Search

The National Ignition Facility (NIF) diagnostic instrument manipulator (DIM) - based high resolution (dHIRES) x-ray spectrometer was used to measure the time evolution of the electron density (n e) and temperature (T e) in the hot spot of four NIF compressed capsules with 25 ps time resolution during the ‘stagnation’ phase. The electron density was inferred by comparing the measured Stark broadening of the krypton (Kr) Heβ spectral complex with theoretical calculations that include ion dynamic effects, and the electron temperature was inferred by comparing the measured ratio of the intensity of a dielectronically excited Li-like Kr line to the intensity of the Kr Heβ resonance line with calculations using the spectroscopic collisional radiative atomic model (SCRAM) and CRETIN collisional-radiative models. The inferred, time averaged n e values mainly agree with n e values from neutron diagnostics within uncertainties, but the neutron time-of-flight values of T ion are consistently higher than dHIRES T e values by 200–700 eV. The dHIRES measurements and measurement techniques, method of uncertainty analysis, and discussion of comparisons with measurements from neutron diagnostics are presented.

The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order to advance the spherical torus (ST) concept. In order to do this, NSTX utilizes up to 7.5 MW of neutral beam injection, up to 6 MW of high harmonic fast waves (HHFWs), and it operates with plasma currents up to 1.5 MA and elongations of up to 2.6 at a toroidal field up to 0.45 T. New facility, and diagnostic and modelling capabilities developed over the past two years have enabled the NSTX research team to make significant progress towards establishing this physics basis for future ST devices. Improvements in plasma control have led to more routine operation at high elongation and high βT (up to ~40%) lasting for many energy confinement times. βT can be limited by either internal or external modes. The installation of an active error field (EF) correction coil pair has expanded the operating regime at low density and has allowed for initial resonant EF amplification experiments. The determination of the confinement and transport properties of NSTX plasmas has benefitted greatly from the implementation of higher spatial resolution kinetic diagnostics. The parametric variation of confinement is similar to that at conventional aspect ratio but with values enhanced relative to those determined from conventional aspect ratio scalings and with a BT dependence. The transport is highly dependent on details of both the flow and magnetic shear. Core turbulence was measured for the first time in an ST through correlation reflectometry. Non-inductive start-up has been explored using PF-only and transient co-axial helicity injection techniques, resulting in up to 140 kA of toroidal current generated by the latter technique. Calculated bootstrap and beam-driven currents have sustained up to 60% of the flat-top plasma current in NBI discharges. Studies of HHFW absorption have indicated parametric decay of the wave and associated edge thermal ion heating. Energetic particle modes, most notably toroidal Alfven eigenmodes and fishbone-like modes result in fast particle losses, and these instabilities may affect fast ion confinement on devices such as ITER. Finally, a variety of techniques has been developed for fuelling and power and particle control.

Significant experimental and theoretical progress has been made in the U.S. heavy ion fusion program on high-current sources, injectors, transport, final focusing, chambers and targets for high energy density physics (HEDP) and inertial fusion energy (IFE) driven by induction linac accelerators. One focus of present research is the beam physics associated with quadrupole focusing of intense, space-charge dominated heavy-ion beams, including gas and electron cloud effects at high currents, and the study of long-distance-propagation effects such as emittance growth due to field errors in scaled experiments. A second area of emphasis in present research is the introduction of background plasma to neutralize the space charge of intense heavy ion beams and assist in focusing the beams to a small spot size. In the near future, research will continue in the above areas, and a new area of emphasis will be to explore the physics of neutralized beam compression and focusing to high intensities required to heat targets to high energy density conditions as well as for inertial fusion energy.

A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7?MW) and high harmonic fast wave (HHFW) heating (up to 6?MW), toroidal fields up to 6?kG, plasma currents up to 1.5?MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fuelling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparision of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10?MW?m?2 has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400?kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.

Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (?), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values ?T of up to 35% with a near unity central ?T have been obtained. NSTX will be exploring advanced regimes where ?T up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction (~60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip ~ 500?kA. In parallel, start-up using a RF current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.

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.

Spontaneous improvements of plasma confinement during auxiliary heating have been observed in many tokamaks when the q profile has been modified from its normal resistive equilibrium so that q is greater than 1 and the magnetic shear is reduced or reversed in a region near the magnetic axis. The effects on the overall plasma confinement result from the formation in the plasma interior of transport barriers, regions where the thermal and particle transport coefficients are substantially reduced. These internal barriers are sometimes tied to unique magnetic surfaces, such as the surface where the shear reverses. The reduction in transport appears to result from the suppression of turbulence by sheared plasma flow, which has now been measured in TFTR. Extensions of the theory for turbulence suppression show that this underlying paradigm may also explain other regimes of improved core confinement. The excitement generated by these discoveries must be tempered by the realization that transport and stability to pressure-driven MHD instabilities are intimately linked in these plasmas through the bootstrap current and the effect of the resulting current profile on the transport. Thus the development of control tools and strategies is essential if these improved regimes of confinement are to be exploited to improve the prospects for fusion energy production.

