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High-density micron-sized particle aerosols might form the basis for a number of applications in which a material target with a particular shape might be quickly ionized to form a cylindrical or sheet shaped plasma. A simple experimental device was built in order to study the properties of high-density aerosol focusing for 1 μm silica spheres. Preliminary results recover previous findings on aerodynamic focusing at low densities. At higher densities, it is demonstrated that the focusing properties change in a way which is consistent with a density dependent Stokes number.

The two-dimensional radial vs poloidal structure and motion of edge turbulence in the National Spherical Torus Experiment (NSTX) were measured using high-speed imaging of the visible light emission from a localized neutral gas puff. Edge turbulence images are shown and analysed for Ohmic, L- and H-mode plasma conditions. The two-dimensional images often show regions of strong localized light emission known as 'blobs', which move both poloidally and radially at a typical speed of ?105?cm?s?1, and sometimes show spatially periodic features.

Coaxial helicity injection (CHI) on the National Spherical Torus Experiment (NSTX) has produced 240 kA of toroidal current without the use of the central solenoid. Values of the current multiplication ratio (CHI produced toroidal current/injector current) up to 10 were obtained, in agreement with predictions. The discharges, which lasted for up to 200 ms, limited only by the programmed waveform, are more than an order of magnitude longer in duration than any CHI discharges previously produced in a spheromak or a spherical torus.

Alpha particle physics experiments were done on TFTR during its DT run from 1993 to 1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single particle confinement model in MHD quiescent discharges. The alpha loss due to toroidal field ripple was identified in some cases, and the low radial diffusivity inferred for high energy alphas was consistent with orbit averaging over small scale turbulence. Finally, the observed alpha particle interactions with sawteeth, toroidal Alfven eigenmodes and ICRF waves were approximately consistent with theoretical modelling. What was learned is reviewed and what remains to be understood is identified.

The article reviews the physics of fusion alpha particles and energetic neutral beam ions studied in the final phase of TFTR operation, with an emphasis on observations in reversed magnetic shear (RS) and enhanced reversed shear (ERS) DT plasmas. Energy resolved measurements of the radial profiles of confined, trapped alphas in RS plasmas exhibit reduced core alpha density with increasing alpha energy, in contrast to plasmas with normal monotonic shear. The measured profiles are consistent with predictions of increased alpha loss due to stochastic ripple diffusion and increased first orbit loss in RS plasmas. In experiments in which a short tritium beam pulse is injected into a deuterium RS plasma, the measured DT neutron emission is lower than standard predictions assuming first orbit loss and stochastic ripple diffusion of the beam ions. A microwave reflectometer measured the spatial localization of low toroidal mode number (n), alpha driven toroidal Alfven eigenmodes (TAEs) in DT RS discharges. Although the observed ballooning character of the n = 4 mode is consistent with predictions of a kinetic MHD stability code, the observed antiballooning nature of the n = 2 mode is not. Furthermore, the modelling does not show the observed strong dependence of mode frequency on n. These alpha driven TAEs do not cause measurable alpha loss in TFTR. Other Alfven frequency modes with n = 2-4 seen in both DT and DD ERS and RS discharges are localized to the weak magnetic shear region near qmin. In 10-20% of DT discharges, normal low n MHD activity causes alpha loss at levels above the first orbit loss rate.

The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approximately}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the P{sub FUS} {approximately}6 MW level. Instability in the TAE mode frequency range has been observed at P{sub FUS} > 7 MW and its effect on performance is under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fullymore » explored.« less

Measurements were made of the motion of the filamentary structures in a plasma ball using high speed cameras and other optical detectors. These filaments traverse the ball radially at ~106 cm/sec at the driving frequency of ~26 kHz, and drift upward through the ball at ~1 cm/sec. Varying the applied high voltage and frequency caused the number, length, and diameter of the filaments to change. A custom plasma ball was constructed to observe the effects of varying gas pressure and species on the filament structures.

