ONO, M. (1982). "EFFECT OF LOW-FREQUENCY DENSITY-FLUCTUATIONS ON ION-CYCLOTRON WAVES." Physics of Fluids 25(6): 990-996.
SUGAI, H., H. KOJIMA, et al. (1982). "EXPERIMENTAL-OBSERVATION OF ELECTROSTATIC ION-CYCLOTRON WAVES MODIFIED BY MINORITY IONS." Physics Letters A 92(8): 392-396.
BRODEUR, P., H. W. H. VANANDEL, et al. (1983). "Experimental study of launched ion-acoustic waves in a plasma using continuous wave CO{sub 2} laser scattering." Canadian Journal of Physics 61(8): 1231-41.
A study of coherent density fluctuations in a low density plasma using continuous wave CO{sub 2} laser scattering diagnostics is reported. A simple and direct description of collective scattering theory from monochromatic electrostatic waves is presented. The diagnostic technique is described in detail and its performance is analyzed. Experimental results on externally launched ion-acoustic waves are presented and it is demonstrated that accurate measurements of certain plasma parameters are possible. (19 refs.)
KRAMER, M., K. LUCKS, et al. (1983). "OBSERVATION OF ION PLASMA-WAVES GENERATED BY LOWER HYBRID MODE CONVERSION." Physics Letters A 96(4): 195-198.
ONO, M., K. L. WONG, et al. (1983). "HARMONIC LAUNCHING OF ION BERNSTEIN WAVES VIA MODE TRANSFORMATION." Physics of Fluids 26(1): 298-309.
WONG, K. L., F. SKIFF, et al. (1983). "ECHO PHENOMENON ASSOCIATED WITH LOWER-HYBRID WAVE LAUNCHING." Physics of Fluids 26(10): 2772-2774.
WURDEN, G. A., K. L. WONG, et al. (1983). "Hot-ion effects and mode conversion of the lower-hybrid wave." Physical Review Letters 50(22): 1779-82.
Finite-ion-Larmor-radius modification of lower-hybrid waves in the ACT-1 toroidal plasma at frequencies omega approximately=(4-8) omega {sub ci}, close to the lower-hybrid resonance layer, is observed experimentally. Probe and CO{sub 2} laser scans of the wave amplitude show patterns at frequencies above each ion cyclotron harmonic. Modeling with a hot-ion, electrostatic-ray-tracing code and Fourier reconstruction of the wave form provide evidence for linear mode conversion to hot-plasma waves. (15 refs.)
ABE, H., H. OKADA, et al. (1984). "Resonant heating due to cyclotron subharmonic frequency waves." Physical Review Letters 53(12): 1153-6.
A direct ion heating process which is resonant with the wave at the cyclotron subharmonic frequency, omega ={sup 3}/{sub 2} Omega {sub i}, is discovered through the particle-simulation investigation of the ion Bernstein-wave heating. The particle trapping in phase space due to the wave of an arbitrary cyclotron subharmonic frequency is studied theoretically and numerically confirmed. (6 refs.)
BERS, A. (1984). "RF HEATING AND CURRENT GENERATION IN MAGNETIC FUSION PLASMAS - OVERVIEW AND PERSPECTIVE OF THIS SPECIAL ISSUE." IEEE Transactions On Plasma Science 12(2): 45-47.
ERICKSON, R. M., P. R. FORMAN, et al. (1984). "POLARIMETRIC MEASUREMENT OF PLASMA POLOIDAL MAGNETIC-FIELD VIA HETERODYNE PHASE-SHIFT METHODS." IEEE Transactions On Plasma Science 12(4): 275-280.
JONES, R. (1984). "Magnetoelectrostatic confinement in Tori." Lettere al Nuovo Cimento 41(4): 107-11.
Electrostatic confinement has been used to reduce cross-field diffusion losses in hollow toroidal magnetoplasmas. (9 refs.)
KOJIMA, H., H. SUGAI, et al. (1984). "Dispersion and resonance-cone near ion-ion hybrid resonance." Plasma Physics and Controlled Fusion 26(1B): 359-71.
An experimental and theoretical investigation of the electrostatic wave propagating across the magnetic field at frequencies around the ion-ion hybrid resonance is presented. Firstly, the dispersion relation for the warm electron mode ( omega /k{sub }//<< nu {sub e}) is measured in a simple toroidal plasma, which is composed of helium and neon ions, with the ion concentration ratio controlled. The experimental results and the calculations reveal that even a small fraction ( approximately 1%) of the minority concentration leads to an additional cutoff and hybrid resonance near the minority-cyclotron-resonance frequency. Secondly, the ion-ion hybrid resonance cone of the warm electron mode is observed for the first time. The dependence of the resonance cone angle on the frequency and the ion concentration ratio is observed to agree with the theoretical predictions. (16 refs.)
ONO, M., G. A. WURDEN, et al. (1984). "Efficient ion heating via finite-Larmor-radius ion-cyclotron waves in a plasma." Physical Review Letters 52(1): 37-40.
Ion heating by externally launched ion Bernstein waves is investigated in the ACT-1 hydrogen plasma. Detailed measurements of wave absorption and of the ion temperature profiles have clearly identified various heating layers near the ion-cyclotron harmonics of deuterium-like and tritiumlike ions. The observed bulk ion heating with heating quality factor of 10 eV/W*(10{sup 10} cm{sup -3}) and the power balance estimates suggest excellent overall efficiency for finite-Larmor-radius-ion-cyclotron resonance-frequency heating. (12 refs.)
OWENS, T. L., F. W. BAITY, et al. (1984). "Ion heating in the range of high ion cyclotron harmonics on EBT." Nuclear Fusion 24(12): 1561-72.
Wave heating of ions is obtained in Elmo Bumpy Torus (EBT) experiments under conditions where 18>or approximately= omega / Omega {sub i}>or approximately=3. Absorption of wave energy in the plasma centre is demonstrated. The fast magnetosonic wave is observed on electrostatically shielded loop probes placed at the edge of the plasma, but it is shown that the fast wave cannot directly heat the ions in EBT. The experiments suggest, however, that the waves that do produce the ion heating are coupled to the fast wave. The possibility that the heating is due to excitation of electrostatic waves is investigated theoretically. (21 refs.)
PARK, H., N. C. LUHMANN, et al. (1984). "OBSERVATION OF 2ND-HARMONIC ION BERNSTEIN WAVES EXCITED BY FAST-WAVE MODE CONVERSION IN THE MICROTOR TOKAMAK." Physical Review Letters 52(18): 1609-1612.
SKIFF, F., M. ONO, et al. (1984). "PARAMETRIC-EXCITATION OF ION BERNSTEIN WAVES BY A FAST WAVE ANTENNA IN THE ION-CYCLOTRON FREQUENCY-RANGE." Physics of Fluids 27(5): 1051-1054.
SUGAI, H., H. KOJIMA, et al. (1984). "ION SPECIES COMPOSITION OF A BOUNDARY-LAYER HYDROGEN PLASMA." Journal of Nuclear Materials 128(DEC): 169-171.
WATT, R. G. and E. M. LITTLE (1984). "RFP ENERGY EVOLUTION AND CONTAINMENT IN THE PRESENCE OF GIANT SAWTEETH." Physics of Fluids 27(4): 784-786.
WONG, K. L. and M. ONO (1984). "Effects of ion cyclotron harmonic damping on current drive in the lower hybrid frequency range." Nuclear Fusion 24(5): 615-26.
The ion cyclotron harmonic damping effects on slow and fast waves in the lower hybrid frequency range for Tokamak reactor parameters are studied. Inclusion of the higher-order terms in the hot-plasma dielectric tensor introduces ion cyclotron harmonic damping; these terms also contribute to the real part of the dispersion relation and affect the wave trajectories. However, wave absorption by 15 keV deuterium and tritium ions can be avoided by choosing the slow-wave frequency above the lower-hybrid frequency and the fast-wave frequency below the loser frequency. But preliminary estimates show that energetic alpha particles tend to absorb both the slow and the fast waves. This absorption may become a serious obstacle to fusion-reactor current drive in the lower hybrid frequency range. (23 refs.)
ARIMOTO, H., S. MASAMUNE, et al. (1985). "Asymmetric flux generation and its relaxation in reversed field pinch." Journal of the Physical Society of Japan 54(4): 1232-5.
The toroidally asymmetric flux enhancement ('dynamo effect') and the axisymmetrization of the enhanced fluxes that follows in the setting up phase of reversed field pinch are investigated on the STP-3(M) device. A rapid increase in the toroidal flux generated by the dynamo effect is first observed near the poloidal and toroidal current feeders. Then, this inhomogeneity of the flux propagates toroidally towards the plasma current. The axisymmetrization of the flux is attained just after the maximum of plasma current. The MHD activities decrease significantly after this axisymmetrization and the quiescent period is obtained. (7 refs.)
DAGAZIAN, R. Y. (1985). "NONLINEAR RESISTIVE HELICAL STATES IN REVERSED-FIELD PINCHES." Physics of Fluids 28(7): 2209-2218.
GOREE, J., M. ONO, et al. (1985). "Fast wave current drive." AIP Conference Proceedings(129): 65-6.
Experiments on the fast wave in the range of high ion cyclotron harmonics in the ACT-1 device show that current drive is possible with the fast wave just as it is for the lower hybrid wave, except that it is suitable for higher plasma densities. A 140 degrees loop antenna launched the high ion cyclotron harmonic fast wave ( omega / Omega =0(10)) into a He{sup +} plasma with n{sub e} approximately=4*10{sup 12} cm{sup -3} and B=4.5 kG. Probe and magnetic loop diagnostics and FIR laser scattering confirmed the presence of the fast wave, and the Rogowski loop indicated that the circulating plasma current increased by up to 40 A with 1 kW of coupled power, which is comparable to lower hybrid current drive in the same device with the same unidirectional fast electron beam used as the target for the RF. A phased antenna array would be used for FWCD in a tokamak without the E-beam. (4 refs.)
GOREE, J., D. K. MANSFIELD, et al. (1985). "Far-infrared laser scattering in the ACT-I toroidal device." Journal of Vacuum Science & Technology A (Vacuum, Surfaces, and Films) 3(3): 1074-1076.
A far-infrared laser scattering diagnostic has been built for the ACT-I toroidal device. The optical system uses a passively stabilized 447- mu m CH{sub 3}I laser. A polyethylene etalon is the beam splitter. The vacuum windows are plastic (TPX), which the authors found to have the vacuum property Q 6.5*10{sup -9} Torr 1/s/cm{sup 2}. Using paraboloidal and ellipsoidal mirrors for detection optics improves the signal strength and allows a better RF enclosure design for the detector. The diagnostic was tested by scattering from an ion Bernstein wave, a technique which can be used for ion temperature diagnostics. (10 refs.)
GOREE, J., M. ONO, et al. (1985). "OBSERVATION OF THE BACKWARD ELECTROSTATIC ION-CYCLOTRON WAVE." Physics of Fluids 28(9): 2845-2847.
JAHODA, F. C., B. T. ANDERSON, et al. (1985). "Application of optical phase conjugation to plasma diagnostics." Review of Scientific Instruments 56(5): 953-957.
Several possibilities for plasma diagnostics provided by optical phase conjugation and, in particular, self-pumped phase conjugation in barium titanate (BaTiO{sub 3}) are discussed. These include placing a plasma within a dye laser cavity equipped with a phase conjugate mirror for intracavity absorption measurements, time differential refractometry with high spatial resolution, and simplified real-time holographic interferometry. The principles of phase conjugation with particular reference to photorefractive media and the special advantages of self-pumped phase conjugation are reviewed prior to the discussion of the applications. Distinctions are made in the applications between those for which photorefractive conjugators are essential and those for which they only offer experimental simplification relative to other types of phase conjugators. (15 refs.)
JONES, R. (1985). "Particle injection for magnetoelectrostatic confinement." Lettere al Nuovo Cimento 42(3): 117-22.
The electrostatic-potential profiles required for magnetoelectrostatic-confinement configurations can be maintained by means of direct particle injection. (12 refs.)
KOJIMA, H., H. KAKO, et al. (1985). "Pressure dependence of carbon film coating by a toroidal methane discharge." Japanese Journal of Applied Physics, Part 1 (Regular Papers & Short Notes) 24(11): 1432-5.
A pure methane discharge under a toroidal magnetic field has been used to deposit thin carbon layers in situ onto the walls of a toroidal vessel, with a view to application to magnetic fusion devices. Hard carbon films with high electrical resistivity were formed on stainless steel, silicon, and NaCl substrates, in a wide range of methane pressure (10{sup -4} approximately 10{sup -2} Torr). The film properties and structures were examined by infrared absorption, X-ray photoemission analysis, electron microscopy (TEM, TED, SEM), and so forth. The results show that the films predominantly consist of tetrahedrally bonded carbon (sp{sup 3} bonding). Both the hydrogen atom concentration in the films and the binding energy of the carbon atoms increase with increasing methane pressure. Also, the size of microcrystallites on a background consisting of an amorphous structure grows with increasing pressure.
KONDOH, Y., K. MIYASHITA, et al. (1985). "Numerical study of reversed field pinch equilibria." Journal of the Physical Society of Japan 54(8): 2927-33.
Numerical results on the reversed-field-pinch (RFP) equilibria are presented with use of a partially relaxed state model derived from an energy principle. The relaxation process is assumed to be dominated by the helical tearing mode with a single helicity. The profiles of current densities concentrate more to the magnetic axis in the relaxed state which is dominated by the mode with the lower n toroidal mode number. The maximum available beta value with respect to the Suydam criterion is higher in the relaxed state dominated by the higher n mode. Numerical results suggest that experimental RFP plasma has a tendency to relax to the partially relaxed state by the higher n dominant mode. (19 refs.)
MASSEY, R. S., R. G. WATT, et al. (1985). "STATUS OF THE ZT-40M REP EXPERIMENTAL PROGRAM." Fusion Technology 8(1): 1571-1580.
OKUDA, H., R. HORTON, et al. (1985). "EFFECTS OF BEAM-PLASMA INSTABILITY ON CURRENT DRIVE VIA INJECTION OF AN ELECTRON-BEAM INTO A TORUS." Physics of Fluids 28(11): 3365-3379.
ONO, M., T. WATARI, et al. (1985). "Ion-Bernstein-wave heating in the JIPPT-II-U tokamak plasma." Physical Review Letters 54(21): 2339-42.
Ion-Bernstein-wave heating is investigated in the JIPPT-II-U tokamak plasma, n{sub e} approximately=1.5*10{sup 13} cm{sup -3}, T{sub eo} approximately=700 eV, and T{sub i0} approximately=300 eV for P{sub rf}<or=100 kW. When the 3/2 Omega {sub H} layer is placed near the plasma minor axis, the bulk-hydrogen-ion temperature shows a significant rise, Delta T{sub i perpendicular to }<or=700 eV and Delta T{sub i}//<or=30 eV. The ion-heating dependence on the magnetic field and RF power suggests a presence of a direct bulk-hydrogen heating mechanism at omega approximately=3/2 Omega {sub H}. (11 refs.)
PARK, H., P. S. LEE, et al. (1985). "Tokamak ion temperature determination via CW far-infrared laser scattering from externally excited ion Bernstein waves." Nuclear Fusion 25(10): 1399-411.
The authors describe a successful proof-of-principle experimental determination of tokamak ion temperature using cw far-infrared (FIR) collective laser scattering from externally excited ion Bernstein waves. It is shown that a viable wave excitation technique for tokamak plasmas is mode conversion of an externally launched fast Alfven wave. A fit of the experimentally determined ion Bernstein wave dispersion to the temperature-dependent theoretical dispersion yields the local ion temperature. Partial ion temperature profiles (chord-averaged) have been obtained with temperature values consistent with charge-exchange measurements. (46 refs.)
PARK, H., P. S. LEE, et al. (1985). "Tokamak ion temperature determination via CW far-infrared laser scattering." Review of Scientific Instruments 56(5): 922-924.
A successful proof-of-principle experimental determination of ion temperature in a Tokamak plasma via CW far-infrared (FIR) collective laser scattering from ion Bernstein waves is reported. The Bernstein waves are excited via mode conversion of an externally launched fast Alfven wave at the second-harmonic cyclotron layer. A fit of the experimentally determined ion Bernstein wave dispersion to the temperature-dependent theoretical dispersion yields the local ion temperature. Partial ion temperature profiles (chord averaged) have been obtained with temperature values consistent with charge exchange measurements. Single-shot measurements should be available in the near future with the use of the multichannel scattering systems and high-power laser sources developed at UCLA. (12 refs.)
PORKOLAB, M. (1985). "Nonlinear Landau heating by ion-Bernstein waves in magnetically confined fusion plasmas." Physical Review Letters 54(5): 434-7.
It is shown that during ion-Bernstein-wave heating experiments, nonlinear ion Landau damping absorbs efficiently the incident Bernstein waves in present-day tokamak and tandem-mirror plasmas. Further, this nonlinear absorption will dominate absorption by minority (impurity) ions. (12 refs.)
SATO, M. (1985). "DISPERSION-RELATION FOR BERNSTEIN WAVES USING A NEW TRANSFORMATION FOR THE MODIFIED BESSEL-FUNCTION." Journal of Plasma Physics 34: 417-425.
SCHNACK, D. D., E. J. CARAMANA, et al. (1985). "3-DIMENSIONAL MAGNETOHYDRODYNAMIC STUDIES OF THE REVERSED-FIELD PINCH." Physics of Fluids 28(1): 321-333.
SHOUCRI, M., I. P. SHKAROFSKY, et al. (1985). "The equilibrium and time evolution of the parameters of the Tokamak de Varennes." Canadian Journal of Physics 63(11): 1473-80.
The equilibrium and time evolution of the density, current, and temperature profiles for a Tokamak are numerically analyzed using the one-dimensional transport code VARTOK 1. The numerical model contains the full neoclassical transport, with coefficients covering transitions between regimes, and includes the Ware pinch and bootstrap currents, empirical anomalous particle and electron heat transports based on Alcator scaling, neutral particle transport, oxygen impurity, and diffusion of the poloidal magnetic field. The solution uses a time- and space-centered Crank-Nicolson scheme. The equilibrium profiles are first determined for a constant current of 200 kA by running the code at constant current. Then the current is ramped down to zero in 20 ms, and the time evolution of the different parameters are followed. (24 refs.)
SMITH, R. A., J. A. KROMMES, et al. (1985). "SATURATION OF DRIFT INSTABILITIES BY ELECTRON DYNAMICS." Physics of Fluids 28(4): 1069-1076.
SUGAI, H., H. TOYODA, et al. (1985). "Ion and radical contributions to hydrogenated amorphous silicon film formation in a DC toroidal discharge." Applied Physics Letters 46(11): 1048-50.
Hydrogenated amorphous silicon films are formed in a low-pressure DC toroidal discharge. The good plasma confinement by a toroidal magnetic field ensures considerable growth rate of the films even at a low pressure (0.1-10 mTorr). The contributions of ions and radicals to a-Si:H film growth are clearly separated by the control of substrate bias and gas pressure. The films formed mainly by ion flux indicate large values of optical band gap and hydrogen content, in contrast with films formed by neutral radicals only. In addition, the spectrum of dominant IR absorption is observed to vary with the ratio of ion to radical contributions. (12 refs.)
SWANSON, D. G. (1985). "RADIO-FREQUENCY HEATING IN THE ION-CYCLOTRON RANGE OF FREQUENCIES." Physics of Fluids 28(9): 2645-2677.
SY, W. N. C., T. AMANO, et al. (1985). "Slow-wave antenna coupling to ion Bernstein waves for plasma heating in ICRF." Nuclear Fusion 25(7): 795-803.
