollow diffusivity, increasing linearly from 1.3 m2/s on axis to 2.4 m2/s at the plasma edge. For r/a > 0.5, the iron diffusivity is significantly smaller than the helium diffusivity measured in identical discharges, while it is larger in the region immediately surrounding the plasma axis.

Measurements of the spatial structure of the transport coefficients of He²+ on the Tokamak Fusion Test Reactor is reported. He²+ profiles were measured using charge-exchange spectroscopy after a helium puff into corotating L-mode plasmas. Modeling shows that the He²+ diffusivity is about 10 m²/s near the plasma edge and drops to below 1 m²/s inside of the q =1 surface. The convective velocity ranges from 1–3 m/s near q =1 to 20–40 m/s near the edge. The helium diffusivity is on the order of the ion momentum and thermal diffusivity and is greater than the electron thermal diffusivity.

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, ...

The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). Its mission is to develop the physics understanding of the compact stellarator and evaluate its potential for future fusion energy systems. Compact stellarators use 3D plasma shaping to produce a magnetic configuration that can be steady state without current drive or feedback control of instabilities. The NCSX has major radius 1.4 m, aspect ratio 4.4, 3 field periods, and a quasi-axisymmetric magnetic field. It is predicted to be stable and have good magnetic surfaces at beta > 4% and to have tokamak-like confinement properties. The device will provide the plasma configuration flexibility and the heating and diagnostic access needed to test physics predictions. Component production has advanced substantially since the first contracts were placed in 2004. Manufacture of the vacuum vessel was completed in 2006. All eighteen modular coil winding forms have been delivered, and twelve modular coils have been wound and epoxy impregnated. A contract for the (planar) toroidal field coils was placed in 2006 and manufacture is in progress. Assembly activities have begun and will be the project's main focus in the next few years. The engineering challenge of NCSX is to meet the requirements for complex geometries and tight tolerances within the cost and schedule constraints of a construction project. This paper will focus on how the engineering challenges of component production have been resolved, and how the assembly challenges are being met.

The National Compact Stellerator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). The complex geometry and tight fabrication tolerances of the NCSX's non-planar coils and vacuum vessel necessitate the use of computerized, CAD-based metrology systems capable of very accurate and reasonably quick measurements. To date, multi-link, portable coordinate measuring machines (pCMM) are used in the fabrication of the non-planar coils. Characterization of the CNC machined coil winding form and subsequent positioning of the conductor centroid (to within +/-0.5 mm) are accomplished via multiple sets of detailed measurements. A laser tracker is used for all phases of work on the vacuum vessel including positioning magnetic diagnostics and vessel ports prior to welding. Future tasks requiring metrology include positioning of the magnet systems and assembly of the three vacuum vessel sub-assemblies onto the final machine configuration. This paper describes the hardware and software used for metrology, as well as the methodology for achieving the required dimensional control and will present an overview of the measurement results to date.

The national compact stellarator experiment (NCSX) will have an extensive set of external magnetic diagnostics. These include flux loops on the exterior surface of the vacuum vessel. Data from these sensors will be integrated with other magnetic sensors and used for plasma control and to constrain magnetic equilibrium reconstructions. NCSX is currently under construction at the Princeton Plasma Physics Laboratory (PPPL). The ex-vessel flux loops must be installed during machine construction since they will ultimately be trapped in the space between the vacuum vessel and the modular coil support shell. Detailed designs have been completed, locator templates have been fabricated and approximately one third of the 225 total loops have been installed as of mid February 2007. Modeling was performed by PPPL to determine the optimum size, placement and number of turns. Engineering of the flux loops was challenging as they must be accurately positioned, optimized geometry maintained and they must be robust and reliable in a bake and cryogenic environment for the lifetime of NCSX. Designs for the ex-vessel flux loops that meet these requirements are presented.