The effect on plasma turbulence of orbit averaging by thermal ions is considered, and illustrated for two modes of potential importance for tokamaks. The effect can reduce the ion response below that in earlier treatments, modifying the predicted mode growth rate, which in turn modifies the turbulent transport. For both modes, the effect modifies earlier transport expressions with a neoclassical factor,'' which makes the scalings of the resultant transport coefficients with plasma current and magnetic field closer to those found experimentally. Additionally, for the trapped electron mode, this mechanism provides a potential explanation of the observed more favorable scaling of {chi}{sub i} with T{sub i} in supershots than in L-modes. 21 refs., 2 figs.

Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

An attempt was made to control global plasma confinement in the Alcator C-Mod tokamak by applying ion cyclotron resonance heating (ICRH) power to the plasma edge in order to deliberately create a minority ion tail loss. In theory, an edge fast ion loss could modify the edge electric field and so stabilize the edge turbulence, which might then reduce the H-mode power threshold or improve the H-mode barrier. However, the experimental result was that edge minority heating resulted in no improvement in the edge plasma parameters or global stored energy, at least at power levels of radio-frequency power is less than or equal to 5.5 MW. A preliminary analysis of these results is presented and some ideas for improvement are discussed.

Edge turbulence images have been made using an ultra-high speed CCD camera on both NSTX and Alcator C-Mod. In both cases, the D-alpha or HeI (587.6 nm) line emission from localized deuterium or helium gas puffs was viewed along a local magnetic field line near the outer midplane. Fluctuations in this line emission reflect fluctuations in electron density and/or electron temperature through the atomic excitation rates, which can be modeled using the DEGAS-2 code. The 2-D structure of the measured turbulence can be compared with theoretical simulations based on 3-D fluid models.

The National Spherical Torus Experiment (NSTX) at Princeton 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 5 MW of neutral beam injection, NBI, at an injection energy of 90 keV and up to 6 MW of high harmonic fast wave, HHFW, at 30 MHz. NSTX has achieved beta T of 32%. A variety of MHD phenomena have been observed to limit eta. NSTX has now begun addressing E scaling, eta limits and current drive issues. During the NBI heating experiments, a broad Ti profile with Ti up to 2 keV, Ti > Te and a large toroidal rotation. Transport analysis suggests that the impurity ions have diffusivities approaching neoclassical. For L-Mode plasmas, E is up to two times the ITER-89P L-Mode scaling and exceeds the ITER-98pby2 H-Mode scaling in some cases. Transitions to H-Mode have been observed which result in an approximate doubling of tE. after the transition in some conditions. During HH FW heating, Te > Ti and Te up to 3.5 keV were observed. Current drive has been studied using coaxial helicity injection (CHI), which has produced 390 kA of toroidal current and HHFW, which has produced H-modes with significant bootstrap current fraction at low Ip, high q and high{sub etap}.

The 2-D radial vs. poloidal structure of edge turbulence in the Alcator C-Mod tokamak [I.H. Hutchinson, R. Boivin, P.T. Bonoli et al., Nuclear Fusion 41(2001) 1391] was measured using fast cameras and compared with 3-D numerical simulations of edge plasma turbulence. The main diagnostic is Gas Puff Imaging (GPI), in which the visible D(subscript alpha) emission from a localized D(subscript 2) gas puff is viewed along a local magnetic field line. The observed D(subscript alpha) fluctuations have a typical radial and poloidal scale of approximately 1 cm, and often have strong local maxima (''blobs'') in the scrape-off layer. The motion of this 2-D structure motion has also been measured using an ultra-fast framing camera with 12 frames taken at 250,000 frames/sec. Numerical simulations produce turbulent structures with roughly similar spatial and temporal scales and transport levels as that observed in the experiment; however, some differences are also noted, perhaps requiring diagnostic improvement and/or additional physics in the numerical model.