The coupling of ICRF power from a slow-wave antenna to a plasma with finite temperature is examined. A heuristic model, allowing explicit representations of ion Bernstein waves, fast waves and slow waves, is used to clarify how the antenna power is partitioned into the various wave energy fluxes. This model is complemented quantitatively by a more elaborate and realistic computer model. It is shown that such antennas can be highly efficient in transferring most of the antenna power directly to ion Bernstein waves, with only a very small fraction going into fast waves. The potentiality of this coupling scheme for plasma heating in ICRF is briefly discussed. (18 refs.)
WERLEY, K. A., R. A. NEBEL, et al. (1985). "TRANSPORT DESCRIPTION OF THE RISE TIME OF SAWTOOTH OSCILLATIONS IN REVERSED-FIELD PINCHES." Physics of Fluids 28(5): 1450-1453.
APPERT, K., G. A. COLLINS, et al. (1986). "Theory of MHD waves." Plasma Physics and Controlled Fusion 28(1A): 133-45.
Reviews recent developments in the theory of MHD waves in connection with radio-frequency heating in the Alfven wave and ion cyclotron ranges of frequencies (AWRF, ICRF). The account focuses on the discussion of full wave solutions and the oscillation spectra in bounded, generally inhomogeneous, plasmas. Original results are presented concerning forced ICRF oscillations in a current-carrying torus. The effects on the wave structures of the equilibrium current, the size of the device, the minority concentration and the phenomenological damping are investigated. The resonant surfaces coincide with the magnetic surfaces as in the AWRF. The poloidal extension of the resonance is small in cases where a WKB approach is permitted so that there is no conflict between full wave solutions and the WKB method. (66 refs.)
BODIN, H. A. B., R. A. KRAKOWSKI, et al. (1986). "THE REVERSED-FIELD PINCH - FROM EXPERIMENT TO REACTOR." Fusion Technology 10(3): 307-353.
BORA, D., K. SATYANARAYANA, et al. (1986). "EFFECT OF RADIOFREQUENCY FIELD ON TOROIDAL PLASMA PARAMETERS." Pramana 27(4): 549-561.
DAHLBURG, J. P., D. MONTGOMERY, et al. (1986). "Turbulent relaxation to a force-free field-reversed state." Physical Review Letters 57(4): 428-31.
The evolution of nonequilibrium initial conditions of an incompressible magnetohydrodynamic Z pinch is described by a three-dimensional, pseudospectral numerical code Magnetohydrodynamic turbulence develops in the resistive, nonviscous magnet fluid, resulting in the selective decay of the energy relative to the magnetic helicity, at Lundquist numbers of only a few hundred. An interior force-free region grows with time and achieves spontaneous reversal of the toroidal magnetic field at the wall, without the necessity of an external electric field. (21 refs.)
FINN, J. M. (1986). "RELAXATION MODEL FOR HELICITY INJECTION IN OHMIC AND OSCILLATING CURRENT-DRIVEN REVERSED FIELD PINCH DISCHARGES." Physics of Fluids 29(8): 2630-2638.
JARDIN, S. C., A. JANOS, et al. (1986). "The effect of a column inductive transformer on the S-1 Spheromak." Nuclear Fusion 26(5): 647-65.
The authors predict the consequences of adding a cylindrical inductive current transformer on the small major radius side of the present S-1 Spheromak experiment. Axisymmetric modelling with only classical dissipation shows an increase of toroidal current and a shrinking and hollowing of the current channel, conserving toroidal flux. These unstable profiles undergo helical reconnection, conserving helicity K= integral A.B d tau while increasing the toroidal flux and decreasing the poloidal flux so that the plasma relaxes toward the Taylor state. This flux rearrangement is modelled by a new current viscosity term in the mean field Ohm's law which conserves helicity and dissipated energy. (19 refs.)
KRAKOWSKI, R. A., R. L. HAGENSON, et al. (1986). "Compact reversed-field pinch reactors (CRFPR)." Nuclear Engineering and Design/Fusion 4(1): 75-120.
The unique confinement properties of the poloidal-field-dominated reversed-field pinch (RFP) are exploited to examine physics and technical issues related to a compact high-power-density fusion reactor. This resistive-coil, steady-state, toroidal device would use a dual-media (i.e. two separate coolants) power cycle that would be driven by a fusion power core (FPC, i.e. plasma chamber, first wall, blanket, shield, and coils) having a power density and mass approaching pressurized-water-fission reactor values. A 1000-MWe(net) base case is selected from a comprehensive trade-off study to examine technological issues related to operating a high-power-density FPC. A general rationale outlining the need for improved fusion concepts is given, followed by a description of the RFP principle, a detailed systems and trade-off analysis, and a conceptual FPC design for the approximately 20-MW/m{sup 2} (neutrons) compact RFP reactor, CRFPR(20). Key FPC components are quantified, and full power-balance, thermal, and mechanical FPC integrations are given. (80 refs.)
KUSANO, K. and T. SATO (1986). "Simulation study of the self-reversal process in the reversed-field pinch based on a nonlinearly driven reconnection model." Nuclear Fusion 26(8): 1051-61.
The self-reversal process in the reversed-field pinch is studied in detail by means of a resistive magnetohydrodynamic simulation. It is confirmed that self-reversal can be caused by nonlinearly driven reconnection resulting from the m=1 global kink instability, as was previously propose by the authors The dependence of the degree of field reversal on the pinch parameter theta and on the instability mode (resonant and nonresonant) is examined. The results are consistent with theoretical predictions. Taylor's conjecture that the total helicity is a better conserved quantity than the total magnetic energy during the relaxation process is numerically confirmed. It is found that this conjecture can be consistently explained by the nonlinearly driven reconnection model. It is also found that the single helicity relaxation process has a definite energy offset from Taylor's minimum energy state. Hence, a totally relaxed state cannot be achieved through the single helicity relaxation process. Finally, the dependence of the reversal process on the resistivity is examined. (33 refs.)
KUTHI, A., H. ZWI, et al. (1986). "RACETRACK: novel device for basic research on magnetized plasmas." Review of Scientific Instruments 57(11): 2720-3.
The construction and operation of a novel magnetized plasma device are described. The device combines characteristics of both linear mirror and toroidal confinement. It opens up new areas of basic plasma physics research. Examples of experiments on potential formation are presented. (10 refs.)
MALACARNE, M. and I. H. HUTCHINSON (1986). "Soft X-ray emission fluctuations in the HBTX1A reversed field pinch." Plasma Physics and Controlled Fusion 28(5): 823-37.
X-ray emission fluctuations from the HBTX1A reversed field pinch have been investigated with an array of silicon surface barrier diodes using statistical analysis techniques. Despite their apparent randomness a new correlation technique involving all the signals has been used to identify the dominant activity at frequencies <or approximately=20 kHz as global modes with poloidal mode number m=0 and 1. This technique also allows the reconstruction of the radial profile of the modes and the calculation of their power spectrum. By using single detectors displaced in the toroidal direction it is shown that the m=0 instabilities have long toroidal wavelength (n approximately 0) and do not propagate in this direction, whereas the m=1 modes have n>>1. Local turbulence becomes the dominant component at higher frequencies; even so correlation lengths parallel to B are much longer than across. Correlating the X-ray diodes with magnetic edge coils enables a direct comparison of the two types of fluctuations and shows that one of the m=0 modes propagates radially outwards. (15 refs.)
MATTOO, S. K. and N. VENKATARAMANI (1986). "Dependence of striking characteristics of plasma discharge on filament orientation in a toroidal device." Journal of Applied Physics 60(8): 2762-5.
In a torus with a toroidal magnetic field it is found that the discharge cannot be struck between the filament and the vessel wall if the filament (the source of thermionic electrons) is oriented parallel to the magnetic field. This observation and other experimental results lead to a conclusion that electric flux plays a crucial role in the determination of the striking characteristics of the discharge. (15 refs.)
MCNEILL, D. H. (1986). "Molecular origin of background light in Thomson scattering measurements." Review of Scientific Instruments 57(11): 2724-8.
The plasma background light in Thomson scattering measurements is often far higher than expected for a pure hydrogen plasma. The spectral distribution of light from three plasmas (duration: 1 ms to steady state; electron density: below 10{sup 12} to over 10{sup 14} cm{sup -3}; temperature: below 20 to over 1000 eV) is studied. Signal-to-noise and intensity data from the Thomson scattering systems used on them are compared with analytic estimates to show that in two of these plasmas molecular light dominates the spectrum, while in the other, molecular light is present, but bremsstrahlung is usually more intense. Knowledge of the mechanism for background light can aid in designing detection systems for Thomson scattering and provide information on the neutral species composition and effective charge of the plasma. (23 refs.)
OKADA, H., H. ABE, et al. (1986). "COMPUTATIONAL STUDY ON EXCITATION, PROPAGATION, AND 2-SPECIES-ION-PLASMA HEATING DUE TO THE ION-BERNSTEIN WAVE." Physics of Fluids 29(2): 489-500.
SCHNACK, D. D., D. C. BARNES, et al. (1986). "Numerical simulation of reversed-field pinch dynamics." Computer Physics Communications 43(1): 17-28.
The reversed-field pinch (RFP) is presented as a driven nonlinear dynamical system that evolves on time scales long compared with the fastest normal modes of the system. Numerical methods for the simulation of these processes, and examples of their application to experimentally observable phenomena, are presented. (21 refs.)
SHINOHARA, S., O. NAITO, et al. (1986). "Localized electron heating experiments by ion Bernstein wave in the TNT-A tokamak." Journal of the Physical Society of Japan 55(8): 2648-52.
Plasma heating by ion Bernstein waves in the range of 2 omega {sub D}<or= omega <or=3 omega {sub D} is investigated in deuterium dominant plasma of the TNT-A tokamak. The localized electron heating is observed at the harmonic (3 omega {sub D}) and subharmonic (2.5 omega {sub D}) resonance layers, while the electron heating on the whole plasma region is observed at omega =2 omega {sub D}. It is also shown that the heating is efficient and the heating layer is localized by ion Bernstein waves in comparison with fast magnetosonic waves.
SHINOHARA, S., O. NAITO, et al. (1986). "Dependence of resistive loading on antenna orientation for ICRE waves in a tokamak." Nuclear Fusion 26(8): 1097-102.
The parameter dependence of antenna loading resistance for ICRF waves using a T-shaped rotatable antenna is investigated in the TNT-A tokamak. When the antenna current is parallel to the toroidal field, the loading resistance increases with the mean plasma density, but depends weakly on the toroidal field, the ratio of hydrogen to deuterium concentrations and the type of the Faraday shield. Similar results with a lower value of the loading resistance are obtained in the case where the antenna current is perpendicular to the toroidal field, using the same rotatable antenna. With the antenna current parallel to the toroidal field, the loading resistance for a low parallel refractive index n{sub z} is larger than that for high n{sub z}. (26 refs.)
TANAKA, M. and T. SATO (1986). "MACROSCALE PARTICLE SIMULATION OF RELATIVISTIC ELECTRON-BEAM INJECTION INTO A MAGNETIZED PLASMA CHANNEL." Physics of Fluids 29(11): 3823-3831.
UEDA, Y., O. NAITO, et al. (1986). "Ion cyclotron heating mechanism of fast magnetosonic wave in the TNT-A Tokamak." Journal of the Physical Society of Japan 55(3): 806-11.
Ion heating mechanism of fast magnetosonic wave in the ion cyclotron range of frequency (ICRF) has been studied in the TNT-A tokamak. Ion temperature measurements have been made with mass separated neutral particle analyser. The effects of the fundamental proton cyclotron resonance layer (ICR) and the two-ion hybrid resonance layer (HBR) on ion heating are observed. In the case of hydrogen minority (ICR and HBR in the plasma), where the ratio of hydrogen-to-deuterium concentration, n{sub H}/n{sub D}, is 0.05, the increase in hydrogen temperature is larger than that in deuterium, and most of the RF power is absorbed by hydrogen. (15 refs.)
WEBER, P. G. and S. MASAMUNE (1986). "High-intensity lithium beam for Zeeman spectroscopy." Review of Scientific Instruments 57(8): 2032-2034.
The Zeeman splitting of neutral lithium introduced to a hot plasma as a beam may be used to determine both the magnitude and direction of the local magnetic field. Dye lasers can improve the sensitivity of the method, either by resonance fluorescence or by intracavity absorption. A limit on the applicability of this diagnostic is set by the lithium density in the plasma. A high density, 80-kV lithium beam has been developed to access plasmas of larger dimension and/or higher densities, or to permit measurements with better time resolution than with previous beams. The beam utilizes a high density lithium ion source. Diagnostics include 2-D ion beam profiling, Faraday cup, magnetic analyzer, and a 'pepperpot' probe for measurements of beam intensity, composition, and divergence. Beam currents can exceed 10 mA in a <or=2-cm-diameter beam with a divergence of several milliradians. (14 refs.)
WEST, W. P., D. M. THOMAS, et al. (1986). "Diagnostic instrument for the measurement of poloidal magnetic fields in Tokamaks." Review of Scientific Instruments 57(8): 1552-1556.
A diagnostic system has been developed to measure the direction of the local magnetic field within the plasma of a Tokamak by making use of Zeeman splitting in an injected neutral lithium beam. A linearly polarized dye laser is used to pump the pi component of the {sup 2}S-{sup 2}P resonant transition. The laser polarization is rotated at a fixed frequency, and phase analysis of the subsequent fluorescence identifies the local field direction. The instrument can monitor many radial points within the plasma simultaneously, allowing a current density profile to be determined on a single shot basis. (22 refs.)
WROBLEWSKI, D., H. W. MOOS, et al. (1986). "Zeeman effect polarimetry of Ti XVII 3834 AA line in Texas Experimental Tokamak." Applied Physics Letters 48(1): 21-3.
The authors present the first measurement of a Tokamak plasma internal magnetic field based on the analysis of polarization of Zeeman components of a magnetic dipole transition in heavy impurity ion. The circular polarization of the Ti XVII 3834 AA line is studied. The values of the component of the toroidal magnetic field in the direction of observation are measured for several observation angles and good agreement with expected values is found. The result indicates that the method may be useful for measuring the local magnetic field and, in particular, the poloidal component of the field in a Tokamak. (12 refs.)
ANTONI, V. and S. ORTOLANI (1987). "RELAXATION PHENOMENA AND ON AXIS-Q LIMIT IN A REVERSED FIELD PINCH." Physics of Fluids 30(5): 1489-1496.
BASTASZ, R. and T. E. CAYTON (1987). "COLLECTOR PROBE MEASUREMENTS OF DEUTERIUM AND IMPURITY FLUXES IN THE ZT-40M REVERSED-FIELD PINCH." Journal of Nuclear Materials 147: 476-479.
CARLSON, A. W. (1987). "A SEARCH FOR LOWER-HYBRID-DRIFT FLUCTUATIONS IN A FIELD-REVERSED CONFIGURATION USING CO2 HETERODYNE SCATTERING." Physics of Fluids 30(5): 1497-1509.
CAYTON, T. E., J. N. DOWNING, et al. (1987). "PLASMA SURFACE INTERACTIONS IN THE ZT-40M REVERSED-FIELD PINCH." Journal of Nuclear Materials 147: 71-80.
HOWELL, R. B., J. C. INGRAHAM, et al. (1987). "ASYMMETRIC MAGNETIC-FLUX GENERATION, M=1 ACTIVITY, AND EDGE PHENOMENA ON A REVERSED-FIELD PINCH." Physics of Fluids 30(6): 1828-1838.
MITARAI, O., S. W. WOLFE, et al. (1987). "Stable AC Tokamak discharges in the STOR-1M device." Nuclear Fusion 27(4): 604-7.
Stable and clean AC tokamak discharges have been achieved in the STOR-1M device. The plasma current is reversed from +4.1 kA to -4.0 kA within 1.9 ms. During the reversal, no disruptive behaviour is observed, the loop voltage changes smoothly from +1 V to -5 V without any spike, and impurities are not released. An electron density of (2-4)*10{sup 12} cm{sup -3} is maintained during current reversal. The possibility of continuous tokamak operation with a low frequency alternating plasma current is discussed. (8 refs.)
NAITO, O., S. SHINOHARA, et al. (1987). "Observation of the ion Bernstein wave propagation by magnetic probe in a tokamak." Journal of the Physical Society of Japan 56(8): 2988-9.
Ion Bernstein wave propagation is observed by magnetic probe in the TNT-A tokamak. The obtained radial B{sub z} profile is in good agreement with that from ray tracing calculation.
NAKAJIMA, N. (1987). "NONLINEAR PHENOMENA DUE TO AN M=1 NONRESONANT CURRENT-DRIVEN MODE IN RFP WITH RELATIVELY HIGH S NUMBER." Journal of the Physical Society of Japan 56(5): 1738-1749.
NAKAJIMA, N. (1987). "Analytical studies on the nonlinear oscillation due to an m=1 non-resonant current-driven mode in RFP plasmas." Journal of the Physical Society of Japan 56(11): 3911-28.
Analytical studies are presented on the nonlinear oscillation due to a single non-resonant current-driven mode with poloidal mode number of m=1 in the vicinity of a marginally stable state in RFP plasmas. A perturbation expansion with respect to the difference of the toroidal wave number from one at the marginally stable state is used together with the multiple-time-scale method. The system of equations governing the time development of the amplitude of the m=1 mode is obtained, which indicates that the nonlinear oscillation is possible in RFP plasmas with a relatively high S number (S>or approximately=10{sup 4}) and that the oscillation is interpreted as an amplitude oscillation of the m=1 mode. The analytical results by using the perturbation theory well explain those of the numerical simulation.
NEWTON, A. A., T. JARBOE, et al. (1987). "CURRENT AND HEAT-FLUX TO THE WALL AND ELECTRON-DENSITY CONTROL IN REVERSED FIELD PINCHES." Journal of Nuclear Materials 147: 487-495.
OGAWA, Y., K. KAWAHATA, et al. (1987). "Characteristics of ion Bernstein wave heating on the JIPP T-IIU tokamak." Nuclear Fusion 27(9): 1379-88.
Ion Bernstein wave (IBW) heating has been examined on the JIPP-T-IIU tokamak under two different conditions referred to as Mode-I and Mode-II. In the Mode-I regime, a wave is launched on an IBW branch between the third and fourth cyclotron harmonics of deuterium ions. In the Mode-II regime, a wave is launched on a branch between the second and third cyclotron harmonics. These two modes show quite different heating characteristics. The causes of this difference are analysed by using a simple model to determine the k{sub }// spectrum of the excited wave and by applying a ray tracing code. In connection with the Mode-I experiment as discussed in a previous report (1985), two important new experimental results are obtained. It is shown that an IBW heats the core of the plasma rather than causing plasma-edge interaction, as anticipated. It is also shown that the energy tail of the hydrogen ions is higher than that of the deuterium ions, which indicates that the responsible heating mechanisms are different. (18 refs.)
OIEN, A. H. (1987). "On steady-state classical transport in collision-dominated toroidal systems." Journal of Plasma Physics 38: 245-262.
Starting from classical transport theory, equations for particle density, particle momentum and electron and ion temperatures are derived for steady-state, toroidal plasma configurations in a parameter regime like that of ACT-I. A set of simplified equations for particle density and electron and ion temperatures are solved numerically. Radial density and temperature profiles are shown and compared with experiments. (6 refs.)
OKADA, H., H. ABE, et al. (1987). "COMPUTATIONAL STUDY ON PROPAGATION AND 2-SPECIES-ION-PLASMA HEATING DUE TO THE ION-BERNSTEIN WAVE." Plasma Physics and Controlled Fusion 29(6): 743-757.