The multilevel physics, massively parallel plasma simulation code, M3D, has been used to study spherical tori (STs) and tokamaks. The magnitude of outboard shift of density profiles relative to electron temperature profiles seen in NSTX [National Spherical Torus Experiment] under strong toroidal flow is explained. Internal reconnection events in ST discharges can be classified depending on the crash mechanism, just as in tokamak discharges; a sawtooth crash, disruption due to stochasticity, or high-beta disruption. Toroidal shear flow can reduce linear growth of internal kink. It has a strong stabilizing effect nonlinearly and causes mode saturation if its profile is maintained, e.g., through a fast momentum source. Normally, however, the flow profile itself flattens during the reconnection process, allowing a complete reconnection to occur. In some cases, the maximum density and pressure spontaneously occur inside the island and cause mode saturation. Gyrokinetic hot particle/MHD hybrid studies of NSTX show the effects of fluid compression on a fast-ion-driven n = 1 mode. MHD studies of recent tokamak experiments with a central current hole indicate that the current clamping is due to sawtooth-like crashes, but with n = 0.

Injection of Lower Hybrid Heating and Current Drive (LHCD) into the current ramp-up phase of JET [Joint European Torus] discharges can produce extremely reversed q-profiles characterized by a core region of near zero current density (within Motional Stark Effect diagnostic measurement errors). Non-inductive, off-axis co-current drive induces a back electromotive force inside the non-inductive current radius that drives a negative current in the plasma core. The core current density does not go negative, although current diffusion calculations indicate that there is sufficient LHCD to cause this. The clamping of the core current density near zero is consistent with n=0 reconnection events redistributing the core current soon after it goes negative. This is seen in reduced MHD simulations and in nonlinear resistive MHD simulations which predict that these discharges undergo n=0 reconnection events that clamp the core current near zero.

Present‐day tokamak x‐ray imaging (XIS) and pulse height analysis (PHA) diagnostics will require special shielding and x‐ray optics to permit use on fusion reactors without prohibitive noise and detector damage from neutrons and gamma rays; x‐ray curved‐crystal spectrometers (XCS) may work with extensive shielding and collimation, but radiation damage of crystals and attainment of adequate impurity concentrations for ion‐temperature measurement are concerns. We consider the use of one or more reflections at grazing incidence from x‐ray mirrors or from Bragg layered synthetic microstructures (LSM) to decouple the x‐ray diagnostic from the direct fusion neutron beam. We present calculations of expected x‐ray line brightnesses from ITER and total instrument throughput. We also consider the use of hollow glass capillaries embedded in radiation shields to precede the XIS detector and reduce the ratio of neutron plus gamma radiation to x rays by a factor of ∼0.01 or better. Compatibility of capillary schemes with the PHA and XCS are discussed.

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.

Local thermal particle and energy transport studies of balanced-injection L-mode and Supershot deuterium plasmas with the same toroidal field, plasma current, and neutral beam heating power have been performed on TFTR. The particle transport of He{sup 2+} and electrons following a small helium gas puff and Fe{sup 24+} induced by laser ablation has been examined and compared to the local energy transport characteristics inferred from power balance analysis. All particle perturbation diffusivities are radially hollow and are similar in magnitude and shape to the effective thermal conductivities found by power balance analysis. All particle diffusivities are 1--2 orders of magnitude larger than neoclassical values, except near the magnetic axis. A reduction in the helium diffusivity D{sub He} in the Supershot as compared to the L-mode is accompanied by a similar reduction in the effective single fluid thermal conductivity {chi}fluid. Also, the helium core convective velocity V{sub He} is found to increase in the Supershot over the L-Mode for r/a < 0.5. A quasilinear model of electrostatic drift waves has been used to calculate ratios between particle and energy fluxes in the Supershot. The measured ratios of the helium and iron particle diffusivities are in good accord with predictions, as aremore » predicted ratios of V{sub He}/D{sub He}. Modelling indicates that the similarity in magnitude and profile shape of D{sub He} and {chi}fluid has generally favorable implications for helium ash content in a future fusion reactor. The core convection found in the Supershot increases the helium concentration on axis but does not reduce the plasma reactivity significantly.« less

X-ray spectroscopy of mid-Z metal impurities is important in the study of tokamak plasmas and may reveal potential problems if their contribution to the radiated power becomes substantial. The analysis of the data from a high-resolution x-ray and extreme ultraviolet grating spectrometer, XEUS, installed on NSTX, was performed focused on a detailed study of x-ray spectra in the range 7-18 Å. These spectra include not only commonly seen iron spectra but also copper spectra not yet employed as an NSTX plasma impurity diagnostic. In particular, the L-shell Cu spectra were modeled and predictions were made for identifying contributions from various Cu ions in different spectral bands. Also, similar spectra, but from much denser Cu plasmas produced on the UNR Z-pinch facility and collected using the convex-crystal spectrometer, were analyzed and compared with NSTX results.