The National Spherical Torus Experiment (NSTX) is a spherical tokamak with nominal plasma major radius R(subscript ''0'') = 0.85 m, minor radius a = 0.66 m, and aspect ratio A > 1.28. Typical discharge parameters are plasma current I (subscript ''p'') = 0.7-1.4 MA, toroidal magnetic field B(subscript ''t0'') = 0.25-0.45 Tesla at major radius R(subscript ''0''), elongation = 1.7-2.2, triangularity 0.3-0.5, line-average electron density = 2-5 x 10(superscript ''19'') m(superscript ''-3''), electron temperature T(subscript ''e'')(0) = 0.5-1.5 keV, and ion temperature T(subscript ''i'')(0) = 0.5-2 keV. The NSTX auxiliary heating systems can routinely deliver 4.5 MW of 80-keV deuterium neutral beams and 3 MW of 30-MHz high-harmonic fast-wave power. Kinetic profile diagnostics presently include a 10-channel, 30-Hz multipulse Thomson scattering system (MPTS), a 17-channel charge-exchange recombination spectroscopy (CHERS) system, a 48-chord ultra-soft X-ray (USXR) array, and a 15-chord bolometry array. Initial experiments utilizing auxiliary heating on NSTX have focused on MHD stability limits, confinement trends, studying H-mode characteristics, and performing initial power balance calculations.

The 2-D structure of plasma density turbulence in a magnetically confined plasma can potentially be measured using a Thomson scattering system made from components of the Nova laser of LLNL. For a plasma such as NSTX at PPPL, the laser would form an {approximately}10 cm wide plane sheet beam passing vertically through the chamber across the magnetic field. The scattered light would be imaged by a CCD camera viewing along the direction of the magnetic field. The laser energy required to make 2-D images of density turbulence is in the range 1--3 kJ, which can potentially be obtained from a set of frequency-doubled Nd:Glass amplifiers with diameters in the range of 208--315 mm. A laser pulse width of 100 nsec would be short enough to capture the highest spatial frequency components of the expected density fluctuations.

A method for visually observing tritium deposition on the surface of the Tokamak Fusion Test Reactor (TFTR) deuterium-tritium (D-T) tiles is being investigated at the Princeton Plasma Physics Laboratory. A green phosphor (P31, zinc sulfide: copper) similar to that used in oscilloscope screens with a wavelength peak of 530 nm was positioned on the surface of a TFTR D-T tile. The approximately 600 gram tile, which contains approximately 1.5 Ci of tritium located on the top approximately 1-50 microns of the surface, was placed in a two liter lexan chamber at Standard Temperature and Pressure (STP). The phosphor plates and phosphor powder were placed on the surface of the tile which resulted in visible light being observed, the consequence of tritium betas interacting with the phosphor. This technique provides a method of visually observing varying concentrations of tritium on the surface of D-T carbon tiles, and may be employed (in a calibrated system) to obtain quantitative data.

A new code, Monte Carlo Collisional Stochastic Orbit Retracing (MCCSOR), has been developed to model the alpha particle loss signal as measured by the outer midplane scintillator detector in TFTR. The shadowing effects due to the outer limiters and the detector itself have been included, along with a pitch angle scattering and stochastic ripple diffusion. Shadowing by the outer limiters has a strong effect on both the magnitude and pitch angle distribution of the calculated loss. There is at least qualitative agreement between the calculated results and the experimental data.

Fast ion losses resulting from MHD modes at the Alfven frequency, such as the TAE, have been observed in TFTR. The modes have been driven both by neutral beam ions, at low B{sub T}, and by H-minority ICRF tail ions at higher B{sub T}. The measurements indicate that the loss rate varies linearly with the mode amplitude, and that the fast ion losses during the mode activity can be significant, e.g. up to 10% of the input power is lost in the worst case.