OKUDA, H., R. HORTON, et al. (1987). "PROPAGATION OF A NONRELATIVISTIC ELECTRON-BEAM IN A PLASMA IN A MAGNETIC-FIELD." Physics of Fluids 30(1): 200-208.
ONO, M., G. J. GREENE, et al. (1987). "Steady-state tokamak discharge via DC helicity injection." Physical Review Letters 59(19): 2165-8.
A tokamak discharge has been formed and maintained through helicity injection, by use of only an external DC low-energy electron beam. The discharge, in a 5-kG toroidal field, evolved to a steady-state circular cross section with q(a)=10, q(0)=4, which was maintained for more than 400 L/R periods, the time (60 msec) limited by the cathode bias supply. The density profile reached a line-averaged n{sub e}=2*10{sup 13} cm{sup -3}, while T{sub e}=25 eV and T{sub i}=15 eV. Strong central peaking occurs and inward pinching is indicated for both density and current. (22 refs.)
PEASE, R. S. (1987). "THE STATUS OF FUSION-RESEARCH." Plasma Physics and Controlled Fusion 29(10A): 1171-1186.
SHIINA, S., K. SAITO, et al. (1987). "The measurement of soft X-rays from the STP-3(M) reversed-field pinch plasma." Journal of the Physical Society of Japan 56(4): 1282-5.
In STP-3(M) reversed-field pinch plasma, the bursts of soft X-rays are periodically observed in association with the toroidal flux regeneration (dynamo action) using surface barrier diode (SBD). The soft X-ray bursts result from high energy electrons colliding with the surface of metal liner and limiters. The energy of colliding electrons analyzed by the absorber method assuming a monochromatic distribution, is found to have a distribution of 2 approximately 8 keV. The energy of 8keV is observed at about 100 mu s before the termination of the plasma current. The mechanisms of the acceleration and the anomalous diffusion for these high-energy electrons are discussed. (12 refs.)
SHINOHARA, S., O. NAITO, et al. (1987). "Parameter dependence of ray trajectory and damping for the ion Bernstein wave in the TNT-A Tokamak." Japanese Journal of Applied Physics, Part 1 (Regular Papers & Short Notes) 26(3): 505-6.
The dependence of ray trajectories and damping on various plasma parameters was studied using three-dimensional ray tracing for an ion Bernstein wave in the TNT-A tokamak. The condition for wave power absorption dominated by electron Landau damping was also estimated.
SUGAYA, R. (1987). "ION HEATING BY NONLINEAR LANDAU DAMPING OF ION BERNSTEIN WAVES." Physics of Fluids 30(6): 1730-1733.
SUN, G. Z. and J. M. FINN (1987). "DOUBLE ADIABATIC RELAXATION AND STABILITY." Physics of Fluids 30(3): 770-778.
TAKASE, Y., J. D. MOODY, et al. (1987). "Study of directly launched ion Bernstein waves in a tokamak." Physical Review Letters 59(11): 1201-4.
Excitation, propagation, and absorption of directly launched ion Bernstein waves were studied with CO{sub 2} laser scattering in the Alcator-C tokamak. Optimum excitation is obtained when an ion-cyclotron harmonic layer is located just behind the antenna. Wave absorption at the omega =3 Omega {sub D}=1.5 Omega {sub H} layer was observed. It may be necessary to control the edge density fluctuation level in order to minimize scattering of the ion Bernstein waves and optimize heating. (18 refs.)
TRIPATHI, V. K., C. S. LIU, et al. (1987). "Kinetic theory of stabilization of the interchange mode by ion Bernstein waves." Nuclear Fusion 27(2): 287-97.
Radiofrequency fields launched in the ion Bernstein mode in a plasma can stabilize the interchange mode by coupling it to ion Bernstein wave side-bands. A kinetic theory for non-linear coupling, based on the guiding centre formalism, is developed and conditions for stabilization are derived. Waves with omega {sub 0}> omega {sub ci}, k{sub 0 perpendicular to }v{sub i}/ omega {sub ci}>or approximately=1 are better suited for stabilization. (10 refs.)
TURNER, L. and S. C. PRAGER (1987). "MAGNETIC TOPOLOGY OF HELICAL OHMIC STEADY-STATES." Physics of Fluids 30(3): 787-796.
UEDA, Y., N. ASAKURA, et al. (1987). "Change of the magnetic field structure during relaxation in the high pinch parameter regime of the reversed field pinch REPUTE-1." Nuclear Fusion 27(9): 1453-60.
Magnetic field measurements in a reversed field pinch plasma have been made with an insertable magnetic probe in REPUTE-1. The relaxation process of the magnetic field configuration in a high theta ( theta is the pinch parameter) regime has been studied. The safety factor on axis, q{sub 0}, decreases as theta increases, and there exists a lower limit to q{sub 0}( approximately 0.10). Close to this limit, q{sub 0} increases when the magnetic field configuration relaxes. These observations agree well with numerical results of non-linear resistive MHD theory. (20 refs.)
WATT, R. G., G. A. WURDEN, et al. (1987). "Pellet injection on the ZT-40M reversed field pinch." Review of Scientific Instruments 58(8): 1401-5.
The four barrel pneumatic pellet injector originally used on the Alcator C tokamak at MIT has been reconfigured for use on the ZT-40M reversed field pinch (RFP) at Los Alamos National Laboratory. The lower temperature and density in this device require operation at velocities significantly lower than on Alcator. Allowing a 2-s warmup prior to firing the injector has enabled operation with pressures as low as 40 psig and velocities ranging as low as 200 m/s with H{sub 2} pellets, without major cryostat modification. Initial injection experiments in the ZT-40M plasma have verified that pellet refueling is effective for density control, but the net particle inventory increase is less than might be expected based on simple models. Indications of nonoptimal penetration and curvature of the pellet trajectory, in addition to comparable time scales for particle confinement and pellet transit, may account for this. (15 refs.)
WILSON, J. R., R. BELL, et al. (1987). "THE EVOLUTION OF PLASMA PARAMETERS AS GOVERNED BY EDGE PHENOMENA DURING ION BERNSTEIN WAVE (IBW) HEATING." Journal of Nuclear Materials 147: 616-620.
WURDEN, G. A., P. G. WEBER, et al. (1987). "Pellet refuelling of the ZT-40M reversed field pinch." Nuclear Fusion 27(5): 857-62.
The first pellet refuelling of a reversed field pinch has been successfully demonstrated in ZT-4OM. By injecting a series of pellets, density increases of 600% have been observed, with no disruptions. Because of a coincidence of pellet transit time, Ohmic reheat time and particle confinement times (about 0.5 ms) a significant fraction of the plasma inventory 'leaks out' before pellet ablation is complete. Strong on-axis peaking of the density profile due to pellet fuelling is not typically observed; this is due to large poloidal and toroidal deflections observed in the pellet trajectory, beginning in the plasma edge. Asymmetric ablation of the pellet by a suprathermal electron population is suspected and simple modelling can account for the observed trajectories. (21 refs.)
ANTANI, S. N. and J. E. SCHARER (1988). "ION-BERNSTEIN WAVE-PROPAGATION IN TOKAMAK EDGE PLASMA DURING HEATING IN THE ION-CYCLOTRON RANGE OF FREQUENCIES." Physics of Fluids 31(10): 3018-3023.
BRODEUR, P., G. RATEL, et al. (1988). "Far-forward and forward CW CO{sub 2} laser scattering measurements of ion-acoustic waves in a quiescent plasma." Plasma Physics and Controlled Fusion 30(7): 805-27.
Far-forward and forward collective light scattering measurements of externally launched ion-acoustic waves in a low density argon plasma column have been performed with a CW CO{sub 2} laser. The scattering processes taking place are described theoretically by the same general formalism. Wave dispersive and damping properties are studied consistently from 2 to 200 cm{sup -1}. Plasma parameters are deduced and compared to electrostatic probe and antenna measurements. It is found that high frequency wave properties are erroneously measured with a grid antenna. (42 refs.)
FERNANDEZ, J. C., C. W. BARNES, et al. (1988). "Energy confinement studies in spheromaks with mesh flux conservers." Nuclear Fusion 28(9): 1555-94.
The authors present experiments and analysis of energy confinement on the CTX spheromak. Compared to previous published results from 0.4 m radius flux conservers, in a 0.67 m radius mesh flux conserver (with the current density kept constant), the magnetic field increases while the plasma density is kept the same. However, the electron temperature does not rise, and hence ( beta ){sub vol} drops. The plasma resistivity remains constant (the resistance drops as the size increases), and the energy confinement time stays the same. Plasma energy content results from spheromaks during sustainment by helicity injection are also presented and show confinement equivalent to that during the decay phase. Increased magnetic field in the same size experiment produces very little improvement in electron temperature and a decrease in confinement time. The resistive decay time is found to be empirically independent of the core electron temperature. It is, instead, proportional to the strength of the magnetic field at constant plasma density, while the ratio of magnetic field to decay time depends on plasma density, consistently with ionization breakdown at the edge of the spheromak dominating helicity dissipation. The possible causes of this observed confinement are examined separately in detailed quantitative and qualitative studies. (65 refs.)
JONES, R. (1988). "Turbulent thermal insulation of toroidal plasmas." Current Science 57(18): 991-5.
Turbulent thermal insulation can substantially improve the performance of a toroidal plasma confinement device, enhancing the confinement time. (14 refs.)
MASAMUNE, S. (1988). "DENSITY-FLUCTUATIONS IN A REVERSED-FIELD PINCH." Physics of Fluids 31(5): 1231-1236.
MILLER, G., J. C. INGRAHAM, et al. (1988). "Improved bolometry system for reversed field pinch research." Review of Scientific Instruments 59(5): 700-8.
A bolometry system for time and spatial resolved energy-loss measurements on the ZT-40M reversed-field pinch is described. This system allows approximately 100- mu s time resolution with a radiation detection limit of approximately 100 mW/cm{sup 2}, in particularly noisy electromagnetic environments. The bolometer element fabrication using evaporative techniques and photolithography is discussed. Increased noise rejection is realized by using an oscillator/FM receiver electronics system based on 5-MHz AC excitation of the resistor. An improved oscillator is described. Because resistive sensing is done at 5 MHz, pulsed heating currents may simultaneously be passed through the bolometer resistor and this allows for absolute in situ system calibration. A new method of such calibration that is independent of the resistor geometry is described. (15 refs.)
MILLER, G. (1988). "MODELING OF REVERSED-FIELD-PINCH MAGNETIC PROBE MEASUREMENTS." Physics of Fluids 31(11): 3330-3337.
MIYAMOTO, K. (1988). "Dynamic processes in relaxation phenomena of RFP plasma." Plasma Physics and Controlled Fusion 30(11): 1493-504.
Research on causal mechanisms and dynamic processes of relaxation phenomena in a Reversed Field Pinch (RFP) plasma is reviewed in relation to toroidal flux generation, anomalous ion heating and loop voltage. The anomalous transport related to MHD fluctuations is briefly mentioned. Recent development of three dimensional, non-linear resistive MHD simulations and detailed measurements in RFP experiments reveal the behavior of MHD activity of RFP plasma more clearly. (36 refs.)
MOODY, J. D., M. PORKOLAB, et al. (1988). "Ion-Bernstein-wave heating and improved confinement in the Alcator C tokamak." Physical Review Letters 60(4): 298-301.
Plasma heating by directly launched ion Bernstein waves is investigated in the Alcator C tokamak at densities n{sub e}>or approximately=1*10{sup 20} m{sup -3}. For injected RF powers P{sub rf}<or approximately=180 kW, bulk-ion-temperature increases of Delta T{sub i}<or approximately=380 eV, as well as improvements in global particle and central impurity confinement times by a factor of 2-3, are observed. In certain regimes, the energy-confinement time improves relative to the initial Ohmic value. At higher densities (n{sub e}>or approximately=2*10{sup 20} m{sup -3}), the heating ceases to be efficient. (15 refs.)
OKAMOTO, M. and M. ONO (1988). "Modelling of ion Bernstein wave heating in JIPPT-II-U." Nuclear Fusion 28(8): 1385-92.
By using a transport code combined with an ion Bernstein wave tokamak ray tracing code, a modelling code for the ion Bernstein wave heating has been developed. With this code, the ion Bernstein wave heating experiment on the JIPPT-II-U tokamak has been analysed. It is assumed that a resonance layer is formed by the third harmonic of deuterium-like ions, such as fully ionized carbon, and oxygen ions near the plasma centre. As wave absorption mechanisms, electron Landau damping, ion cyclotron harmonic damping, and collisional damping are considered. The characteristics of the ion Bernstein wave heating experiment, such as an increase in ion temerature, a strong dependence of the quality factor on the magnetic field strength and a dependence of the ion temperature increment on the input power are well reproduced. (19 refs.)
ONO, Y., R. A. ELLIS, et al. (1988). "Relaxation phenomena in the high-temperature S-1 spheromak." Physical Review Letters 61(25): 2847-50.
Operation of the S-1 device in a high-current-density (j/n{sub e}>or=2*10{sup -14} A m) regime has created high-electron-temperature spheromaks (50 eV<or=T{sub e}<or=130 eV). The mechanisms and causes of the periodic relaxation events often observed in these hotter plasmas were made clear. A relationship between the MHD relaxation cycle and confinement was revealed for the first time. Resistive loss at the outer edge causes a departure from the initial force-free minimum-energy Taylor state to an MHD profile unstable to low-n ideal MHD modes; a relaxation event then returns the configuration to a Taylor state. (12 refs.)
ONO, M., P. BEIERSDORFER, et al. (1988). "Effects of high-power ion Bernstein waves on a tokamak plasma." Physical Review Letters 60(4): 294-7.
Effects of high-power ion-Bernstein-wave heating on the PLT tokamak plasmas have been investigated with up to 650 kW of RF power coupled to the plasma. A significant improvement in the deuterium particle confinement (as much as 300%) during high-power ion-Bernstein-wave heating has been observed. Associated with it, a longer injected-impurity-confinement time, reduced drift-wave turbulence activity, frequency shifts of drift-wave turbulence, and development of a large negative edge potential were observed. The ion heating efficiency, Delta T{sub i}(0)n{sub e}/P{sub rf}=6*10{sup 13} eV cm{sup -3}/kW, without high-energy tail ions, is relatively constant up to the maximum rf power. (22 refs.)
PHILLIPS, J. A., D. A. BAKER, et al. (1988). "Startup of reversed field pinches and current ramping using dynamo action." Nuclear Fusion 28(7): 1241-54.
The different startup modes of a reversed-field pinch (RFP) are examined and compared. The RFP startup is not the same as startup of other toroidal devices such as the tokamak because of the spatial and temporal variations of the toroidal field and the reversed toroidal field in the outer region of the pinch near the wall. In matched mode startup, used in many RFP experiments today, the toroidal flux is held constant during the current rise, with the field reversal occurring before the peak current. This mode, with its short rise time, has a low volt-second (V.s) input but requires a high toroidal voltage to reach a specific current in a relatively short time. In a ramped current mode, a low current RFP discharge is ramped to its final peak value. The toroidal flux needed as the current rises is generated by a dynamo action. This slower startup mode can be driven by a lower voltage but requires more V.s input than the matched mode. Different startup modes in the ZT-40M experiment at Los Alamos are compared and an analytic expression is given for characterizing the V.s contributions. The resistive component of the toroidal loop voltage during the current rise in ramped discharges is found to depend on theta ( Theta =B{sub pol}(wall)/B{sub phi ave}). At a theta of equivalent to 1.45, the resistive voltage has a minimum and it has been possible to reduce the V.s input by as much as 40% in ramp discharges by keeping theta close to this value. (40 refs.)
SAHA, S. K. and S. N. SENGUPTA (1988). "Selective ion heating by energetic ion beam injection into a plasma." Physics Letters A 129(1): 33-7.
High energy neutral beam injected into a plasma has been predicted to be susceptible to microinstabilities. If the growth time of these microinstabilities is shorter than the collisional relaxation time, then the energy transfer from the beam to the plasma may take place, in the initial phase, by collisionless wave-particle interaction. The authors report here an experiment where this effect has indeed been observed. Here ion Bernstein waves excited by an energetic ion beam lead to selective ion heating of the plasma in a collisionless manner. (14 refs.)
SCARDOVELLI, R. A. and G. H. MILEY (1988). "CURRENT REQUIREMENTS FOR OHMIC IGNITION IN OHMICALLY HEATED TOROIDAL EXPERIMENT AND REVERSED-FIELD PINCH REACTORS." Fusion Technology 13(3): 510-514.
SHARMA, A. and V. K. TRIPATHI (1988). "EXCITATION OF KINETIC ALFVEN WAVES IN THE ION-BERNSTEIN WAVE HEATING OF A PLASMA." Physics of Fluids 31(12): 3697-3702.
SHINOHARA, S., O. NAITO, et al. (1988). "EXCITATION OF THE 2ND HARMONICS DUE TO A RESONANT MODE-MODE COUPLING FOR THE ION BERNSTEIN WAVE IN TOKAMAKS." Journal of the Physical Society of Japan 57(3): 707-710.
SHOKAIR, I. R., R. W. CONN, et al. (1988). "THEORY AND MODELING OF 2ND HARMONIC ICRF COUPLING AND HEATING IN INHOMOGENEOUS PLASMAS." Nuclear Fusion 28(8): 1393-1412.
SKIFF, F., M. ONO, et al. (1988). "EXCITATION OF ION BERNSTEIN WAVES FROM LOOP ANTENNAS." Physics of Fluids 31(7): 2030-2036.
SUGAWA, M. (1988). "Observation of self-interaction of Bernstein waves by nonlinear Landau damping." Physical Review Letters 61(5): 543-6.
It is confirmed experimentally that the self-interaction of electron Bernstein waves occurs by means of nonlinear cyclotron (Landau) damping. This wave-particle interaction occurs in a relatively broad frequency range: 1.46< omega / omega {sub c}<1.54, 1.60< omega / omega {sub c}<1.75. The virtual wave is observed most strongly at the half-harmonic electron cyclotron frequency near omega / omega {sub c} equivalent to 1.5. Although this self-interaction always occurs for k{sub }// nu {sub t}/ omega {sub c}<or approximately=0.1, the virtual wave cannot be detected for k{sub }// nu {sub t}/ omega {sub c}>or approximately=0.15. (12 refs.)
TRIPATHI, Y. K. and R. P. SHARMA (1988). "SOME PARAMETRIC-INSTABILITIES OF AN ORDINARY ELECTROMAGNETIC-WAVE IN MAGNETIZED PLASMAS." Physical Review A 38(6): 2991-2995.
WENZEL, K. W. and R. D. PETRASSO (1988). "X-ray response of silicon surface-barrier diodes at 8 and 17.5 keV: evidence that the X-ray sensitive depth is not generally the depletion depth." Review of Scientific Instruments 59(8): 1380-1387.
The absolute X-ray response of 18 EG&G Ortec partially depleted silicon surface-barrier diodes (SBDs) has been measured at 8 and 17.5 keV. In addition the X-ray response of four Tennelec and two United Detector Technology partially depleted SBDs has been examined. The variation in response to 8 keV X-rays, for which the optical depth is about 5, is comparatively slight (<or approximately=12%). The variation in response to 17.5-keV X-rays, for which the optical depth is only approximately 0.7, is comparatively large ( approximately 100%). These variations are mainly attributable to differences in the SBD physical thicknesses, and thus to differences in optical depth. At both 8 and 17.5 keV the diodes respond linearly to large variations in incident flux (over three orders of magnitude). This linearity, and the diode X-ray response, in general, is insensitive to large changes in the detector bias voltage; thus the depletion depth, proportional to the square root of the bias voltage, does not play a critical role in determining the X-ray sensitive depth. It is important to emphasize that this finding is contrary to the commonly held belief that the X-ray sensitive depth is equivalent to the depletion depth. In addition, this result has a direct bearing on both SBD and p-i-n detectors intended for fully depleted operation, but used in an underbiased mode such that they are actually partially depleted. It is concluded that SBDs have attractive features for quantitatively measuring X radiation from high-intensity sources for which hv>or approximately=10 keV. (60 refs.)