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.

New or improved wavelengths are given for several long-wavelength magnetic dipole lines in highly ionized nickel, which have been observed in the Tokamak Fusion Test Reactor plasmas, as well as the 2s22p4(3P0–3P1) line in Cu xxii.

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.

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.

Plasma-materials interactions studied in recent ICRF heating and lower hybrid current drive experiments are reviewed. The microscopic processes responsible for impurity generation are discussed. In ICRF experiments, improvements in machine operation and in antenna and feedthrough design have allowed efficient plasma heating at RF powers up to 3 MW. No significant loss of energy from the plasma core due to impurity radiation occurs. Lower hybrid current drive results in the generation and maintenance of hundreds of kiloamperes of plasma current carried by suprathermal electrons. The loss of these electrons and their role in impurity generation are assessed. Methods to avoid this problem are evaluated.

The paper presents a study of impurity transport in ohmically heated TFTR plasmas by computer modelling of VUV line emissions from impurities injected using the laser blow-off technique. The results are sensitive to uncertainties in the ionization and recombination rates used in the modelling and, therefore, only a spatially averaged diffusion coefficient and parameterized convective velocity can be measured. Measurements of these transport parameters are presented for deuterium and helium discharges with Ip = 0.8−2.5 MA, e = (0.6-6.0) × 1019 m−3 and Zeff = 2-6. The diffusion coefficients are found to be in the range of 0.5-1.5 m2 s−1, considerably larger than neoclassical values. Non-zero inward convective velocities are necessary to fit the data in most cases. No dependence of the diffusion coefficient on injected element, working gas species or plasma current is found, but, at a given current, the diffusion coefficient in plasmas near the density limit is smaller by approximately a factor of two than in discharges with e < 3 × 1019 m−3.

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.

During the q-mode in TFTR, the Dα emissions decrease and the electron density, Zineff, and confinement times increase, as seen during the H-mode in other tokamaks. The interaction of surface islands with the limiter is believed to increase the carbon and electron concentration. Analysis of the deuterium recycling indicates that the recycling coefficient at the last closed flux surface decreases about 15% during the q-mode.

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.

Power loss studies were carried out to determine the impurity radiation and energy transport characteristics of various Tokamak Fusion Test Reactor (TFTR) operation and confinement regimes including low mode, detached plasma, co‐only neutral beam injection (energetic ion regime), and the enhanced confinement (‘‘supershot’’) regime. Combined bolometric, spectroscopic, and infrared photometry measurements provide a picture of impurity behavior and power accounting in TFTR. The purpose of this paper is to make a survey of the various regimes with the aim of determining the radiated power signatures of each.

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

The paper describes the present (end of February 1985) status of the plasma confinement studies in the TFTR tokamak with emphasis on those with neutral beam injection (NBI). Recent improvements in the device capabilities have substantially extended operating parameters: B /sub T/ increased to 4.0 T, I /sub p/ to 2.0 MA, injection power (P /sub b/ ) to /sup 5/ MW with H/sup 0/ or D/sup 0/ beams, n /sub e/ to 5 x 10/sup 19/ m/sup -3/ and Z /sub eff/ reduced to 1.4. With ohmic heating (OH) alone, the previously established scaling for gross energy confinement time (T /sub E/ /sup 0/ n /sub e/ q) has been confirmed at higher I /sub p/ and B /sub T/ , and the maximum T /sub E/ of 0.4 sec has been achieved. With NBI at P /sub b/ substantially (by factor > 2) higher than P /sub OH/ , excellent power and particle accountability have been established. This suggest that the less-than-expected increase in stored energy with NBI is not due to problems of power delivery, but due to problems of confinement deterioration. T /sub E/ is observed to scale approximately as I /sub p/ P /sub b/more » /sup 0.5/ (independent of n /sub e/ ), consistent with previous L-mode scalings. With NBI we have achieved the maximum T /sub E/ of 0.2 s and the maximum T /sub i/ (o) of 4.4 keV in the normal operating regime, and even higher T /sub i/ (o) in the energetic-ion regime with low-n /sub e/ and low I /sub p/ operation.« less

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.