Circular limiter DD and DT H-modes have been obtained through transitions in supershot and high-poloidal-beta plasmas in TFTR. High tritium concentrations have been realized, with tritium fuelling mostly through the heating beams, varying from all to all beams. Density fluctuations were reduced by a factor of two across the plasma and and profiles have features consistent with simultaneous transport barriers in the plasma core and edge. values from TRANSP decrease by a factor of 2 to 5 in these regions at the transition and remain low until the onset of ELMs.

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

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.

Supershot to limiter H-mode transitions have been studied on TFTR. In several cases an edge poloidal rotation was observed prior to the initial Dalpha drop at the transition. This spin-up is consistent with formation of a radial electric field, -Er. The Del pI contribution to Er is approximately 30 kV/m and the density fluctuation level is reduced by a factor of 1 to 1.5 over a 5 cm region at the plasma edge across the transition.

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

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

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.

The principle of profile consistency states that for fixed limiter safety factor qa there are unique natural equilibrium profile shapes for the current density j(r) (and consequently q(r)) and the electron temperature Te(r) for any tokamak plasma, independent of the shapes of the heating power deposition profiles. The mathematical statements of the three basic consequences of this principle for sawtoothing discharges (i.e. discharges with q(0) ? 1) are: (1) r1/a = F1(l/qa) (? 1/qa, empirically); (2) Te/Te0 = F2(l/qa); and (3) a unique scaling law for the central electron temperature , where q(r1) = 1, Te is the volume average electron temperature, and F3(qa) = qa/q0. Since Ohm's law relates j(r) to Te(r), the principle of profile consistency dictates that this unique set of functions F1, F2 and F3 remain the same for all sawtoothing discharges in any tokamak, regardless of its size (i.e. a and R), Ip, VL, BT, etc. The paper presents a rather complete and detailed analysis of this self-consistency of the measured values of Te(r), F1, F2 and F3 for sawtoothing TFTR discharges. In particular, the analytical predictions of the profiles of Coppi's Gaussian, exponential, modified exponential and trapezoidal models, as well as the model profiles of Kadomtsev and Campbell et al. are compared with TFTR and TFR data. Some of the principal results are: (1) The empirical profile consistency relation r1/a = 1/qa is an acceptable solution of q(r1) = 1 for all qa dependent profiles. (2) A comparison between experiment and the present analytic results yields [Te/Te0]EXP ? [Te/Te0]AN + 0.05 for the profiles of Coppi's Gaussian model, and for the profiles of the models of Kadomtsev and Campbell et al. (3) For all qa independent profiles, F3(qa) = qa/q0 = constant, and, consequently, for all qa dependent profiles, F3(qa) = qa/q0 ~ qa when r1/a ? 1/qa, and, consequently, . (4) Coppi's and Ohkawa's forms of xe(r) yield , while the INTOR value of xe(r) yields . (5) For r1/a ? 1/qa, the profiles of Coppi's Gaussian model and those of the models of Kadomtsev and Campbell et al. all predict that the normalized sawtooth amplitude ?Te/Te ~ 1/qa, in agreement with the experimental observations. (6) For qa dependent models, universality of profiles exists in suitably reduced co-ordinates when r1/a ? 1/qa.

High-β spherical tokamak (ST) plasma conditions cut off propagation of electron cyclotron (EC) waves used for heating and current drive in conventional aspect ratio tokamaks. The electron Bernstein wave (EBW) has no density cutoff and is strongly absorbed and emitted at the EC harmonics, allowing EBWs to be used for heating and current drive in STs. However, this application requires efficient EBW coupling in the high-β, H-mode ST plasma regime. EBW emission (EBE) diagnostics and modelling have been employed on the National Spherical Torus Experiment (NSTX) to study oblique EBW to O-mode (B–X–O) coupling and propagation in H-mode plasmas. Efficient EBW coupling was measured before the L–H transition, but rapidly decayed thereafter. EBE simulations show that EBW collisional damping prior to mode conversion (MC) in the plasma scrape off reduces the coupling efficiency during the H-mode phase when the electron temperature is less than 30 eV inside the MC layer. Lithium evaporation during H-mode plasmas was successfully used to reduce this EBW collisional damping by reducing the electron density and increase the electron temperature in the plasma scrape off. Lithium conditioning increased the measured B–X–O coupling efficiency from less than 10% to 60%, consistent with EBE simulations.