ANTONI, V., P. MARTIN, et al. (1989). "Mean magnetic field profiles and their dynamics in reversed field pinches." Nuclear Fusion 29(10): 1759-68.
Magnetic field profiles have been measured in the ETA-BETA II experiment by means of insertable probes. The main properties of the mean magnetic field profiles and the mu -profiles at various values of the pinch parameter theta are discussed as well as experimental evidence of the existence of constraints on the radial distribution of mu . The dynamics of the magnetic distributions is investigated and given in terms of the cyclic current redistribution associated with the sustainment of the configuration. (34 refs.)
ASAKURA, N., Y. NAGAYAMA, et al. (1989). "Soft X-ray measurements on the repute-1 reversed field pinch." Nuclear Fusion 29(6): 893-904.
Soft X-ray measurements have been made with surface barrier diode (SBD) arrays in the REPUTE-1 RFP device (R/a=82/22 cm). Arrays of seven and eleven SBDs are arranged so that they view a minor cross-section of the plasma from the vertical and horizontal diagnostic ports, respectively. Applying the tomography technique to the soft X-ray data, a two-dimensional emissivity profile can be reconstructed. In the sustainment phase of the RFP discharge, the soft X-ray emissivity profile exhibits a periodic peaking and a subsequent large crash across the minor cross-section. The fluctuation period is in the range of 60-150 mu s, whereas the duration of the large crash is always about 30 mu s (corresponding to approximately 100* tau {sub A}, where tau {sub A} is the Alfven transit time). The soft X-ray emission is due to two modes: one is a dominantly m=0 mode, which corresponds to the axisymmetric increase or decrease of the soft X-ray emissivity profile, and the other is an m=1 mode. (31 refs.)
BATHKE, C. G., R. A. KRAKOWSKI, et al. (1989). "TECHNOLOGY AND PHYSICS IMPLICATIONS OF OSCILLATING-FIELD CURRENT DRIVE IN REVERSED-FIELD PINCHES." Fusion Technology 15(2): 1082-1087.
BELLAN, P. M. (1989). "Non-observation of AC helicity injection." Nuclear Fusion 29(1): 78-80.
Despite a careful measurement technique, an experiment to observe AC helicity injection in the Caltech Encore tokamak was not successful. A criterion used here for discriminating AC helicity injection from spurious effects should be useful for judging the data of any future attempts to observe AC helicity injection in other devices. (8 refs.)
CARTER, M. D., E. F. JAEGER, et al. (1989). "NON-LINEAR CORE PLASMA RESPONSE TO ICRF HEATING WITH TRANSPORT." Nuclear Fusion 29(12): 2141-2154.
ENGLAND, A. C., O. C. ELDRIDGE, et al. (1989). "Power transmission and coupling for radiofrequency heating of plasmas." Nuclear Fusion 29(9): 1527-33.
RF power is widely used as an auxiliary heating method in fusion devices. The authors review the relevant theoretical considerations for the ion cyclotron, lower hybrid and electron cyclotron ranges of frequency, and present the history, the state of the art, and the plans and prospects for antennas and transmission lines for RF heating. Reactor-relevant concerns are discussed, and the information needed to develop realistic antenna designs for a reactor environment is assessed. (455 refs.)
HATTORI, K., Y. UEDA, et al. (1989). "Asymmetric perturbations of toroidal flux in ramped-up discharges on REPUTE-1 reversed field pinch." Journal of the Physical Society of Japan 58(1): 24-7.
Toroidally asymmetric perturbations of toroidal flux in a reversed field pinch (RFP) plasma have been observed with a set of 18 flux loops in ramping current discharges on the REPUTE-1 RFP. The toroidal flux perturbation is induced toroidally localized and spreads as it propagates in the electron-diamagnetic drift direction. The dominant toroidal mode number of the flux perturbation is usually 1, but becomes 2 when the toroidal flux is strongly generated, and the second peak merges with the first one as it propagates in the toroidal direction. Thus, a locally generated toroidal flux becomes toroidally symmetric. (16 refs.)
HAYDEN, R. J. and B. ALPER (1989). "Coherent oscillations in the HBTX1B Reversed Field Pinch." Plasma Physics and Controlled Fusion 31(2): 193-204.
Large amplitude low frequency (0.7-2.5 kHz) oscillations which are in phase across the minor radius have been observed in soft X-ray emission since the removal of limiters from HBTX1B Reversed Field Pinch for plasmas with small field reversal, F, at all toroidal currents. These oscillations, due to variations in electron temperature, are in antiphase with oscillations in plasma resistance and ion temperature and occur in the majority of discharges where the resistance is closest to its lowest value. Correlated magnetic oscillations indicate cyclic behaviour in the F- theta plane similar to that observed at high values of theta on other experiments, where theta is the pinch parameter. (15 refs.)
HORIUCHI, R., K. KUSANO, et al. (1989). "Simulation study of self-organization and self-relaxation processes in plasma." Progress of Theoretical Physics Supplement(99): 206-19.
Recent progress in super-computers has opened a new tool for the study of nonlinear dynamics in plasmas. The authors review the recent computer simulation studies for the self-organization and self-relaxation processes in MHD plasmas carried out by the Hiroshima group. Taylor's conjecture for the selective dissipation of magnetic energy keeping the total magnetic helicity constant is confirmed and the mechanisms for formation and maintenance of the reversed field pinch configuration are explained in terms of the self-relaxation of the plasma du to a driven magnetic reconnection triggered by ideal MHD helical kink unstable modes and their nonlinear couplings. (22 refs.)
LEE, G. S., P. H. DIAMOND, et al. (1989). "Dynamics of magnetic relaxation in a high-temperature, current-carrying plasma." Physics of Fluids B (Plasma Physics) 1(1): 99-108.
The results of a study of the role of multiple-helicity nonlinear interaction of tearing modes and the dynamics of magnetic relaxation and dynamo activity in a high-temperature, current-carrying plasma are described. A set of fluid equations for tearing modes in the semicollisional regime is derived, and a previous resistive magnetohydrodynamic study of tearing mode turbulence is thus extended to high-temperature regimes. Because of the direct connection between fluctuation evolution and configuration evolution, a generalized nonlinear theory of the turbulent dynamo and magnetic relaxation is proposed, and a two-point (( Del {sub perpendicular to }p) psi )- correlation evolution is determined by calculating the relaxation time tau {sub cl}. This calculated relaxation time is shown to serve as the phase shift between nu {sub r} and psi and hence to control magnetic energy relaxation and dynamo processes. Careful study of the two separated regions of kink-tearing modes (i.e., the resonant region and the exterior region) reveals the direct relationship between the equilibrium magnetic energy relaxation and average magnetic flux evolution. Thus a theoretical interpretation of the observed correlation between magnetic fluctuations and maintenance of field reversal is provided. Finally, the correlation lifetime of the flux transport term is evaluated from the renormalized two-point theory, and the saturated magnetic fluctuation level is estimated. The result shows that the saturation level is independent of or very weakly dependent on the Lundquist number S. The implications of this result for anomalous thermal transport and dynamo activity in high-temperature reversed field pinch experiments are discussed. (19 refs.)
MADARAME, H., T. TERAI, et al. (1989). "A conceptual design of RFP fusion power core-REPUTER-I." Fusion Engineering and Design 7(3/4): 389-416.
REPUTER-I is a conceptual design study on a compact reversed field pinch fusion reactor with a solid breeder blanket. The first wall is structurally independent from the blanket. The blanket is composed of SiC reinforced blocks which form a stable arch, and the stresses in SiC are basically compressive. Pressurized water is chosen for the first wall and as a limiter coolant in order to remove a high heat flux, while helium gas is used for the blanket coolant. A pile of thin Li{sub 2}O tiles and a thick beryllium tile provides a satisfactory tritium breeding ratio. A pumped limiter is shown to meet the demand for heat and particle exhaust removal in a compact RFP core. Magnet system design and plasma performance analysis are made as well on the first wall, blanket and limiter design. (56 refs.)
MILLER, G. (1989). "Nonlinear tearing modes in the reversed field pinch." Physics of Fluids B (Plasma Physics) 1(2): 384-91.
Finite-amplitude islands, which are the saturated states of tearing modes in the reversed field pinch, are calculated. These states are bifurcated noncylindrical equilibrium states. With sigma (r) ( sigma identical to j.B/B{sup 2}) nonuniform across the plasma, as is consistent with experiment, a variety of m=1 and m=0 bifurcated equilibria are possible, instead of just the m=1 helix calculated for uniform sigma (r) by Taylor (in Pulsed High Beta Plasmas, edited by D. Evans (Pergamon, Oxford, 1976), p.59). Assuming the magnetic field lines in the reversed field pinch are weakly stochastic, the growth time of an unstable tearing mode is on the inertial time scale, as in the Taylor model, in contrast to growth on the resistive time scale predicted from nonlinear tearing mode theory when magnetic surfaces exist. The dependence of the saturated island width on radius of a conducting shell is investigated. Islands in the reversed field pinch often have magnetic wells in the island interior, which may result in improved confinement in the island regions. (20 refs.)
MIYAMOTO, K. (1989). "MHD fluctuation level and transport in reversed field pinch." Journal of the Physical Society of Japan 58(5): 1501-3.
MHD fluctuation level in RFP plasma is deduced from a model that supposes that MHD relaxation and sustainment of reversed field pinch (RFP) configuration are caused by the reconnection of lines of magnetic field driven by (m=0,n=1) mode nonlinearly excited by m=1 (poloidal mode number) internal kink modes with the different toroidal mode numbers n and n+1. Thus the lines of magnetic forces become stochastic by the MHD fluctuations and the transport of RFP plasma in the braided magnetic field is clarified. (19 refs.)
PETRILLO, V., C. MAROLI, et al. (1989). "ION-CYCLOTRON ELECTROSTATIC-WAVES IN A LOW-TEMPERATURE LOW-DENSITY MAGNETIZED PLASMA." Nuovo Cimento Della Societa Italiana Di Fisica D-Condensed Matter Atomic Molecular and Chemical Physics Fluids Plasmas Biophysics 11(9): 1337-1348.
SHIMOTOHNO, H. and S. KONDO (1989). "A parametric system study for the selection of RFP fusion reactor design point." Fusion Engineering and Design 7(3/4): 377-88.
Based on simple physics, engineering, and economic models among various design parameters of reversed-field pinch (RFP) fusion reactors, a parametric system study program has been developed giving cost of electricity (COE) for a given set of design parameters or a set of design parameters giving the minimum COE. This program has been used extensively to select design points and evaluate the sensitivity of physics, engineering, economic and operational parameters in the REPUTER RFP fusion reactor design study effort. Several key features emerged from the study in relation to high power density RFP fusion reactors with solid breeder blanket. They are: (1) selection of normal conductor for TF and PF coils to reduce the thickness of radiation shielding, (2) selection of a total first wall loading of 18 MW/m{sup 2} considering the potential economic merit and nonlinear effects which might shorten the lifetime of the first wall, and (3) a maintenance policy of package replacement of the fusion power core, including TF coils, to significantly reduce maintenance time. The uncertainty range of COE was evaluated based on the sensitivity and uncertainty of design variables and parameters. The result warn one of the use of COE evaluated in this program for the purpose of comparison of economy with the result of other fusion system studies. (16 refs.)
TRIPATHI, Y. K. and R. P. SHARMA (1989). "Stimulated Brillouin scattering of obliquely propagating extraordinary electromagnetic waves off ion-Bernstein waves in plasmas." IEEE Transactions on Plasma Science 17(4): 569-75.
The parametric decay of an obliquely propagating extraordinary electromagnetic wave into another extraordinary electromagnetic wave and a low-frequency electrostatic ion-Berstein wave via the stimulated Brillouin scattering process is considered. Explicit expressions for the growth rate and threshold power for this instability are presented. Applications of the present investigation to fusion plasmas in the end cells of tandem mirrors and the magnetospheric and ionospheric plasmas to explain wave phenomena are pointed out. The dependence of threshold power on the pump wave angle is studied. The general expressions for the growth rate and threshold derived here can also be useful to explain future experimental data. (17 refs.)
YOSHIZAWA, A. (1989). "Compatibility of the turbulent dynamic concept with reversed field pinches." Physics of Fluids B (Plasma Physics) 1(5): 983-5.
The sustainment phase of reversed field pinches (RFP's) is investigated with the concept of the turbulent dynamo. In the case of a finite toroidal voltage, the familiar alpha-beta dynamo, which is equivalent to Taylor's and modified Bessel function models, cannot satisfy the condition of vanishing currents at the plasma edge. The importance of another alpha term is stressed, and the deep reversal resulting from a large toroidal voltage, the enhanced dynamo effect in ramped discharges, and the F-(-) pumping are discussed in light of the characteristics of the edge region that have a great influence on the global structure of the RFP. (21 refs.)
ANTONI, V., M. BAGATIN, et al. (1990). "Edge plasma investigation on the reversed field pinch ETA BETA II." Journal of Nuclear Materials 176/177: 1076-82.
The boundary region of the ETA BETA II RFP experiment has been investigated by movable instrumented limiters equipped with Langmuir probes, heat flux probes, and surface collectors. Edge plasma temperature and density are of the order of 10 eV and 0.5-1*10{sup 19} m{sup -3}. The heat deposition on the movable limiter is asymmetric between the electron and the ion drift side, and a power flux of the order of 300 MW/m{sup 2} is measured on the electron drift side. This high value has been explained by the presence of current carrying electrons coming from the hotter region of the plasma, through a stochastic region extending into the plasma. Surface collectors have been inserted in the plasma and analyzed by RBS and ERD techniques. The implanted impurities are mainly oxygen and metallic elements of the liner, both showing a quite linear build up as a function of the exposure time whereas the amount of trapped deuterium tends to a stationary value after a few discharges. The deuterium flux incident on the wall has been inferred, the impurity yield estimated, and the generation mechanisms discussed. The deposition radial profiles give information on the screening capability of the boundary region. (24 refs.)
BODIN, H. A. B. (1990). "The reversed field pinch." Nuclear Fusion 30(9): 1717-37.
Summarizes the theoretical basis for the RFP and reviews the status of research in this field. The RFP is a relaxed state system well described by Taylor's theory which explains many observations. The RFP is of interest because its study will increase the understanding of toroidal confinement in general, which might lead to better reactor designs, and the RFP itself has potential as a reactor, for example the improved, high energy density, compact RFP reactor. In the last five or ten years, RFP research has expanded, with some 15-20 machines operating or under construction, and the plasma parameters have improved substantially, with a confinement time of 0.5-1 ms and temperatures approaching 1 keV; values of beta {sub theta } approximately 5-15% are reached routinely. Following an overview of recent results, three key problems are discussed in more detail: (1) resistivity, edge physics and ion heating (the ions are heated by fluctuations which drive the RFP dynamo); (2) operation of the RFP with an (electrically) thin shell which permits the growth of new unstable modes which degrade the confinement; and (3) scaling. Over the current range of 0.5 MA, favourable scaling trends of temperature and confinement with current are identified, but experiments at much higher currents on the two megaampere machines-RFX at Padua and CPRF at Los Alamos-due to operate in the early 1990s, are needed. A brief account of the compact RFP reactor is given, followed by a summary with an indication of future trends. (115 refs.)
BOODY, F. P. and M. A. PRELAS (1990). "Fast multiline low-Z impurity imaging diagnostic for fluctuation and particle confinement studies." Review of Scientific Instruments
8TH TOPICAL CONF ON HIGH TEMPERATURE PLASMA DIAGNOSTICS 61(10): 3064-3066.
A system for imaging complete Tokamak edge cross sections using radiation from intrinsic low-Z impurities, such as carbon, with a frequency response of 500 kHz and a spatial resolution of 1-4 cm (depending on the thickness of the radiating region) is being developed. The noise level is determined by the electronics and is <or=2% for the full 500 kHz bandwidth and <or=1% for a 100 kHz subset of that bandwidth. Photon noise is 0.3% for the full 500 kHz bandwidth. This diagnostic can be used for studying fluctuations in P{sub rad}( lambda ), Gamma {sub z}, and T{sub e}. Additionally it may be useful for studying marfes and detached plasmas, transport and oscillations during pellet injection, and as a fast disruption precursor monitor. (12 refs.)
BOOZER, A. H., D. E. BALDWIN, et al. (1990). "Alternate transport." Physics of Fluids B (Plasma Physics) 2(12): 2870-8.
The understanding of tokamak transport depends on the exploration of a wide range of theoretical models and of a variety of toroidal experiments. This report considers the contributions that nontokamak, but toroidal, experiments can make to the understanding of tokamak transport as well as theoretical alternatives to the standard drift wave model of tokamak transport. (48 refs.)
DARROW, D. S., M. ONO, et al. (1990). "Properties of DC helicity injected tokamak plasmas." Physics of Fluids B (Plasma Physics) 2(6): 1415-1420.
Several DC helicity injection experiments using an electron beam technique have been conducted on the current drive experiment (CDX) (Phys. Rev. Lett. 59, 2165 (1987)) and the continuous current tokamak (CCT) (Phys. Rev Lett. 63, 2365 (1989)). The data strongly suggest that tokamak plasmas are being formed and maintained by this method. The largest currents driven to date are 1 kA in CDX (q{sub a}=5) and 6 kA in CCT (q{sub a}=3.5). An initial comparison of discharge properties with helicity theory indicates rough agreement. Current drive energy efficiencies are 9% and 23% of Ohmic efficiency in two cases analyzed. Strong radial electric fields are observed in these plasmas that cause poloidal rotation and, possibly, improved confinement. (15 refs.)
FERNANDEZ, J. C., T. R. JARBOE, et al. (1990). "Ion heating and current drive from relaxation in decaying spheromaks in mesh flux conservers." Nuclear Fusion 30(1): 67-80.
The plasma energy confinement in CTX mesh flux conservers appears to be dominated by relaxation fluctuations which drive current in the resistive plasma edge region. Strong ion heating is expected from these relaxation processes. The paper presents new data, from previously studied decaying discharges, which show impurity oxygen ion Doppler temperatures much above the Thomson scattering electron temperature. In addition, using the standard MHD equilibrium model by Knox et al. (1986), and the resistivity model previously described by Fernandez et al. (1988), the contribution of relaxation fluctuations to spheromak current drive along the open magnetic field lines at the spheromak edge can be estimated. (26 refs.)
GIUBBILEI, M., P. MARTIN, et al. (1990). "A mechanism for plasma heating in driven relaxing magnetic field configurations." Plasma Physics and Controlled Fusion 32(5): 405-11.
Magnetic configurations used for laboratory plasma confinement and those naturally occurring in the solar corona are both driven systems: in the former, the energy input is generally supplied by externally applied voltages; in the latter, by boundary fluid motions in the photosphere. Moreover, solar corona dynamics has been described in a very similar way to Reversed Field Pinch dynamics. Using these analogies as a starting point, a description is presented, supported by experimental results, of plasma heating in driven partially relaxed RFP configurations resulting from energy dissipation during reconnection processes. The power deposited in the plasma through this mechanism adds to the Joule heating of the mean field dissipation and may also provide a direct channel for ion heating through viscous dissipation. (34 refs.)