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.

Hydrogenlike oxygen and carbon lines from transitions (Δn=1) between levels of high principal quantum number n, with wavelengths above 2000 A have been observed. Observations of such transitions were possible due to charge‐exchange processes during neutral beam injection of hydrogen atoms into the Princeton Large Torus tokamak. The lines are O VIII 2976 A (8–7 transitions), C VI 3434 A (7–6), and C VI 5291 A (8–7). Application of these lines for ion temperature measurements and initial observations of neutral beam vertical distributions in the plasma are presented.

The soft x‐ray continuum radiation in tokamak fusion test reactor (TFTR) low‐density neutral‐beam discharges can be much lower than its theoretical value obtained by assuming a corona equilibrium. This reduced continuum radiation is caused by an ionization equilibrium shift toward lower states, which strongly changes the value of the average recombination coefficient of metallic impurities γ, even for only slight changes in the average charge Z. The primary agent for this shift is the charge exchange between the highly ionized impurity ions and the neutral hydrogen, rather than impurity transport, because the central density of the neutral hydrogen is strongly enhanced at lower plasma densities with intense beam injection. In the extreme case of low‐density, high neutral‐beam power TFTR operation (energetic ion mode) the reduction in γ can be as much as one‐half to two‐thirds. We calculate the parametric dependence of γ and Z for Ti, Cr, Fe, and Ni impurities on neutral density (equivalent to beam power), electron temperature, and electron density. These values are obtained by using either a one‐dimensional impurity transport code or a zero‐dimensional code with a finite particle confinement time. As an example, we show the variation of γ and Z in different TFTR discharges.

Metal impurity concentrations are measured by the pulse‐height‐analyzer (PHA) diagnostic from Kα x‐ray peak intensities by use of an averaged excitation rate 〈σv〉. Low‐Z impurity concentrations are inferred from the continuum enhancement (relative to a pure plasma) minus the enhancement due to metals. Since the PHA does not resolve lines from different charge states, 〈σv〉 is a weighted sum of rates; coronal equilibrium is usually assumed. The 〈σv〉 used earlier omitted the intercombination and forbidden lines from the dominant heliumlike state. The result was an overestimate of metals and an underestimate of low‐Z impurities in cases where metals were significant. Improved values of 〈σv〉 using recent calculations for H‐, He‐, and Li‐like Fe range from 10% to 50% larger than the earlier rates and yield metal concentrations in better agreement with those from vacuum ultraviolet (VUV) spectroscopy.

Two improvements to the SPRED multichannel VUV spectrographs used on the TFTR and PBX tokamaks have been made: (1) a new 2100‐g/mm grating covering the 100–320‐A region with 0.4‐A resolution (FWHM) has been added to the existing 450‐g/mm grating (100–1100 A with 2‐A resolution), and (2) the TFTR SPRED has been absolutely calibrated using synchrotron radiation from the NBS SURF II facility, while the PBX system has been calibrated using conventional branching ratios along with line ratios from charge‐exchange‐recombination excited lines. The availability of high‐resolution spectra in the 100–320‐A range provides improved measurements of metallic ion emissions and, when the instrument views across a neutral beam as in PBX, allows carbon and oxygen densities to be measured via charge‐exchange‐recombination spectroscopy.

Synchrotron Radiation Calibration of the SPRED VUV Spectograph at the NBS SURF II Electron Storage RingB.C. Stratton, R.J. Fonck, and A.T. RamseyPrinceton University, Plasma Physics LaboratoryJames Forrestal Campus, P.O. Box 451, Princeton, New Jersey 08544AbstractAn absolute intensity calibration of the SPRED VUV spectrograph currently in use on the Tokamak FusionTest Reactor (TFTR) has been performed using synchrotron radiation from the NBS SURF II electron storage

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.