An experiment was performed in TFTR DT plasmas to attempt to destabilize the alpha particle driven Toroidicity-induced Alfven Eigenmode (TAE) by transiently cooling the ions, which should have lowered the ion Landau damping of the TAE modes. Transient cooling perturbations were made during the NBI heating phase of high powered DT supershots using He gas puffs or deuterium (D) or lithium (Li) pellet injection. The ion temperature was successfully lowered from T{sub i}(O) {approx} 20 keV to T{sub i}(O) {approx} 10 keV in about 0.2 sec; however, no signs of alpha-driven TAE modes were observed. Theoretical analyses of these discharges suggested that the alpha pressure required for TAE instability was about a factor of 2--3 greater than actually obtained in this experiment, consistent with the absence of alpha-driven TAE modes.

An Alfven frequency mode (AFM) is very often seen in TFTR neutral beam heated plasmas as well as ohmic plasmas. This quasi-coherent mode is so far only seen on the magnetic fluctuation diagnostics (Mirnov coils). A close correlation between the plasma edge density and the mode activity (frequency and amplitude) has been observed, which indicates that the AFM is an edge localized mode with r/a > 0.85. No direct impact of this mode on the plasma global performance or fast ion loss (e.g., the {alpha}-particles in DT experiments) has been observed. This mode is apparently not the conventional TAE (toroidicity-induced Alfven eigenmodes). The present TAE theory cannot explain the observation. Other possible explanations are discussed.

Alpha particle loss to the wall of TFTR has been measured during the initial TFTR DT run period. These measurements were made with the same lost alpha scintillator detector system used previously for DD fusion products, except for a switch of the scintillator material from zinc sulfide (P31) to yttrium aluminate (P46) to insure a linear response up to the maximum alpha flux expected in DT. The alpha loss signals in DT are {approx} 100 times larger than the DD fusion product loss signals, as expected from the neutron rates and the relative sensitivity to DT vs. DD fusion products.

High energy ion loss during hydrogen minority ICRF heating is measured and compared with the loss of the D-D fusion products. During H minority heating a relatively large loss of high energy ions is observed at 45{degrees} below the outer midplane, with or without simultaneous NBI heating. This increase is most likely due to a loss of the minority tail protons, a possible model for this process is described.

A detailed evaluation of various potential mechanisms for the generation of strong runaway beams during disruptions of largetokamak devices, including TFTR, JET, DIIID and ITER, is performed based on typical operating parameters of these devices and the presently accepted disruption model. The main results include: (1) In the existing devices, the evaporative preicer'' process by itself can lead to sizable runaway beams in disruptions of high-current-medium-to-low-ne discharges. In ITER, such runaways are expected mainly for discharges with ne values sizably smaller than the projected typical ones. (2) Runaway generation also may occur during post-thermal-quench period through the untrapping of trapped hot-thermal electrons remaining from the pre-thermal-quench plasma; this process may be directly important in particular in disruptions of high-T[sub e] discharges with details depending on the time required for reclosure of the magnetic surfaces. Both processes (1) and (2) will occur and be completed mostly during the initial few 100 [mu]sec after the thermal quench. (3) Subsequently, close collisions of runaways with cold plasma electrons generally will lead to an exponential growth ( avalanching'') of runaway populations generated by processes (1) and/or (2) and/or others; this process will be effective in particular during the current quench phase and will continuemore » until the resulting runaway beam will carry essentially all of the remaining discharge current. In presently existing devices, possible avalanche factors of up to 10[sup 2]--10[sup 5] may be expected; in ITER, avalanche factors of up to 10[sup 10]--10[sup 15] -- if not properly suppressed -- are expected to lead to strong runaway beams in most disruptions, except those at particularly high densities. At the same time, avalanching will shift the main part of their energy spectrum down to relatively low energies around 10--20 MeV, and may sizably change the spatial distribution of the runaways.« less

This report summaries the experimental, theoretical, and diagnostic talks presented at the Third Workshop on Alpha Particle Physics in TFTR, which was held at MIT May 28--29, 1992.