GUREVICH, A. V., K. P. ZYBIN, et al. (1990). "THE TURBULENT-PLASMA IN RELAXED STATE AND THE PROCESSES OF THE ANOMALOUS TRANSFER." Zhurnal Eksperimentalnoi I Teoreticheskoi Fiziki 98(2): 468-484.
HEIKKINEN, J. A. and M. BURES (1990). "Fast wave antenna coupling to slow waves and ion Bernstein waves during ICRF heating of tokamak plasmas." Plasma Physics and Controlled Fusion 32(3): 173-88.
The coupling of the fast wave antenna to short wavelength waves is analysed by taking into account the imperfect alignment of the antenna screens along the magnetic field lines of a Tokamak plasma. Analytical and numerical estimates for the fraction of the power going into the short wavelength modes are given for various edge plasma parameters and their profiles. The functional dependence of this fraction on the screen angle and the excited spectrum of the short wavelength waves are computed. According to the results a nonnegligible amount of power could be radiated from the present ICRF antennas in the form of ion Bernstein waves or slow waves for realistic edge density profiles. For edge plasmas extending to the screen, the excited electrostatic wave spectrum is peaked for large parallel refractive indices n{sub z} and the electrostatic wave coupling is due to the parallel electric field along the magnetic field lines which therefore requires some misalignment of the screens. When the density near the screens is sufficiently low to allow the presence of the slow mode resonance, the excited electrostatic wave spectrum is peaked for mod n{sub z} mod <1 and the coupling is mainly due to the confluence between the fast wave and slow wave and does not necessarily require any misalignment. Possible connections of these phenomena to the deleterious edge modification and enhanced impurity influx from the antennae, observed in the heating experiments of JET, are discussed. (24 refs.)
INGRAHAM, J. C., R. F. ELLIS, et al. (1990). "Energetic electron measurements in the edge of a reversed-field pinch." Physics of Fluids B (Plasma Physics) 2(1): 143-59.
The edge plasma of the ZT-40M (Fusion Technol.8, 1571 (1985)) reversed-field pinch has been studied using a combination of three different plasma probes: a double-swept Langmuir probe, an electrostatic energy analyzer, and a calorimeter-Langmuir probe. The edge plasma has been measured both with and without a movable graphite tile limiter present nearby in the plasma. Without a limiter a fast nonthermal tail of electrons (T approximately=350 eV) is detected in the edge plasma with nearly unidirectional flow along B and having a density between 2% and 10% of the cold edge plasma (T approximately=20 eV). The toroidal sense of this fast electron flow is against the force of the applied electric field. A large power flux along B is measured flowing in the same direction as the fast electrons and is apparently carried by the fast electrons. With the limiter present the fast electrons are still detected in the plasma, but are strongly attenuated in the shadow of the limiter. The measured scrape-off lengths for both the fast electrons and the cold plasma indicate cross-field transport at the rate of, or less than, Bohm diffusion. Estimates indicate that the fast electrons could carry the reversed-field pinch current density at the edge and, from the measured transverse diffusion rates, could also account for the electron energy loss channel in ZT-40 M. the long mean-free-path kinetic nature of these fast electrons suggests that a kinetic process, rather than a magnetohydrodynamic process that is based upon a local Ohm's law formulation, is responsible for their generation. (50 refs.)
JARBOE, T. R., F. J. WYSOCKI, et al. (1990). "Progress with energy confinement time in the CTX spheromak." Physics of Fluids B (Plasma Physics) 2(6): 1342-1346.
Large improvements in spheromak parameters and new understanding have been obtained from the CTX experiment at Los Alamos (Phys. Rev. Lett. 51, 39 (1983); 61, 2457 (1988)). In one experiment the global energy confinement time has been increased an order of magnitude over previous experiments to 0.2 msec and the magnetic-energy decay time increased to 2 msec. These results were achieved in a decaying spheromak by reducing the helicity dissipation in the edge. In another smaller spheromak, record electron temperatures ( approximately 400 eV) and record magnetic field strengths ( approximately 30 kG) have been obtained. (10 refs.)
LEVINTON, F. M., D. D. MEYERHOFER, et al. (1990). "CONFINEMENT AND POWER BALANCE IN THE S-1 SPHEROMAK." Nuclear Fusion 30(5): 871-879.
MATHESON, P. L., R. A. NEBEL, et al. (1990). "MODELING OF A TOROIDAL FIELD DIVERTOR IN A REVERSED-FIELD PINCH." Fusion Technology 18(2): 257-272.
MURPHY, A. B. (1990). "Waves in the edge plasma during ion cyclotron resonance heating." Fusion Engineering and Design
MEETING OF THE TECHNICAL COMMITTEE OF INTERNATIONAL ATOMIC ENERGY AGENCY ON ION CYCLOTRON RESOURCE HEATING / EDGE PHYSICS 12(1/2): 79-92.
The waves that can propagate in the edge plasma during fast wave ion cyclotron resonance heating (ICRH) for densities sufficiently high that omega < omega {sub pi} are considered; in particular surface wave modes of the compressional Alfven wave, the torsional Alfven wave and the ion Bernstein wave. The basic properties of each wave type are outlined, and previous theoretical and experimental studies briefly reviewed. Evidence is presented that excitation of waves in the edge plasma by the antenna is responsible for the edge heating, and possibly indirectly the increase in impurities, generally observed during ICRH. (63 refs.)
PRAGER, S. C. (1990). "Transport and fluctuations in reversed field pinches." Plasma Physics and Controlled Fusion 32(11): 903-16.
In the reversed field pinch various turbulent transport mechanisms might be simultaneously active, including MHD flow effects, free-streaming in a stochastic magnetic field, direct transfer of fluctuating energy to ions, and electrostatic fluctuation driven transport. By comparing fluctuations and transport in the RFP to that in the related toroidal configurations of the tokamak and stellarator, a greater understanding of toroidal confinement in general might be achieved. In addition to particle and energy transport, current transport sustains the reversed field magnetic configuration and constitutes the dynamo effect. Recent results are available from RFP experiments on electrostatic fluctuations, anomalous ion heating, the presence of free-streaming fast electrons, and the extension of these effects to plasmas of large size. The reactor promise of the RFP is based upon the conjecture that the beta value will remain high and the resistance scaling classical, as has been the case in past experiments. One might expect beta to be limited by resistive interchange turbulence. The MST, RFX, and ZTH experiments will form a sequence of devices to test this scaling conjecture to large size and large current values. (55 refs.)
PRAGER, S. C., A. F. ALMAGRI, et al. (1990). "1ST RESULTS FROM THE MADISON SYMMETRIC TORUS REVERSED FIELD PINCH." Physics of Fluids B-Plasma Physics 2(6): 1367-1371.
SCHAFFER, M. J. (1990). "Magnetohydrodynamic marginal stability study of force-free reversed field pinch configurations with a surrounding vacuum region." Nuclear Fusion 30(4): 635-44.
The ideal and resistive MHD free-boundary stability of m=0, 1, 2 and 3 modes is calculated for the circular force-free reversed field pinch (RFP). Functions mu (r)= mu {sub 0}J.B/B{sup 2} specifying the equilibrium configurations are chosen to approximate the deviations of experimental RFPs from the fully relaxed state. Families of unstable modes are identified. All are resistive in the usual RFP parameter range, unless the distance to the conducting shell is very large. The usually stable m=2 and 3 modes, as well as m=0 and m=1, are unstable near the plasma edge for practical shell separations, and these modes might promote greater edge relaxation than if plasma and shell were in contact. When a small central peaking of mu (r) is present, the unstable mode spectrum hardly varies as the shell is moved outward. It is suggested that extreme closeness of the shell to the plasma might not be required. Moderate plasma-shell separation might also eliminate sawtoothing at high reversals. Supporting evidence from OHTE and HBTX 1C experiments is cited. (30 refs.)
SHINOHARA, S., K. F. SCHOENBERG, et al. (1990). "REP plasma performance in the F- Theta pumping experiments on ZT-40M." Journal of the Physical Society of Japan 59(4): 1232-41.
Plasma performance in the F- Theta pumping condition has been studied on the reversed field pinch (RFP) device ZT-40M. Compared with an un-modulated case, slight changes of plasma parameters are found for the case of current drive phasing. On the other hand, the increases in radiated power and impurity intensity, and the decreases in plasma current and soft X-ray intensity are observed for the anti-drive phasing. Changes of effective minor radius, inferred internal structures and phases between various plasma parameters during modulation are also studied. Frequency spectra up to the 10th harmonic of the deriving frequency (1.3 and 2 kHz) are investigated. The fast penetration (>10 kHz) of changes of plasma parameters to the central region and the increased impurity and soft X-ray intensities with an increase in the frequency are found.
STIX, T. H. (1990). "Waves in plasmas: highlights from the past and present." Physics of Fluids B (Plasma Physics) 2(8): 1729-43.
To illustrate the development of some fundamental concepts in plasma waves, a number of experimental observations, going back over half a century, are reviewed. Particular attention is paid to the phenomena of dispersion, collision-free damping, ray trajectories, amplitude transport, plasma wave echoes, finite-Larmor-radius and cyclotron and cyclotron-harmonic effects, nonlocal response, and mode conversion; to the straight-trajectory approximation and two-level phase mixing; and to quasilinear diffusion and its relation to radio-frequency heating, current drive, and induced neoclassical transport, and to stochasticity and superadiabaticity. Not only is the constructive interplay between experiment and theory noted, but also that major advances have come from each of the many disciplines that invoke plasma physics as a tool, including radio communication, astrophysics, controlled fusion, space physics, and basic research. (49 refs.)
TERRY, P. W. and P. H. DIAMOND (1990). "A self-consistent theory of radial transport of field-aligned current by microturbulence." Physics of Fluids B (Plasma Physics) 2(6): 1128-1137.
The radial transport of field-aligned current resulting from collisionless microturbulence is examined self-consistently. The self-consistent treatment of mode coupling is shown to constrain the transport in such a way that the relaxation of current gradients is regulated solely by electrostatic fluctuations that couple to ion dissipation, even in the presence of temperature gradients and temperature anisotropy. As a consequence, the radial flux of parallel current induced by collisionless microinstabilities is insufficient to account for the dynamo in reversed field pinch plasmas. (27 refs.)
WURDEN, G. A., K. BUCHL, et al. (1990). "Pellet imaging techniques in the ASDEX Tokamak." Review of Scientific Instruments 61(11): 3604-8.
As part of a USDOE/ASDEX collaboration, a detailed examination of pellet ablation in ASDEX with a variety of diagnostics has allowed a better understanding of a number of features of hydrogen ice pellet ablation in a plasma. In particular, fast-gated photos with an intensified Xybion CCD video camera allow in situ velocity measurements of the pellet as it penetrates the plasma. With time resolution of typically 100 ns and exposures every 50 mu s, the evolution of each pellet in a multipellet ASDEX Tokamak plasma discharge can be followed. When the pellet cloud track has striations, the light intensity profile through the cloud is hollow (dark near the pellet), whereas at the beginning or near the end of the pellet trajectory the track is typically smooth (without striations) and has a gaussian-peaked light emission profile. New, single pellet Stark broadened D{sub alpha }, D{sub beta }, and D{sub lambda } spectra, obtained with a tangentially viewing scanning mirror/spectrometer with Reticon array readout, are consistent with cloud densities of 2*10{sup 17} cm{sup -3} or higher in the regions of strongest light emission. A spatially resolved array of D{sub alpha } detectors shows that the light variations during the pellet ablation are not caused solely by a modulation of the incoming energy flux as the pellet crosses rational q surfaces, but instead are a result of dynamic, nonstationary, ablation process. (20 refs.)
ANTONI, V., M. BAGATIN, et al. (1991). "INVESTIGATION OF THE ETA-BETA-II PLASMA EDGE BY SURFACE-ANALYSIS OF COLLECTOR PROBES." Nuovo Cimento Della Societa Italiana Di Fisica D-Condensed Matter Atomic Molecular and Chemical Physics Fluids Plasmas Biophysics 13(4): 435-453.
To investigate the ion flux escaping from the plasma and the impurity flux released by the wall, collector probes made of graphite, silicon and titanium have been exposed to the deuterium plasma confined in the toroidal device ETA BETA II. The damages on the collector surfaces have been surveyed by a scanning electron microscopy (SEM) apparatus. The deuterium and impurity retention have been measured by elastic recoil detection (ERD) and Rutherford backscattering spectroscopy (RBS) techniques respectively. The implantation build-up has been investigated as a function of the exposure time. The deuterium dose in graphite saturates after a few discharges, whereas the metal impurities exhibit a linear increase in time. The deuterium flux and its radial dependence, inferred from the implanted concentrations, have been compared with those measured by Langmuir probes. Metal impurities have been identified and their relative abundances have been compared with the material wall composition. The impurity flux is found consistent with the global content in the plasma derived by spectroscopic measurements. The deuterium dose measured in different samples has been related to the backscattering coefficient of the materials. Finally, to investigate the damage on sample probes facing the plasma particle flow, erosion probes made of vitreous graphite with silver implanted at a fixed depth have been exposed to the plasma and the thickness change after exposure recovered.
BELLAN, P. M. and M. A. SCHALIT (1991). "Can useful toroidal current be driven by classical viscoresistive Alfven waves?." Physics of Fluids B (Plasma Physics) 3(2): 423-7.
Simple, yet exact, analytic solutions for the shear and compressional Alfven wave are obtained for helical magnetohydrodynamic (MHD) waves in cylindrical geometry with both resistivity and viscosity included. The current driven by all possible combinations of these waves is examined in the quasilinear regime (i.e., where the magnetic field produced by the driven current is not self-consistently included in the equilibrium where the wave equations are derived). It is found in all cases that it is not possible to drive significant bulk axial current with small amplitude wave fields. Thus, any useful low-frequency current drive scheme will have to be based on phenomena more complicated than those discussed here. (27 refs.)
CASE, L. C. (1991). "THE REALITY OF COLD FUSION." Fusion Technology 20(4): 478-480.
Despite the unreproducibility, doubt, and controversy involved in the question of the "cold fusion" of deuterium, enough good data have been published to clearly indicate the reality of some sort of nuclear fusion. Yamaguchi and Nishioka reported a thrice-repeated event in which large amounts of heat and definite bursts of neutrons evolved simultaneously with considerable out-gassing of absorbed deuterium. These results are consistent with nuclear fusion and not with a chemical reaction. A detailed mechanism is proposed that is consistent with these events and that also generally explains many of the scattered indications of cold fusion that have been reported. There must be an adventitiously large enough presence of tritium to initiate the nuclear reaction. The results of previously successful experiments cannot now be reproduced because currently available D2O (and D2) is so low in adventitious tritium as to preclude initiation of the nuclear reaction.
CHANG, C. T. (1991). "Pellet-plasma interactions in tokamaks." Physics Reports 206(4): 143-96.
The ablation of a refuelling pellet of solid hydrogen isotopes is governed by the plasma state, especially the density and energy distribution of the electrons. On the other hand, the cryogenic pellet gives rise to perturbations of the plasma temperature and density. Based on extensive experimental data, the interaction between the pellet and the plasma is reviewed. Among the subjects discussed are the MHD activity, evolution of temperature and density profiles, and the behaviour of impurities following the injection of a pellet (or pellets). The beneficial effect of density peaking on the energy confinement time, offset by the accumulation of impurities at the plasma core is brought into focus. A possible remedy is suggested to diminish the effect of the impurities. Plausible arguments are presented to explain the apparent controversial observations on the propagation of a fast cooling front ahead of the plasma. The appearance of striations and the curving of the pellet trajectory are discussed in detail. (161 refs.)
CHRIEN, R. E., J. C. FERNANDEZ, et al. (1991). "Evidence for runaway electrons in a spheromak plasma." Nuclear Fusion 31(7): 1390-3.
The first evidence for runaway electrons in a spheromak plasma is presented; it is based on the observation of hard X-ray emission with an energy of about 1 MeV. The hard X-rays are produced in one or more bursts occurring early in the spheromak decay phase, after the coaxial plasma gun voltage is turned off. No obvious correlation is found between the amplitude of the hard X-ray emission and other spheromak parameters. Since there is no direct acceleration mechanism for electrons to reach MeV energy levels, these observations imply that the runaway electrons are confined for an acceleration time of >or=20-50 mu s and a path length of >or=5-10 km, assuming that the electrons are accelerated in the electric field of the decaying spheromak. (10 refs.)
FUJISAWA, A., H. JI, et al. (1991). "Anomalous ion temperature and plasma resistance due to MHD fluctuations in REPUTE-1 reversed field pinch plasmas." Nuclear Fusion 31(8): 1443-53.
Anomalous ion temperature and plasma resistance have been studied in reversed field pinch discharges of the REPUTE-1 device. The ion temperature measured by Doppler broadening of carbon V (2271 AA) has been observed to be anomalously high in hydrogen discharges with low electron density (n{sub e}<or=0.5*10{sup 14} cm{sup -3}); also the plasma resistance is anomalously large in these discharges. The ion temperature and the plasma resistance are shown to increase with the MHD fluctuation level. The profiles of ion temperature and soft X-ray emission have been found to be more peaked in the lower electron density regime than in the higher one. A high ion temperature and a large plasma resistance are observed simultaneously when the profile of the soft X-ray emission I{sub SX} is more peaked. (35 refs.)
HO, Y. L. and S. C. PRAGER (1991). "Nonlinear reversed-field pinch dynamics with nonideal boundaries." Physics of Fluids B (Plasma Physics) 3(11): 3099-110.
The nonlinear behavior of the reversed-field pinch bounded by a resistive shell or a distant conducting wall is investigated with a three-dimensional magnetohydrodynamic code. Nonlinear interaction between modes enhances fluctuation levels as the conducting wall is removed. The enhanced fluctuation-induced v*b electric field, which produces the dynamo effect, suppresses toroidal current, and enhances surface helicity dissipation. Thus loop voltage must increase to sustain the current and maintain helicity balance. (46 refs.)
HO, Y. L. (1991). "Numerical simulation of fluctuation suppression via DC helicity injection in a reversed field pinch." Nuclear Fusion 31(2): 341-50.
The sustainment of a reversed field pinch using electrostatic helicity injection in conjunction with the usually applied axial electric field is investigated. Nonlinear magnetohydrodynamic simulations indicate that it may be possible to sustain the plasma in a resistive steady state free from current driven modes and the associated dynamo relaxation. Moreover, the core region is well confined by nested flux surfaces. Practical considerations such as the impedance and efficiency of the injection circuit are discussed. (25 refs.)
HOKIN, S., A. ALMAGRI, et al. (1991). "Global confinement and discrete dynamo activity in the MST reversed-field pinch." Physics of Fluids B (Plasma Physics) 3(8): 2241-2246.
Results obtained on the Madison Symmetric Torus (MST) reversed-field pinch (Fusion Technol. 19, 131 (1991)) after installation of the design poloidal field winding are presented. Values of beta {sub theta e0} identical to 2 mu {sub 0}n{sub e0}T{sub e0}/B{sub theta }{sup 2}(a) approximately 12% are achieved in low-current (I=220 kA) operation; here, n{sub e0} and T{sub e0} are central electron density and temperature, and B{sub theta }(a) is the poloidal magnetic field at the plasma edge. An observed decrease in beta {sub theta e0} with increasing plasma current may be due to inadequate fueling, enhanced wall interaction, and the growth of a radial field error at the vertical cut in the shell at high current. Energy confinement time varies little with plasma current, lying in the range of 0.5-1.0 msec. Strong discrete dynamo activity is present, characterized by the coupling of m=1, n=5-7 modes leading to an m=0, n=0 crash (m and n are poloidal and toroidal mode numbers). The m=0 crash generates toroidal flux and produces a small (2.5%) increase in plasma current. (25 refs.)