During the past two years, the U.S. heavy ion fusion science program has made significant experimental and theoretical progress in simultaneous transverse and longitudinal beam compression, ion-beam-driven warm dense matter targets, high brightness beam transport, advanced theory and numerical simulations, and heavy ion target designs for fusion. First experiments combining radial and longitudinal compression of intense ion beams propagating through background plasma resulted in on-axis beam densities increased by 700X at the focal plane. With further improvements planned in 2007, these results will enable initial ion beam target experiments in warm dense matter to begin next year at LBNL. We are assessing how these new techniques apply to low-cost modular fusion drivers and higher-gain direct-drive targets for inertial fusion energy.

Blackbody electron cyclotron emission was used to ascertain and study the evolution and behaviour of the electron temperature profile in ohmically heated plasmas in the Tokamak Fusion Test Reactor (TFTR). The emission was measured with absolutely calibrated millimetre wavelength radiometers. The electron temperature profile normalized to the central temperature and the minor radius was observed to broaden substantially with decreasing limiter safety factor qa and to be insensitive to the plasma minor radius. Sawtooth activity was seen in the core of most TFTR discharges and appeared to be associated with a flattening of the electron temperature profile within the plasma core where q 1. Two types of sawtooth behaviour were identified in large TFTR plasmas (minor radius a 0.8 m): a 'normal' sawtooth with typically 35–40 ms period and a 'compound' sawtooth with 70–80 ms period. The compound sawtooth was characterized by what appeared to be a partial reconnection, which occurs away from the plasma core, in addition to a full reconnection. A multiple linear regression analysis of the central electron temperature Te(0), with toroidal field BT, major radius R, minor radius a, plasma current Ip and effective ion charge Zeff, results in a scaling of the form , for which the major radial dependence was obtained by comparison with earlier PLT data.

Lower hybrid current-drive experiments have been carried out on the PLT Tokamak. Steady currents up to 175 kA have been maintained for three seconds and 400 kA for 0.3 sec by the rf power alone. The principal current carrier appears to be a high energy (approx. 100 keV) electron tail, concentrated in the central 20 to 40 cm diameter core of the 80-cm PLT discharge. Effective current drive is observed only for anti n/sub e/ less than or equal to 8 x 10/sup 12/ cm/sup -3/. This limitation may be a wave propagation phenomenon and not a fundamental plasma physics effects.

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.

The bremsstrahlung emission from the PLT tokamak during lower-hybrid current drive has been measured as a function of angle between the magnetic field and the emission direction. The emission is peaked strongly in the forward direction, indicating a strong anisotropy of the electron velocity distribution. The data demonstrate the existence of a nearly flat tail of the velocity distribution, which extends out to approximately 500 keV and which is interpreted as the plateau created by Landau damping of the lower-hybrid waves.

TSTR (Tokamak Fusion Test Reactor) has operated since December 1982 with ohmically heated plasmas. Routine operation with feedback control of plasma current, position and density has been obtained for plasmas with Ip800 kA, a = 68 cm, R = 250 cm, and Bt=27 kG. A maximum plasma current of 1 MA was achieved with q2.5. Energy confinement times of ~150 msec were measured for hydrogen and deuterium plasmas with e = 2 x 1013 cm-3, Te(0) 21.5 keV, Ti(0) = 1.5 keV and Zeff1 3. The preliminary results suqgest a size-cubed scaling from PLT, and are consistent with Alcator C scaling where T ~ nR2a. Initial measurements of plasma disruption characteristics indicate current decay rates of ~ 800 kA in 8 ms which is within the TFTR design requirement of 3 MA in 3 ms.

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.

Radio-frequency power of up to 110 kW for 14 ms at 28 GHz is launched in a horizontal polarization from the outside midplane at an angle nearly perpendicular to the 10-kG toroidal field of the JFT-2 tokamak. With initial Ohmic power comparable to the RF power, the central electron temperature increases from 600 to 1000 eV in 10 ms. Details of the electron heating are reported.

The Ohmic targets of supershot (Strachan et al., 1987) plasmas in the Tokamak Fusion Test Reactor (TFTR) are often characterized by the presence of nonthermal electron cyclotron emission (ECE) from a runaway or slideaway electron population. This nonthermal ECE decays away within the first 100-150 ms after the start of neutral beam injection (NBI) as the electron density rises and the loop voltage drops. After this usual slideaway-induced, nonthermal ECE is quenched a new nonthermal feature develops at frequencies just below both the optically thick, thermal emission of the ECE fundamental ordinary mode, and the second harmonic extraordinary mode. This paper presents an analysis of the ECE data which indicates a possible source of this nonthermal emission feature. In addition, correlation of the peak intensity of the nonthermal feature with various plasma parameters for 240 TFTR discharges from 1988 provides some insight toward identifying possible physical conditions which give rise to the feature.