In contrast to earlier expectations, it is estimated that generation of runaway electrons from close collisions of existing runaways with cold plasma electrons can be significant even for small electric fields, whenever runaways can gain energies of about 20 MeV or more. In that case, the runaway population will grow exponentially with the energy spectrum showing an exponential decrease towards higher energies.Energy gains of the required magnitude may occur in large Tokamak devices as well as in cosmic-ray generation.

The loss of MeV ions during {sup 3}He ICRH minority heating experiments has been measured using scintillator detectors near the wall of TFTR. The observed MeV ion losses to the bottom (90{degrees} poloidal) detector are generally consistent with the expected first-orbit loss of D-{sup 3}He alpha particle fusion products, with an inferred global reaction rate up to {approx}10{sup 16} reactions/sec. A qualitatively similar but unexpectedly large loss occurs 45{degrees} poloidally below the outer midplane. This additional loss might be due to ICRH tail ions or to ICRH wave-induced loss of previously confined fusion products.

Results obtained from experiments utilizing high power ICRF (ion cyclotron range of frequency) heating to stabilize sawtooth oscillations on TFTR are reviewed. The key observations include existence of a minimum ICRF power required to achieve stabilization, a dependence of the stabilization threshold on the relative size of the ICRF power deposition profile to the q=1 volume, and a peaking of the equilibrium pressure and current profiles during sawtooth-free phases of the discharges. In addition, preliminary measurements of the poloidal magnetic field profile indicate that q on axis decreases to a value of 0.55{plus minus}0.15 after a sawtooth-stabilized period of {approximately}0.5 sec has transpired. The results are discussed in the context of theory, which suggests that the fast ions produced by the ICRF heating suppress sawteeth by stabilizing the m=1 MHD instabilities believed to be the trigger for the sawtooth oscillations. Though qualitative agreement is found between the observations and the theory, further refinement of the theory coupled with more accurate measurements of experimental profiles will be required in order to complete quantitative comparisons.

A transport code (TRANSP) is used to simulate future deuterium-tritium experiments (DT) in TFTR. The simulations are derived from 14 TFTR DD discharges, and the modeling of one supershot is discussed in detail to indicate the degree of accuracy of the TRANSP modeling. Fusion energy yields and {alpha}-particle parameters are calculated, including profiles of the {alpha} slowing down time, average energy, and of the Alfven speed and frequency. Two types of simulations are discussed. The main emphasis is on the DT equivalent, where an equal mix of D and T is substituted for the D in the initial target plasma, and for the D{sup O} in the neutral-beam injection, but the other measured beam and plasma parameters are unchanged. This simulation does not assume that {alpha} heating will enhance the plasma parameters, or that confinement will increase with T. The maximum relative fusion yield calculated for these simulations is Q{sub DT} {approx} 0.3, and the maximum {alpha} contribution to the central toroidal {beta} is {beta}{sub {alpha}}(0) {approx} 0.5%. The stability of toroidicity-induced Alfven eigenmodes (TAE) and kinetic ballooning modes (KBM) is discussed. The TAE mode is predicted to become unstable for some of the equivalent simulations, particularly after the terminationmore » of neutral beam injection. In the second type of simulation, empirical supershot scaling relations are used to project the performance at the maximum expected beam power. The MHD stability of the simulations is discussed.« less

The use of ICRF heating to stabilize the core plasma sawtooth relaxations has been extended to TFTR where such stabilization has been produced at relatively low power in the L Mode regime at moderate density (P{sub RF} = 4 MW, 2.6 MW in helium and deuterium discharges, respectively, for the minority hydrogen ICRF heating regime with {bar n}{sub e}{approx}2.5 {times} 10{sup 13} cm{sup {minus}3}). These results, as in the case of those obtained on JET, are qualitatively consistent with energetic ion stabilization of the m = 1, n = 1 ideal/resistive kink mode. The relevance of sawtooth stabilization to the primary regimes of interest on TFTR -- the high-Q supershot regime and the high density pellet injection regimes -- and on CIT -- the high density ICRF heated regime -- is considered in the context of the present theory and the projected ICRF power deposition characteristics. 35 refs., 11 figs.