KRAKOWSKI, R. A. (1991). "PROGRESS IN COMMERCIAL MAGNETIC FUSION ENERGY REACTOR DESIGNS." Fusion Technology 20(2): 121-143.
Two decades of fusion reactor conceptual design have led to a clearer definition of an "attractive" fusion power plant. Recent advances in commercial reactor designs have pushed in the direction of smaller, more compact systems while stressing material and configurational choices that amplify safety and environmental (S & E) advantages (e.g., inherent or passive safety and significantly reduced long-term radioactive waste). When intelligently amalgamated, compactness and favorable S & E characteristics can enable fusion power to be competitive. The history of fusion reactor conceptual design, the constituents of an attractive fusion end product, and recent progress in fusion reactor studies as embodied in the TITAN reversed-field pinch and the more recent and ongoing Advanced Reactor Innovations and Evaluation Study (ARIES) advanced tokamak reactor designs, are reviewed. The future for magnetic fusion energy can be bright if the right physics, technology, and materials research and development (R & D) choices are made now. An important ingredient in this "right choice" is design simplification and subsystem combination to achieve requisite levels of reliability and ease of maintenance, while ensuring competitive energy costs and acceptable S & E features. Significant departures from the "conventional" (i.e., the current R & D direction) tokamak physics embodiment are required to achieve these goals.
KUMAR, A., R. P. SHARMA, et al. (1991). "Parametric decay instabilities of the fast wave in the lower hybrid frequency regime." IEEE Transactions on Plasma Science 19(4): 590-7.
Three possible delay instabilities of the fast wave in the lower hybrid frequency regime are presented. It has been considered that a fast wave may decay into a high-frequency, lower hybrid wave and a low-frequency wave, either the ion-Bernstein wave, ion-acoustic wave, or kinetic Alfven wave. Explicit expressions for growth rates, homogeneous thresholds, and convective thresholds have been given. Applications have been pointed out in tokamak plasma during current drive experiments and magnetospheric plasmas. For example, for PLT tokamak parameters, the convective threshold for the decay instability of the fast wave into lower hybrid and ion acoustic wave is approximately 180 W/cm{sup 2}. (22 refs.)
MAROLI, C. and V. PETRILLO (1991). "Transition from ion-cyclotron to ion-acoustic longitudinal waves in a Maxwellian plasma." Physics Letters A 154(3/4): 160-3.
The roots k{sub perpendicular to } versus omega and k{sub }// of the electrostatic dispersion relation in a magnetized Maxwellian plasma in the frequency range that covers the first few ion-cyclotron harmonics, are followed when the external magnetic field is decreased down to arbitrary low values and compared with the roots of the non-magnetic dispersion relation (the ion-acoustic roots). It is found that the 'forward' branch of the magnetic case tends to the less damped ion-acoustic root and that the whole series of backward Bernstein waves become, in the limit, the second, third, etc., less damped waves in the ion-acoustic series. (13 refs.)
MAYO, R. M., J. C. FERNANDEZ, et al. (1991). "Time of flight measurement of ion temperatures in spheromaks." Nuclear Fusion 31(11): 2087-95.
The energy distribution of neutral particles is detected from spheromak plasmas by a time of flight neutral atom spectrometer. These particles originate in charge exchange interactions within the bulk plasma and carry the temperature of the majority ion component. This is the first direct measurement of temperatures of the main ion species in spheromaks. The data are found to be in the range of 500 eV during the sustainment and late decay phases and are in agreement with impurity Doppler temperatures for species expected to be localized to the plasma centre. (26 refs.)
MCNEILL, D. H., G. J. GREENE, et al. (1991). "Spectroscopic measurements of the parameters of the ablation clouds of deuterium pellets injected into tokamaks." Physics of Fluids B (Plasma Physics) 3(8): 1994-2009.
Spectroscopic measurements have been made of the parameters of the luminous cloud surrounding deuterium pellets injected into the Princeton Large Torus (PLT) (Phys. Rev. Lett. 55, 1398 (1985)) and the Tokamak Fusion Test Reactor (TFTR) (Plasma Physics and Controlled Nuclear Fusion Research (IAEA, Vienna, 1987), vol.1, p.171), with the measurements on the latter described formally here for the first time. The electron densities determined from the Stark broadening of the Balmer alpha and beta lines ranged from 5*10{sup 15}-1*10{sup 18} cm{sup -3}, while the temperatures obtained from the line to continuum intensity ratios and the ratio of the intensities of the Balmer alpha and beta lines ranged from 0.9 to 3 eV. Balmer alpha emission powers as high as 100 kW were measured. The electron temperature (1.5+or-0.2 eV) and density (3+or-1*10{sup 17} cm{sup -3}) at the time of maximum emission from intact pellets were essentially the same for the two tokamaks, although the volumes of both the discharges and the pellets were roughly ten times larger in TFTR than in PLT. The data were taken in Ohmically heated plasmas, except for a limited number of discharges in PLT with low-power (approximately half the Ohmic heating power) ion cyclotron heating for which the results are the same as with Ohmic heating alone. Various factors that may influence the results of measurements of this type are discussed, including self-absorption, low-density plasma surrounding the most intense region of an ablation cloud, fluctuations in the emission intensity, and nonMaxwellian electron distributions. The results of the present experiments are similar to measurements of the pellet cloud parameters in other machines. (65 refs.)
MILLER, G. (1991). "Equivalence of resistive evolution and kinetic dynamo models for stochastic magnetic fields." Physics of Fluids B (Plasma Physics) 3(5): 1182-5.
The kinetic dynamo model of Jacobson and Moses (Phys. Rev. A 29, 3335 (1984)) and the resistive evolution model of Miller (Phys. Fluids 31, 1133 (1988)) are compared. Fourier analyzing the x dependence in slab geometry, both models reduce to an Ohm's law for Fourier components of the form j{sub k}=F(k)E{sub k}/ eta , where j{sub k} and E{sub k} are the Fourier components of j{sub }// and E{sub }//, eta is the parallel resistivity, and F(k) is some function of k. The field line distribution function in the resistive evolution model can be chosen to make both models exactly equivalent, and implies that the characteristic field line length over which j{sub }///B is constant is the electron mean-free path. The time scale for validity of both models is t> tau , where tau is the electron collision time. The stochastic current spreading implied by these models is in rough quantitative agreement with earlier measurements. (11 refs.)
MYRA, J. R., D. A. DIPPOLITO, et al. (1991). "Sheath-plasma waves and anomalous loading in ion-Bernstein-wave experiments." Physical Review Letters 66(9): 1173-6.
It is proposed that the anomalous loading observed in tokamak ion-Bernstein-wave (IBW) experiments is due to a class of sheath-plasma waves (SPW) which propagate on high-voltage RF sheaths near the antenna. Particle simulations in one and two dimensions confirm the existence and scalings of the SPW. Coupling to the SPW modes can dominate coupling to the IBW, and qualitatively yields the observed dependences of antenna loading on parallel wave number, plasma density, and magnetic field. Antennas which minimize RF sheaths should improve the effectiveness of IBW heating. (15 refs.)
ONO, Y., M. YAMADA, et al. (1991). "Experimental study of the relaxation cycle of a decaying spheromak in an external magnetic field." Physics of Fluids B (Plasma Physics) 3(6): 1452-60.
For high electron temperature plasma discharges in the S-1 spheromak device (Plasma Physics and Controlled Nuclear Fusion Research, 1984 (IAEA, vienna, 1985), Vol.2, p.535), 'sawtooth'-like oscillations appear on signals of magnetic field, flux, q value, and electron temperature. Based on the internal magnetic field profiles measured by magnetic probe scans, the mechanisms and causes of these oscillations are revealed. The cycle of one oscillation consists of a toroidal current peaking phase and a subsequent relaxation phase. In the peaking phase, resistive current decay at the edge causes the spheromak to deviate from the initial minimum-energy Taylor state. The deviation was revealed experimentally by the preferential decay of toroidal flux over poloidal flux. A simple calculation shows that a peaking of the electron temperature profile is the most probable cause for the preferential decay of toroidal flux over poloidal flux. During the peaking phase, q decreases so low as to make the configuration unstable to low-n magnetohydrodynamic (MHD) modes (mainly n=2 mode). In the relaxation phase, these modes invoke current redistribution (relaxation), restoring the Taylor state. A significant finding in the relaxation phase is that a reversed toroidal field similar to that of the reversed-field pinch (RFP) configuration is sometimes measured at the edge of the plasma for a brief period. The disappearance or resistive decay of the reversed toroidal flux is attributed to a flux conversion through the magnetic reconnections caused by the low-n modes. (23 refs.)
PHILLIPS, J. A., D. A. BAKER, et al. (1991). "Experimental observations on plasma-circuit interactions and their effect on low frequency fluctuations in the reversed field pinch ZT-40M." Nuclear Fusion 31(8): 1556-62.
Experimental observations of the amplitude and frequency of low frequency magnetic field and soft X-ray fluctuations in the reversed field pinch (RFP) ZT-40M, which has toroidal and poloidal gaps in its stabilizing shell, have shown that for high values of the pinch parameter these fluctuations can be strongly affected by the configuration of the toroidal field circuit. Parallel connection of coil sections improves the toroidal flux uniformity, which results in a reduction of the fluctuation level, the toroidal loop voltage and the implied energy losses compared with those of coils connected in a series or series-parallel configuration. These results indicate that the operation of RFP experiments with shell gaps and field coils well removed from the plasma can be further improved by the addition of close fitting, parallel connected, low current coils. (21 refs.)
RUSBRIDGE, M. G. (1991). "The relationship between the 'tangled discharge' and 'dynamo' models of the magnetic relaxation process." Plasma Physics and Controlled Fusion 33(12): 1381-9.
It is shown that these two models are closely related. The 'Tangled Discharge Model' (TDM) necessarily involves a non-vanishing <v*b>, the so-called 'dynamo' effect, while the effective functioning of this term in the 'Dynamo' models is shown to require a stochastic magnetic field structure. There is no inconsistency between this conclusion and the fact that the fully-relaxed state may be described by analytical forms such as the 'Bessel Function Model' which apparently contain good flux surfaces. The TDM is not a satisfactory description of the behaviour of present relaxed-state systems, but might be appropriate in the limit of large Lundquist number S. (25 refs.)
SCHMID, P., G. BARRICK, et al. (1991). "Flux generation in ultra-low-q tokamak discharges." Physics of Fluids B (Plasma Physics) 3(4): 1113-15.
Spontaneous generation of toroidal flux is observed in a reversed-field pinch device operated as an ultra-low-q tokamak with a safety factor on axis of 1/4. An initial toroidal bias field of 150 G is increased to 600 G on axis in 300 mu sec while the field at the wall is held nearly constant. Sawteeth are observed, which indicate cyclic magnetic reconnection. (19 refs.)
SRINIVASAN, M. (1991). "Nuclear fusion in an atomic lattice: an update on the international status of cold fusion research." Current Science 60(7): 417-39.
It is now two years since the first reports of the occurrence of nuclear reactions at ambient temperatures in deuterated metals such as PD or Ti were published. 'Cold fusion', as this phenomenon has now come to be known, has, however, become embroiled in intense controversy with scientific community becoming sharply polarized into 'believers' and 'non-believers' of this novel phenomenon. This ambivalence is primarily because of the non-reproducibility of the claimed results by many reputed research groups that have often used sophisticated experimental equipment. However, as the present review clearly shows, a large number of laboratories in many different countries have now obtained very reliable experimental evidence confirming the generation of 2.45-MeV neutrons, tritium, charged particles, X-rays, etc., both in electrolysis experiments and in a variety of other D{sub 2}-/plasma/ion-beam-loading experiments, thereby confirming the nuclear origin of the phenomenon. These experimental results are such that they cannot be dismissed as being 'experimental artefacts' any more. It is understandable that the scientific community finds it difficult to accept a phenomenon that is not repeatable at will as 'science'. It would seem that the sporadicity of the results is due to some as yet unknown parameters that seem to be controlling the onset of nuclear phenomena in solid deuterated matrices. (174 refs.)
STORMS, E. (1991). "REVIEW OF EXPERIMENTAL-OBSERVATIONS ABOUT THE COLD FUSION EFFECT." Fusion Technology 20(4): 433-477.
The experimental literature describing the cold fusion phenomenon is reviewed. The number and variety of careful experimental measurements of heat, tritium, neutron, and helium production strongly support the occurrence of nuclear reactions in a metal lattice near room temperature as proposed by Pons and Fleischmann and independently by Jones.
TSUI, H. Y. W., C. P. RITZ, et al. (1991). "Fluctuations and transport in a reversed field pinch edge plasma." Nuclear Fusion 31(12): 2371-82.
Edge fluctuations are characterized and their associated transport is determined from Langmuir probe measurements in the ZT-40M reversed field pinch. It is found that the fluctuations have high normalized amplitudes (e mod phi {sub f} mod /T{sub e}=1.1, mod T{sub e} mod /T{sub e}=0.5 and mod n mod /n=0.4). There are significant contributions from magnetic perturbations acting on the equilibrium gradients. Compared to the global estimates, the fluctuation driven particle flux is large, whereas the corresponding electron energy flux is not. In the limiter shadow, the equilibrium density and electron temperature scale lengths are shorter and the fluctuation levels are higher. The fluctuation driven particle flux in the limiter shadow is 60% less than that in the plasma edge; most of the reduction is in the low frequency spectral region, which is where global MHD magnetic flux fluctuations are strongest. (31 refs.)
VEERASINGAM, R., K. E. HAYNES, et al. (1991). "Impurity studies on ZT-40M, ZTH and RFX using a zero-dimensional model for the reversed field pinch." Nuclear Fusion 31(9): 1735-45.
A time dependent zero-dimensional model of plasma transport for reversed field pinch devices is developed to perform parametric studies on next generation experiments. The model includes energy balance equations for electrons and ions. Particle density and current are specified quantities. Impurities are treated using a coronal nonequilibrium model. Particle leakage and recycling are included for plasma and impurities. The particle lifetime is a specified parameter. The energy confinement times for electrons and ions are specified as multiplicative coefficients times the Ohmic energy confinement time. Ohmic heating is enhanced by an anomaly factor, and a direct ion heating term is included via anomalous resistive heating represented by a multiplicative coefficient times Ohmic heating. A single set of coefficients is found to simulate a variety of ZT-40M runs with different currents and wall conditions. The resulting model is employed to perform parametric analysis on the behaviour of ZTH at 2 MA operation. (27 refs.)
WEBER, P. G., J. C. INGRAHAM, et al. (1991). "Effects of limiters on reversed-field pinch confinement." Physics of Fluids B (Plasma Physics) 3(7): 1701-7.
The authors document the effects on confinement of introducing discrete limiters into the edge plasma of the ZT-40M reversed-field pinch (RFP) (Fusion Technol. 8, 1571 (1985)). RFP confinement is not significantly degraded by appropriately designed single limiters inserted to a sufficient depth for effective local vacuum vessel wall protection. Inserting limiters too deeply into the plasma results in excess limiter heating, and a consequent increase in the impurity content of the plasma. Under these conditions the plasma loop voltage increases. The heating of the limiters is observed to be asymmetric, with the majority of the heat flux in ZT-40M being attributable to suprathermal electrons (T{sub supra} approximately 2-3*T{sub e} (0)) reaching the edge moving almost unidirectionally along magnetic-field lines. (19 refs.)
WONG, K. L. (1991). "An alpha particle diagnostic based on measurements of lower-hybrid wave fluctuations." Physics of Fluids B (Plasma Physics) 3(6): 1501-3.
It is shown that the one-dimensional alpha particle velocity distribution function can be determined from a first-order differential equation based on measurements of lower-hybrid wave fluctuations in an equilibrium plasma. This method uses collective Thomson scattering data from an array of detectors placed on a prescribed curve. It is applicable only when the alpha particles have an isotropic velocity distribution. (9 refs.)
YAGI, Y., Y. HIRANO, et al. (1991). "Plasma current and energetic electrons in the core plasma of a reversed field pinch." Plasma Physics and Controlled Fusion 33(12): 1391-406.
The electron distribution function is statistically examined from measurements by Thomson scattering at +or-7 cm off-axis points (about half of the minor radius, 13.5 cm) from the center of the vacuum vessel of the TPE-1RM15 reversed field pinch (RFP) machine. Four different RFP discharge conditions are examined at both ports on a shot-by-shot basis. The wavelength shift of the distribution function and corresponding poloidal current density are deduced so that the same bulk electron temperatures at both the inside (-7 cm) and outside (+7 cm) ports are obtained. The deviation of the distribution function from the shifted Maxwellian is examined from the difference in fitted temperature with different weighting functions. The results indicate that in two of the experimental groups, the plasma current density carried by the tail electrons above 2.6 keV exceeds that carried by the shifted bulk Maxwellian distribution. For all of the data groups, the ratio of the current density carried by tail electrons to the total current density has a positive correlation with both E/E{sub C} (=0.14-0.22) and eta {sub k}/ eta {sub S}, where E{sub C} is the critical electric field for thermal electrons to run away, and eta {sub k} and eta {sub S}, are the plasma resistivity estimated from the helicity balance equation and Spitzer's formula, respectively. These tail electrons observed in the core plasma of an RFP can be attributed to the origin of the high-energy electrons recently observed at the edge region of RFPs. (31 refs.)
ANTONI, V., M. BAGATIN, et al. (1992). "Langmuir and heat flux probes for RFX." Review of Scientific Instruments
9TH TOPICAL CONF ON HIGH TEMPERATURE PLASMA DIAGNOSTICS 63(10): 4711-4713.
A movable limiter made of graphite and instrumented with an array of Langmuir and heat flux probes has been developed for the reversed field pinch experiment RFX. The limiter is mushroom shaped and designed to be exposed to the edge plasma for a discharge duration approximately=250 ms. The distribution of the probes on the limiter surface allows simultaneous collection, in a single discharge, of edge parameters at different angles and radial positions. A bellows drive system with three degrees of freedom allows the instrumented limiter to be rotated and translated to investigate the spatial structure of the edge plasma. Cables and feedthroughs have been designed for 1-kV voltage and 5-A current. The power supply provides an output voltage +or-100 V and a maximum current 4 A. Specific software has been developed for the automatic analysis of the Langmuir probes data. (6 refs.)
ANTONI, V., M. BAGATIN, et al. (1992). "PARTICLE AND ENERGY-BALANCE IN THE SOL GENERATED BY A LIMITER IN A RFP PLASMA." Journal of Nuclear Materials
10TH INTERNATIONAL CONF ON PLASMA-SURFACE INTERACTIONS IN CONTROLLED FUSION DEVICES ( PSI-10 ) 196: 577-580.
The plasma outer region of the RFP experiment ETA BETA II has been extensively investigated. In particular by an insertable graphite limiter, instrumented with thermocouples and Langmuir probes, the energy and particle fluxes in the shadow of the limiter have been measured at different insertions. The results are compared with those obtained by small-sized calorimeter/Langmuir probes inserted to investigate, with fine space resolution, the outer region without limiter. Almost 80% of the power to the limiter is estimated to be carried by fast suprathermal electrons flowing along the magnetic field line direction. From an energy balance equation, applied in the SOL locally generated, the connection length of the limiter has been derived. Thus a particle balance has been applied to determine the particle diffusion coefficient at the plasma edge.
ANTONI, V., M. BAGATIN, et al. (1992). "Energy flux investigation in the outer region of the ETA BETA II experiment." Plasma Physics and Controlled Fusion 34(5): 699-714.