The time-scales for sawtooth repetition and heat pulse propagation are much longer (tens of milliseconds) in the large tokamak TFTR than in previous, smaller tokamaks. This extended time-scale, coupled with more detailed diagnostics, has led us to revisit the analysis of heat pulse propagation as a method to determine the electron heat diffusivity χe in the plasma. A combination of analytic and computer solutions of the electron heat diffusion equation is used to clarify previous work and to develop new methods for determining χe. Direct comparison of the predicted heat pulses with soft-X-ray and ECE data indicates that the space-time evolution is diffusive. However, the χe determined from heat pulse propagation usually exceeds that determined from background plasma power balance considerations by a factor ranging from two to ten. Some hypotheses for resolving this discrepancy are discussed.

Lower-hybrid current drive requires the generation of a high-energy electron tail anisotropic in velocity. Measurements of bremsstrahlung emission produced by this tail are compared with the calculated emission from reasonable model distributions. The physical basis and the sensitivity of this modelling process are described, and the plasma properties of current-driven discharges which can be derived from the model are discussed.

Absorber foils have been installed in the TFTR X-Ray Imaging System to permit measurement of the electron temperature along 10 to 30 chords spaced at 5-12.5 cm with a time resolution of less than 100 ..mu..s. The technique uses the ratio of x-ray fluxes transmitted through two different foils. The ratio depends mainly on electron temperature. Simulations show that strong impurity line radiation can distort this ratio. To correct for these effects, special beryllium-scandium filters are employed to select the line-free region between 2 and 4.5 keV. Other filter pairs allow corrections for Fe L and Ni L line radiation as well as Ti K and Ni K emission. Good accuracy is also obtained with simple beryllium filters, provided that impurity corrections are incorporated in the analysis, taking line intensities from the x-ray pulse-height analysis diagnostic. A description of modeling calculations and a comparison of temperature values from this diagnostic with data from the x-ray pulse height analysis, the electron cyclotron emission, and the Thomson scattering diagnostics are presented. Several applications of the absorber foil electron temperature diagnostic on TFTR are discussed.

Transport simulations of ohmically heated TFTR experiments with recently developed microinstability based, profile consistent models for the anomalous thermal diffusivities, χe and χi, give good agreement with experimental data. The steady state temperature profiles and the total energy confinement times, τE, were found to agree for each of the Ohmic TFTR experiments simulated, including three high radiation cases and two plasmas fuelled by pellet injection. Both collisional and collisionless models are tested. The trapped electron drift wave microinstability model results are consistent with the thermal confinement of large plasma Ohmic experiments on TFTR. It is also found that transport due to the profile consistent model based on toroidal ion temperature gradient (ηi) mode transport can cause saturation in τE at the highest densities comparable to that observed on TFTR and equivalent to a neoclassical anomaly factor of three. Predictions based on stabilized ηi mode driven ion transport are found to be in agreement with the enhanced global energy confinement times for pellet fuelled plasmas.

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.

Electron cyclotron wave absorption by moderately relativistic electrons in the low-density regime of the PLT tokamak is investigated. Appreciable wave damping is found for vertical propagation at frequencies of 50, 60, and 70 GHz when the spatially constant cyclotron frequency is 89 GHz. The perpendicular temperature T⊥ (v||) of the fast tail is estimated from emission of radiation in the same direction. The results obtained are in satisfactory agreement with the theory of wave emission and absorption.

Fusion neutron emission of 1.5 × 1014 neutrons s−1 and 2 × 1013 neutrons/pulse has been observed for PLT deuterium discharges with up to 2.5 MW of deuterium neutral-beam injection. The neutron time evolution and magnitude are consistent with theoretical calculations of the fusion reactions caused by energetic injected ions which are confined and slow down classically. The factor-of-two accuracy in the absolute neutron calibration is the major uncertainty in the comparison with theory. Neutron sawtooth oscillations ( 3%) are observed which can also be explained classically.

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.

Introduction of large amounts of neon into Ohmically heated deuterium discharges in the PLT tokamakl results in higher central electron temperature (Te(0) 3 keV) and values of electron energy containment time that are larger than in regular discharges at the same electron density (τEe = 44 ms at e = 2 × 1019 m−3). For steady-state discharges with high effective Z (~ 5–8) the conductance is larger than that predicted by neoclassical theory by as much as a factor of two. Transport rates of hydrogen-like and helium-like ions can be fairly well approximated by assuming a diffusion constant of between 0.4 and 1 m2s−1. Within experimental uncertainties the diffusion of hydrogen-like neon is the same for co- and counter-directed high-power neutral beams.