The energy transport in the outer region of the RFP experiment ETA BETA II has been experimentally investigated by measuring the energy flux intercepted by the surface of an insertable limiter, instrumented with heat flux probes. It has been found that the energy deposited on the surface facing the electron flow is almost an order of magnitude larger than that deposited on the opposite side. The spatial dependence of the energy flux has been studied and the results are reported for different main plasma densities. The results are explained by fast electrons flowing in the outer region of the plasma, parallel to the magnetic field in agreement with the results found in other RFP experiments. (25 refs.)
ANTONI, V., M. BAGATIN, et al. (1992). "Energy transport and magnetic stochasticity at the edge of a plasma in RFP configuration." Plasma Physics and Controlled Fusion 34(11): 1639-50.
The energy flux carried by energetic electrons at the edge of the RFP experiment ETA BETA II has been derived by comparing the total heat collected by calorimetric probes oriented parallel and antiparallel to the magnetic field. Assuming that the energetic electrons flow along the stochastic magnetic field lines without undergoing collisions, the magnetic diffusion coefficient has been inferred from the radial behaviour of the energy flux. The diffusion coefficient is then compared with the magnetic fluctuation amplitude measured at the edge for different reversal of the toroidal field at the wall. It is found that the magnetic diffusion coefficient increases with magnetic fluctuations and the energy flux carried by energetic electrons decreases with Theta. (23 refs.)
ARIMOTO, H., A. NAGATA, et al. (1992). "Coupling of tearing modes in normal RFP discharge." Journal of the Physical Society of Japan 61(4): 1202-10.
The dynamic process of the coupling of tearing modes in the normal RFP discharge is studied with the help of 3-D MHD simulations. Almost all studies on the mechanism of the dynamo and the mode coupling have been carried out in the special operation mode of 'the deep reversal discharge' where the toroidal field at the wall is decreased. The dynamic process in the normal (ordinary) RFP discharge is reported. The process is as follows: the m=1 modes resonant near the plasma axis start growing in the initial stage. After growing enough, they interact with each other and the subsequent reconnection of magnetic field lines takes place. Consequently, m=0 modes are excited. In the normal RFP discharge, these couplings of tearing modes are frequently and periodically generated.
BRUNDTLAND, T. (1992). "A description of the vacuum chamber for the Blaamann plasma experiment." Vacuum 43(3): 185-7.
The construction and operation of the toroidal vacuum chamber for the Blaamann plasma experiment is described. The lowest pressure achieved was approximately 5*10{sup -9} mbar which compares well with the theoretically calculated value of 1.3*10{sup -9} mbar. (3 refs.)
CARDINALI, A. and F. ROMANELLI (1992). "Linear propagation and absorption of ion Bernstein waves in toroidal geometry." Physics of Fluids B (Plasma Physics) 4(3): 504-11.
The ion Bernstein wave (IBW) propagation is investigated in toroidal geometry with particular regard for the absorption mechanism of the power along the trajectory which shows a strong dependence on the value of the poloidal angle at the antenna. A large power fraction launched by poloidally wide antennas turns out to be absorbed in the very edge of the plasma by electron Landau damping and no penetration of the wave up to the plasma core is allowed except for rays starting at poloidal angles very close to the horizontal plane. This result could provide an explanation for the absence of efficient core heating in some of the IBW heating experiments. (15 refs.)
CRADDOCK, G. G. and P. H. DIAMOND (1992). "Theory of kinetic Alfven wave helicity injection and current drive." Physics of Fluids B (Plasma Physics) 4(8): 2560-6.
Kinetic shear Alfven wave (KSAW) helicity injection motivated by the goal of simultaneous implementation of radio frequency current and flow (Phys. Rev. Lett. 67, 1535 (1991)) drive is analyzed. The quasilinear helicity flux results in a net helicity increase only via transport through the boundaries. This, in turn, requires a compressional component at the plasma edge boundary and electron dissipation at the Alfven resonance, as well as throughout the region, where helicity transport occurs. A comparison is made to direct KSAW current drive, which can add constructively with helicity injection by tailoring of the sign of poloidal and parallel wave numbers. The KSAW helicity flux is related to the alpha effect due to MHD (magnetohydrodynamic) resistive kink and shear-Alfven-driven helicity fluxes. The helicity flux need not have a zero flux surface average. Finally, the efficiency (I{sub p}R/P{sub abs}) of helicity injection by the KSAW is found to be significantly smaller than that by viscoresistive shear Alfven waves, and scales as (I{sub p}R/P{sub abs}) approximately n{sub 0}{sup -3/2}B{sub 0}{sup 3}, where n{sub 0} is the density and B{sub 0} is the toroidal magnetic field. Application to edge current drive for ELM control is discussed. (19 refs.)
DUBOIS, M. A., R. SABOT, et al. (1992). "Determination of the safety factor profile in TORE SUPRA from the largest striations observed during pellet ablation." Nuclear Fusion 32(11): 1935-40.
The largest striations in H{sub alpha }(D{sub alpha }) light observed across the ablation cloud of injected hydrogen pellets can be explained in most cases by the presence of resonant magnetic surfaces. The authors suggest that it is possible to identify these striations and therefore to obtain a very detailed profile of the safety factor. This q profile exhibits shear plateaus related to rational q values, thus indicating the presence of magnetic turbulence. (22 refs.)
FISCHER, B. and M. KRAMER (1992). "Study of lower-hybrid wave propagation by CO{sub 2}-scattering and RF probe diagnostics." Plasma Physics and Controlled Fusion 34(9): 1467-92.
Coherent CO{sub 2}-scattering as well as phase-sensitive probe diagnostics are applied to investigate small-amplitude lower-hybrid (LH) waves launched by a slow-wave antenna to a linear plasma column. Special attention is focussed on the detailed description of the LH wave field in the plasma taking into account the wavenumber spectrum of the antenna. With the knowledge of the LH density perturbations, the RF detector current is computed under the actual scattering conditions. To fit the theoretical results to the experimental results, an effective collisional frequency nu {sub eff} for electron momentum transfer is introduced. The observed LH wave damping is considerably higher than expected from collisional and electron Landau damping, but it is correlated with the level of low-frequency fluctuations. In particular, the scattering and the probe signals are strongly affected by the damping as well as by the low-frequency turbulence itself. (21 refs.)
HOKIN, S. (1992). "An X-ray target probe for superthermal electron and field line pitch measurements." Review of Scientific Instruments
9TH TOPICAL CONF ON HIGH TEMPERATURE PLASMA DIAGNOSTICS 63(10): 5041-5044.
An insertable target X-ray probe has been designed which may be used for two different purposes: measurement of superthermal electron temperature and drift speed, and measurement of the local magnetic field line pitch, both with fast time resolution (1 mu s). The diagnostic is intended for use on the Madison Symmetric Torus (MST) reversed-field pinch. For the first application, the probe is fitted with six detectors, three viewing the part of the target facing toward the incident electrons, and three viewing the part facing away; each set of three has 0.3, 0.6 and 1.2 mil Be filters for energy discrimination. By subtracting signals from detectors with the same filter but opposing views and dividing signals from detectors with different filters but the same view, one can derive the electron drift speed and parallel temperature. For the second application, the six detectors are replaced by a single position-sensitive detector, which produces a signal indicating the position of the 'hot spot' on the target, and therefore the field line pitch. The diagnostic can withstand high temperatures and can therefore be inserted further into the plasma than electrostatic probes. (4 refs.)
JI, H., H. TOYAMA, et al. (1992). "Fluctuation and edge-current sustainment in a reversed-field pinch." Physical Review Letters 69(4): 616-19.
The simple Ohm's law eta j{sub }//=E{sub }// is not satisfied in the a/2<or approximately=r<or=a region of the REPUTE-1 reversed-field-pinch plasma. Fluctuation-induced electric fields, such as (v*B), are not sufficient to account for eta j{sub }//-E{sub }// at the edge. Current diffusion due to magnetic fluctuations, carried by fast electrons rather than bulk cold electrons, sustains the edge parallel equilibrium current. (21 refs.)
LENGYEL, L. L. (1992). "Evolution of particle clouds around ablating pellets in magnetically confined hot plasmas." IEEE Transactions on Plasma Science 20(6): 663-8.
Cryogenic hydrogen isotope pellets are being used for introducing fuel particles into the plasma interior in magnetic confinement fusion experiments. The spatial and the time evolution of the initially low-temperature, high-density particle clouds forming around such pellets are considered, with particular attention given to such physical processes as heating of the clouds by the energy fluxes carried by incident plasma particles, gas-dynamic expansion with j*B-produced deceleration in the transverse direction, finite-rate ionization and recombination processes, and magnetic field convection and diffusion. While the dynamic processes associated with the ionization and radial confinement processes are characterized by the relatively short Alfven time scale ( mu s range), the subsequent phase of axial expansion is associated with a notably larger hydrodynamic time scale defined by the heat input and gas-dynamic expansion rates (ms range). Data stemming from experimental measurements in toroidal confinement machines are compared with results of model calculations. Some similarities with extraterrestrial plasma scenarios, such as the earlier magnetospheric barium release experiments, are discussed. (23 refs.)
MITARAI, O., A. HIROSE, et al. (1992). "Plasma density at the current reversal in the STOR-1M tokamak with AC operation." Nuclear Fusion 32(10): 1801-9.
The plasma density behaviour in the STOR-1M tokamak with alternating current (AC) operation is described using the Murakami-Hugill diagram (1/q{sub a}, nR/B{sub t}). At the current reversal, I{sub p}=0 (1/q{sub a}=0), the plasma density remains finite and the Murakami parameter is nR/B{sub t}=(0.66+or-0.22)*10{sup 18}m{sup -2}.T{sup -1}. Gas puffing before the current reversal does not noticeably increase the plasma density at the current reversal, but allows AC operation with larger currents and improves its reproducibility. A qualitative explanation for the finite plasma density at the current reversal is given on the basis of a short circuit effect by the limiter. (18 refs.)
ORTOLANI, S. (1992). "The behaviour of fast electrons in reversed field pinches." Plasma Physics and Controlled Fusion
9TH KIEV INTERNATIONAL CONF ON PLASMA THEORY / 9TH INTERNATIONAL CONGRESS ON WAVES AND INSTABILITIES IN PLASMAS / 19TH EPS CONF ON CONTROLLED FUSION AND PLASMA PHYSICS 34(13): 1903-8.
High energy electrons are typically measured in the outer region of Reversed Field Pinch plasmas. An explanation for the origin of this small population (2-10%) of fast electrons can be proposed on the basis of the models used to describe the relaxation process of the magnetic field profiles. In particular, stochastic models where electrons can travel unconfined from the plasma core to the edge, and MHD models where fast electrons originate in high current density reconnection layers, are briefly reviewed. (22 refs.)
REMPEL, T. D., A. F. ALMAGRI, et al. (1992). "Turbulent transport in the Madison Symmetric Torus reversed-field pinch." Physics of Fluids B (Plasma Physics)
33RD ANNUAL MEETING OF THE DIVISION OF PLASMA PHYSICS OF THE AMERICAN PHYSICAL SOC 4(7): 2136-2141.
Measurements of edge turbulence and the associated transport are ongoing in the Madison Symmetric Torus (MST) reversed-field pinch (Fusion Technol. 19, 131 (1991)) using magnetic and electrostatic probes. Magnetic fluctuations are dominated by m=1 and n approximately 2R/a, tearing modes. Particle losses induced by magnetic field fluctuations have been found to be ambipolar ((J{sub }//B{sub r})/B{sub 0}=0). Electrostatic fluctuations are broadband and turbulent, with mode widths Delta m approximately 3-7 and Delta n approximately 70-150. Particle, parallel current, and energy transport arising from coherent motion with the fluctuating E*B drift have been measured. Particle transport via this channel is comparable to the total particle loss from MST. Energy transport (from (PE)/B{sub Q}) due to electrostatic fluctuations is relatively small, and parallel current transport (from (J{sub }//E{sub phi })/B{sub 0}) may be small as well. (20 refs.)
SCIME, E., S. HOKIN, et al. (1992). "Ion heating and magnetohydrodynamic dynamo fluctuations in the reversed-field pinch." Physical Review Letters 68(14): 2165-7.
Ion temperature measurements, time resolved to 10 mu s, have been made in the Madison Symmetric Torus (MST) reversed-field pinch with a five-channel charge-exchange analyzer. The ion temperature, T{sub i} approximately=300 eV for I=360 kA, increases by as much as 100% during discrete dynamo bursts in MST discharges. Magnetic-field fluctuations in the range 0.5-5 MHz were also measured. Structure in the fluctuation frequency spectrum at the ion cyclotron frequency appears as the bursts terminate, suggesting that the mechanism of ion heating involves the dissipation of dynamo fluctuations at ion gyro-orbit scales. (15 refs.)
SCIME, E., M. CEKIC, et al. (1992). "Ion heating and magnetohydrodynamic dynamo fluctuations in the reversed-field pinch." Physics of Fluids B (Plasma Physics) 4(12): 4062-71.
Ion temperatures have been measured in the Madison Symmetric Torus (MST) (Dexter et al., Fusion Technol. 19, 131 (1991)) reversed-field pinch (RFP) with a five channel charge exchange analyzer. The characteristic anomalously high ion temperature of RFP discharges has been observed in the MST. The ion heating expected from ion-electron collisions is calculated and shown to be too small to explain the measured ion temperatures. The charge exchange determined ion temperature is also compared to measurements of the thermally broadened CV 227.1 nm line. The ion temperature, T{sub i} approximately=250 eV for I=360 kA, increases by more than 100% during discrete dynamo bursts in MST discharges. Magnetic field fluctuations in the range 0.5-5 MHz were also measured during the dynamo bursts. Structure in the fluctuation frequency spectrum at the ion cyclotron frequency suggests that the mechanism of ion heating involves the dissipation of dynamo fluctuations at ion cyclotron frequencies. (30 refs.)
SHARMA, R. P., A. KUMAR, et al. (1992). "Some parametric decay instabilities of fast magnetosonic waves in the ion cyclotron harmonic frequency regime." Physics of Fluids B (Plasma Physics) 4(1): 79-85.
In the present paper, some possible parametric decay instabilities of fast magnetosonic waves (FMSW) near the second harmonic of ion cyclotron frequency have been considered in one-ion species hydrogen plasmas. The decay channels of FMSW include an ion Bernstein wave as a high-frequency decay wave and a low-frequency decay wave either as an ion-acoustic wave or kinetic Alfven wave. Applications have been pointed out to the ASDEX (Plasma Phys. Controlled Fusion 28, 235 (1986)) and ACT-1 (Rev. Sci. Instrum. 53, 4 (1982)) toroidal devices where parametric decay instabilities of FMSW near the second harmonic of ion cyclotron frequency have been observed in the scrape-off layer and near the edge plasma. It has been shown that the growth rate is sufficiently high at the edge and may thus contribute to the energy deposition to the edge plasma. A comparison between various decay processes has been discussed for different toroidal devices. Applications have also been pointed out to Earth's magnetospheric plasmas where some magnetic fluctuations have been observed. (37 refs.)
SUGAWA, M. (1992). "Numerical study of self-interaction of Bernstein waves by nonlinear Landau damping." Journal of Plasma Physics 48: 465-476.
When Bernstein waves (B waves) are excited in a magnetized plasma, their self-interaction by nonlinear Landau damping (NLD) becomes the dominant mechanism for the electron heating of the bulk plasma. The authors examine this behaviour numerically. This occurs only for B waves with relatively small k{sub }//, because the damping of the B waves becomes very small. This occurs in the relatively broad B-wave frequency range between omega / omega {sub c}=1.45 and 1.78. For B waves with large k{sub }// (k{sub }//R>0.15), virtual waves are not generated via self-interaction due to NLD because the quasi-linear cyclotron damping of the B waves becomes the dominant mechanism. The numerical results agree well with experimental ones. (16 refs.)
TERRY, J. L., E. S. MARMAR, et al. (1992). "Imaging of lithium pellet ablation trails and measurement of q profiles in TFTR." Review of Scientific Instruments
9TH TOPICAL CONF ON HIGH TEMPERATURE PLASMA DIAGNOSTICS 63(10): 5191-5194.
Video images with 2 mu s exposures of the Li{sup +} emission in Li pellet ablation clouds have been obtained in a variety of Tokamak Fusion Test Reactor tokamak discharges. The pellet clouds are viewed from behind the pellet, which is injected from the outside midplane. In this view, the emission forms an elongated cigar shape with the long dimension of the cigar aligned with the local magnetic field. In some cases, two distinct parallel cigars can be seen simultaneously, displaced vertically from one another by approximately 5 cm. Measurements using a ten channel array of position sensitive photodiodes show that the mean position of the ablation cloud emission can oscillate vertically by approximately 4 cm with periods in the 60-100 mu s range, and that these oscillations are highly correlated with 'bursts' in the cloud emission. The tilt of the cloud is also measured as a function of time as the pellet traverses the plasma, and in this way the poloidal field profile is obtained. (The total transit time of the pellet is approximately 1 ms.) Magnetohydrodynamic equilibrium reconstructions of q profiles have been determined using these measurements. (13 refs.)
YASAKA, Y., H. TAKENO, et al. (1992). "Observation of ion-cyclotron heating due to mode conversion of fast waves in a simple mirror." Physics of Fluids B (Plasma Physics) 4(6): 1486-91.
The observation of mode conversion of the m=-1 (right-rotating propagation) fast magnetosonic wave into the slow ion-cyclotron wave via the ion-ion hybrid resonance in a mirror magnetic field is presented. The left-circularly polarized component of the wave magnetic field strongly depends on the minority ion concentration in a helium-hydrogen plasma. The associated ion heating up to 250% of the initial value is in correlation with the enhancement of the left-circularly polarized component of the wave field, showing evidence of ion cyclotron heating due to the mode conversion. The two-dimensional calculation of the wave propagation and mode conversion is performed and compared with the experiment. (13 refs.)
ALBA, S., M. FONTANESI, et al. (1993). "Experimental investigation of minority ion concentration in a weakly ionized hydrogen plasma." Plasma Physics and Controlled Fusion 35(2): 263-8.
H{sub 3}{sup +} and H{sub 2}{sup +} ion minority concentrations in a weakly ionized hydrogen plasma are determined through the comparison between the measured wavelength of the neutralized in Bernstein waves and the theoretical one, given by the solution of the complete electrostatic dispersion relation. A simple model based on the resolution of fluid balance equations permits one to explain the presence of a high percentage of H{sub 3}{sup +} ( approximately=30%). (12 refs.)
ANDO, T., A. NAGATA, et al. (1993). "New features of the reversed-field pinch dynamo." Journal of the Physical Society of Japan 62(9): 2979-82.
New features of dynamo action were found in normal- Theta discharge on the STP-3M reversed-field pinch (RFP), where Theta =B{sub theta }(a)/(B{sub z}) is the pinch parameter. The results indicate that as the magnetic Reynolds number (S) increases, the self-organization effect works smoother in the RFP plasma, hence the toroidal magnetic flux increases continuously and uniformly in the toroidal direction. Furthermore, the effective inductance in the RFP plasma increases with increasing S and is kept constant during the plasma current duration. These features lead to the conclusion that usual discrete dynamo action does not play an essential role in the RFP dynamo.
ANTONI, V., L. APOLLONI, et al. (1993). "Confinement studies on RFX." Plasma Physics and Controlled Fusion
20th European Conference on Controlled Fusion and Plasma Physics 35: B333-42.
The results of the first year of operation of the experiment RFX are reported. Profiles of electron density, electron and ion temperature and impurity emission have been measured at plasma current I<0.7 MA. The energy confinement parameters at different density are reported, the best values ( tau {sub E} approximately 1ms, beta {sub theta } approximately 8%) being obtained operating at higher density. The role of the impurity content in determining the present performance of the experiment is discussed. (26 refs.)
ARIMOTO, H., A. NAKAMURA, et al. (1993). "Macroscopic behavior and discrete dynamo in high- Theta reversed-field pinch discharges." Physics of Fluids B (Plasma Physics) 5(6): 1836-41.
The magnetohydrodynamic (MHD) activity and the discrete dynamo in high- Theta reversed-field pinch (RFP) discharges are studied through comparisons with those in normal- Theta RFP discharges, where Theta =B{sub theta }(a)/(B{sub z}) is the pinch parameter. In high- Theta RFP discharges, the mode coupling of m=1 is enhanced by the high magnetic shear configuration, and the MHD activity is more turbulent as compared with that in normal- Theta RFP discharges. Furthermore, the discrete increase of the toroidal magnetic flux is obviously observable. The increase of the toroidal magnetic flux generated by the discrete dynamo, which has an asymmetric structure in the toroidal direction, is only 1%-2% of the whole toroidal magnetic flux. The experimental results indicate that the discrete dynamo does not play an essential role in the RFP dynamo. (17 refs.)
BAGATIN, M., D. DESIDERI, et al. (1993). "The power supply system for Langmuir probes on RFX." Measurement Science & Technology 4(11): 1269-74.
A power supply system for Langmuir probes to be used on the reversed field pinch (RFP) thermonuclear fusion experiment RFX has been developed. It is based on a full-bridge converter concept, using a pulse width modulation (PWM) technique, and is characterized by a 5 A, +or-150 V rating. The power supply has been designed to ensure compatibility with the electromagnetic noise close to the machine and with the plasma, which behaves as a nonlinear rapidly varying load. The system has been optimized by numerical simulations. Tests in the laboratory and on the experiment have been performed in a wide range of working conditions and the system has been proven to suit fairly well the RFX experimental conditions. (20 refs.)
BALDZUHN, J. and W. SANDMANN (1993). "Analysis of luminescence signals observed during pellet injection." Plasma Physics and Controlled Fusion 35(10): 1413-31.
We report on a numerical analysis of luminescence signals observed during pellet injection into magnetically confined plasmas. Data from the ASDEX tokamak and the Wendelstein W7-AS advanced stellarator are considered in the form of a comparative study. We investigate how far the safety factor q and other plasma parameters influence the luminescence signals. Thus, we try to clarify the validity of the so-called neutral gas shielding model and an extension of it that takes into account the influence of resonant surfaces on the ablation process. Resonant values play only a minor role in the appearance of minima in the luminescence signals. (28 refs.)
BIGLARI, H. and P. H. DIAMOND (1993). "A MEAN-FIELD OHM LAW FOR COLLISIONLESS PLASMAS." Physics of Fluids B-Plasma Physics 5(11): 3838-3840.
A mean field Ohm's law valid for collisionless plasmas is derived kinetically. It is shown that contrary to conventional thinking, the resulting hyperresistivity is significantly smaller than its fluid counterpart due to the fact that the turbulent decorrelation rate is linked to the rapid electron ballistic motion rather than the slower nonlinear mixing time. Moreover, the off-diagonal contributions to the parallel electron momentum flux are shown to result in Ohm's law renormalizations that dwarf the current diffusivity and break radial parity symmetry.
BRAMBILLA, M. (1993). "Modelling heating and current drive in the ion cyclotron frequency range." Plasma Physics and Controlled Fusion
EUROPHYSICS TOPICAL CONF ON RADIOFREQUENCY HEATING AND CURRENT DRIVE OF FUSION DEVICES 35: A141-65.
Numerical simulation of heating and current drive in the ion cyclotron frequency range has progressed to the point that both the propagation and absorption of IC waves on the one hand, and the evolution of ion and electron distribution functions due to interaction with the HF fields on the other hand, can be described in realistic geometry and to some extent consistently with each other. The author reviews work in this field, with particular emphasis on the derivation and numerical solution of adequate equations to describe wave propagation in hot plasmas in toroidal geometry. Codes solving the quasilinear kinetic equations and their interface to wave propagation codes are also briefly mentioned. The complete problem is so complex, however, that in most cases simplified models must be used for extensive numerical studies of IC heating and current drive scenarios. Plane stratified models are appropriate to investigate antenna loading and global power balance; ray tracing gives excellent results in large hot plasmas and are easily interfaced to quasilinear kinetic codes. Full-wave codes in toroidal geometry are often simplified by neglecting either finite Larmor radius effects or the poloidal component of the static magnetic field, or both. (73 refs.)
DENHARTOG, D. J., M. CEKIC, et al. (1993). "B{sub 4}C solid target boronization of the MST reversed-field pinch." Journal of Nuclear Materials 200(2): 177-83.
A solid rod of hot-pressed boron carbide is being used as the source of boron during boronization of MST. The most striking result of this procedure is the reduction in oxygen contamination of the plasma (O III radiation, characteristic of oxygen at the edge, falls by about a factor of 3 after boronization). The radiated power fraction drops to about half its initial value. Particle reflux from the wall is also lowered, making density control simpler. The rod (12.7 mm diameter) is inserted into the edge plasma of normal high-power RFP discharges. B{sub 4}C is ablated from the surface of the rod and deposited in a thin film (a-B/C:H) on the walls and limiters. The energy flux carried by 'superthermal' (not 'runaway') electrons at the edge of MST appears to enhance the efficient, nondestructive ablation of the boron carbide rod. (26 refs.)
GREENE, P. and S. ROBERTSON (1993). "Spectra and growth rates of resistive wall modes in a reversed-field pinch." Physics of Fluids B (Plasma Physics) 5(2): 556-63.
The Reversatron II (IEEE Trans. Plasma Sci PS-16 667 (1988)) reversed-field pinch has been operated with shells having magnetic penetration time comparable to and shorter than the discharge duration. The spectra and growth rates of the resistive wall modes have been measured and compared for the two cases. The growth rates of modes are typically in the expected range gamma tau {sub s}=0.5-2 where gamma is the growth rate and tau {sub s} is the magnetic penetration time of the shell. The spectra, however, are broader than expected from linear theory and there are indications that both the growth rates and the mode spectra are affected by nonlinearities. (21 refs.)
HOFFMAN, A. L. and J. T. SLOUGH (1993). "Field reversed configuration lifetime scaling based on measurements from the Large s Experiment." Nuclear Fusion 33(1): 27-38.
Flux, energy and particle lifetimes have been measured in the new Large s Experiment field reversed configuration (FRC) facility. By careful control of the formation process, it was possible to form symmetric, quiescent FRCs, with s values higher than 4, in the one year of operation of the device. A wide range of plasma conditions was achieved, with ion temperatures varying between 0.1 and 1.5 keV. The lifetimes continue to scale approximately with the r{sub s}{sup 2}/ rho {sub i} parameter found in earlier work, with a coefficient proportional to x{sub s} to a power between 0.5 and 1. (22 refs.)
JAIN, K. K. (1993). "Observation of improved behavior by electrode biasing of a toroidal plasma having no poloidal magnetic field." Physical Review Letters 70(6): 806-9.
The author observes an increase in plasma density and a significant reduction in fluctuations with bias electrode in a pure toroidal plasma having no poloidal magnetic field. The improved toroidal plasma behavior is found to depend upon the radial currents driven by positive biasing of a ring electrode. Although biasing establishes a radial electric field at the edge, the suppression of fluctuations and density rise in the central core region of the plasma. (14 refs.)
MANSFIELD, D. K., A. T. RAMSEY, et al. (1993). "Observation of rational magnetic surfaces in TFTR using the emission from ablating deuterium pellets." Nuclear Fusion 33(1): 150-6.
The locations of the q=2 and q=1.5 magnetic surfaces have been determined to within +or-2 cm in TFTR using the emission from ablating deuterium pellets. As predicted theoretically, pronounced reductions in the average emission are found to occur when the pellet crosses these low order rational surfaces. Further, radial displacement of these surfaces is observed when the current distribution is modified. This modification has been accomplished by conditioning the graphite inner wall of TFTR so as to cause a higher effective pumping rate. The positions and radial displacements of these surfaces are predicted reasonably well by the transport analysis code TRANSP. (16 refs.)
NAJMABADI, F., R. W. CONN, et al. (1993). "Introduction and synopsis of the TITAN reversed-field-pinch fusion-reactor study." Fusion Engineering and Design 23(2/3): 69-80.
The TITAN reversed-field-pinch (RFP) fusion-reactor study has two objectives: to determine the technical feasibility and key developmental issues for an RFP fusion reactor operating at high power density; and to determine the potential economic (cost of electricity), operational (maintenance and availability), safety and environmental features of high mass-power-density fusion-reactor systems. Mass power density (MPD) is defined as the ratio of net electric output to the mass of the fusion power core (FPC). The FPC includes the plasma chamber, first wall, blanket, shield, magnets, and related structure. Two different detailed designs, TITAN-I and TITAN-II, have been produced to demonstrate the possibility of multiple engineering-design approaches to high-MPD reactors. TITAN-I is a self-cooled lithium design with a vanadium-alloy structure. TITAN-II is a self-cooled aqueous loop-in-pool design with 9-C ferritic steel as the structural material. Both designs use RFP plasmas operating with essentially the same parameters. Both conceptual reactors are based on the DT fuel cycle, have a net electric output of about 1000 MWe, are compact, and have a high MPD of 800 kWe per tonne of FPC. The inherent physical characteristics of the RFP confinement concept make possible compact fusion reactors with such a high MPD. The TITAN designs would meet the US criteria for the near-surface disposal of radioactive waste (Class C, 10CFR61) and would achieve a high Level of Safety Assurance with respect to FPC damage by decay afterheat and radioactivity release caused by accidents. Very importantly, a 'single-piece' FPC maintenance procedure has been worked out and appears feasible for both designs. Parametric system studies have been used to find cost-optimized designs, to determine the parametric design window associated with each approach, and to assess the sensitivity of the designs to a wide range of physics and engineering requirements and assumptions. (47 refs.)
ONO, M. (1993). "Ion Bernstein wave heating research." Physics of Fluids B (Plasma Physics) 5(2): 241-80.
Ion Bernstein wave heating (IBWH) utilizes the ion Bernstein wave (IBW), a hot plasma wave, to carry the radio frequency (RF) power to heat the tokamak reactor core. Earlier wave assessibility studies have shown that this finite-Larmor-radius (FLR) mode should penetrate into a hot dense reactor plasma core without significant attenuation. Moreover, the IBW's low perpendicular phase velocity ( omega /k{sub perpendicular to } approximately=V{sub Ti}<<V{sub alpha }) greatly reduces the otherwise serious wave absorption by the 3.5 MeV fusion alpha particles. In addition, the property of IBW's that K{sub perpendicular to } rho {sub i} approximately=1 makes localized bulk ion heating possible at the ion cyclotron harmonic layers. Such bulk ion heating can prove useful in optimizing fusion reactivity. In another vein, with proper selection of parameters, IBW's can be made subject to strong localized electron Landau damping near the major ion cyclotron harmonic resonance layers. This property can be useful, for example, for RF current drive in the reactor plasma core. IBW's can be excited with loop antennas or with a low-hybrid-like waveguide launcher at the plasma edge, the latter structure being one that is especially compatible with reactor application. In either case, the mode at the plasma edge is an electron plasma wave (EPW). Deeper in the plasma, the EPW is mode transformed into an IBW. Such launching and mode transformation of IBW's were first demonstrated in experiments in the Advanced Concepts Torus-1 (ACT-1) (Phys. Rev. Lett. 45, 1105 (1980)) plasma torus and in particle simulation calculations. These and other aspects of IBW heating physics have been investigated through a number of experiments performed on ACT-1, the Japanese Institute of Plasma Physics Tokamak II-Upgrade (JIPPTII-U) (Phys. Rev. Lett. 54, 2339 (1985)), the Tokyo University Non-Circular Tokamak (TNT) (Nucl. Fusion 26, 1097 (1986)), the Princeton Large Tokamak (PLT) (Phys. Rev. Lett. 60, 294 (1988)), and Alcator-C (Phys. Rev. Lett. 60, 298 (1988)). In these experiments both linear and nonlinear heating processes have been observed. Interestingly, improvement of plasma confinement was also observed in the PLT and Alcator-C experiments, opening up the possible use of IBW's for the active control of plasma transport. Two theoretical explanations have been proposed: one based on four-wave mixing of IBW with low-frequency turbulence, the other on the nonlinear generation of a velocity-shear layer. Both models are consistent with the observed threshold power level of a few hundred kW in the experiments. Experiments on lower field plasmas on JFTII-M (Eighth Topical Conference on Radio-Frequency Power in Plasmas, Irvine, CA, 1989 (American Institute of Phys., New York, 1989), p.350) and DIII-D (Eighth Topical Conference on Radio-Frequency Power in Plasmas, Irvine, CA, 1989 (American Institute of Physics, New York, 1989), p.314) have raised some concern with the IBW wave-launching process. The experiments showed serious impurity release from the walls but little or no core heating, a combination of circumstances strongly suggestive of edge heating. Possible parasitic channels could include the excitation of short wavelength modes by the Faraday shield's fringing fields, antenna-sheath-wave excitation, and axial-convective loss channel, and nonlinear processes such as parametric instability and ponderomotive effects. Suggested remedies include changes in the antenna phasing, the use of low-Z insulators, operating at higher frequencies, positioning the plasma differently with respect to the antenna, eliminating the Faraday shields, and using a waveguide launcher. The recent JIPPTII-U experiment, employing a 0- pi phased antenna array with a higher frequency 130 MHz source, demonstrated that those remedies can indeed work. Looking to the future, one seeks additional ways in which IBWH can improve tokamak performance. The strong ponderomotive potential of the IBWH antenna may be used to stabilize external kinks and, acting as an RF limiter, to control the plasma edge. Control of the plasma pressure profile with local IBWH heating is already an important part of the Princeton Beta Experiment-Modified (PBX-M) (Ninth Topical Conference on Radio-Frequency Power in Plasmas, Charleston, SC, 1991 (American Institute of Physics, New York, 1991), p.129) program in its exploration of the second-stability regime. Application of IBWH may also improve the performance of neutral beam heating and the efficiency and localization of lower-hybrid current drive for current profile control. Used with pellet injection, IBWH may also prolong the period of good confinement. The three planned high-power IBWH experiments covering vastly different parameters: f=40-80 MHz for PBX-M; f=130 MHz for JIPPT-II-U; and f=430 MHz for the Frascati Tokamak-Upgrade (FT-U) (16th European Physical Society Conference on Controlled Fusion and Plasma Physics, Venice, Italy, 1989 (European Physical Society, Amsterdam, 1989), vol.III, p.1069) appear to be well positioned to explore these possibilities and to clarify other issues including the physics of wave launching and associated nonlinear processes. (117 refs.)
PEGOURIE, B. and J. M. PICCHIOTTINO (1993). "Pellet ablation theory and experiments." Plasma Physics and Controlled Fusion
20th European Conference on Controlled Fusion and Plasma Physics 35: B157-66.
The ultimate aim of pellet ablation studies is to predict what the plasma profiles are just after a pellet injection. This requires description of the pellet ablation process, the parallel expansion of the ablatant and the fast outward motion of the deposited material since these three phenomena successively occur from the time of pellet injection to the moment when new axisymmetric profiles are reached. Only the first two points have been quantitatively modelled. If the most important processes of ablation physics are identified and although current models reproduce both measured penetrations and averaged characteristics of ablation clouds, some debatable points remain, mainly bearing on the drifts associated with the pellet motion and, consequently, on the effective shielding efficiency of the ionized part of the ablation cloud. During its parallel expansion, the ablated material experiences a strong poloidal rotation which depends on the ratio of the pellet and plasma masses and is due to the total kinetic momentum conservation on each magnetic surface. The fact that this rotation occurs on the same timescale as the outward motion suggests that both phenomena can be linked and that a comprehensive model of the whole fuelling process may emerge from considering the pellet and the plasma as a unique system. (23 refs.)
PINSKER, R. I., C. C. PETTY, et al. (1993). "Observation of parametric decay correlated with edge heating using an ion Bernstein wave antenna on DIII-D." Nuclear Fusion 33(5): 777-93.
Significant levels of parametric decay activity and correlated edge ion heating were observed during injection of high power ion Bernstein waves (IBWs) in DIII-D. Both minority hydrogen ions and majority deuterium ions showed the formation of a high energy perpendicular tail; no parallel heating was observed. The edge ion heating and the parametric decay activity were both strongest when an ion cyclotron harmonic was present at the plasma edge. Ion tail formation had a power threshold of several hundred kilowatts, above which the tail size increased with antenna power; a comparable power threshold for parametric decay instability (PDI) was observed. Both the PDI and the associated edge deuterium heating were found to be sensitive to the hydrogen-to-deuterium ratio. (31 refs.)
SCHOENBERG, K. F., R. A. GERWIN, et al. (1993). "Preliminary investigation of power flow and performance phenomena in a multimegawatt coaxial plasma thruster." IEEE Transactions on Plasma Science 21(6): 625-44.
Preliminary experimental and theoretical research that was directed toward the study of quasi-steady-state power flow in a large, unoptimized, multimegawatt coaxial plasma thruster is summarized. Large coaxial thruster operation is discussed, and the experimental results are evaluated and interpreted with a view to the development of efficient, steady-state, megawatt-class magnetoplasmadynamic (MPD) thrusters. (25 refs.)
SHEN, W. and S. C. PRAGER (1993). "The fluctuation-induced Hall effect." Physics of Fluids B (Plasma Physics) 5(7): 1931-1933.
The fluctuation induced Hall term, (J{sup approximately }*B{sup approximately }), has been measured in the Madison Symmetric Torus (MST) (R.N. Dexter et al., Fusion Technol. 19, 131 (1991)), reversed-field pinch. The term is of interest as a possible source of current self-generation (dynamo). It is found to be nonnegligible, but small in that it can account for less than 25% of the dynamo driven current. (9 refs.)
SHINOHARA, S. (1993). "Preliminary experiments with edge current injection in a reversed field pinch." Plasma Physics and Controlled Fusion 35(12): 1661-7.
The first dc helicity injection is tried applying a voltage between two parallel plates, and effects on plasma performance are studied in the reversed field pinch plasma. The injected current is nearly proportional to the applied voltage V{sub H} with a small offset-induced voltage ( approximately -30 V), and a transient increase in the toroidal flux and a decrease in the poloidal field (plasma takes the more relaxed state) is found near the injection port, regardless of the polarity of V{sub H}. The induced voltage measured as a function of the insertion depth shows the presence of high-energy electrons. (21 refs.)
SLOUGH, J. T. and A. L. HOFFMAN (1993). "Stability of field-reversed configurations in the large s experiment (LSX)." Physics of Fluids B (Plasma Physics) 5(12): 4366-77.
Data from several diagnostics employed on the large s Experiment (LSX) field-reversed theta pinch were analyzed to seek correlation between plasma distortions and the confinement properties of the field-reversed configurations (FRC's) formed. In particular, an array of B{sub theta } probes was used to determine separatrix movement, which might indicate the existence of low-order modes, such as a tilt instability. No correlation between the quality of confinement and signal was observed. The parameter s, equal to the average number of ion gyroradii inside the separatrix, has been postulated as a measure of FRC stability with values above 2, leading to instability and loss of confinement. However, the confinement observed in experiments conducted over a large range of s (1<s<8) appeared to correlate more with the shape of the equilibrium radial density profile produced during formation rather than s. Flatter profiles correlated with poorer confinement. (23 refs.)
TSUI, H. Y. W., A. J. WOOTTON, et al. (1993). "A comparison of edge turbulence in tokamaks, stellarators, and reversed-field pinches." Physics of Fluids B (Plasma Physics) 5(7): 2491-2497.
Edge equilibrium, turbulence and transport related parameters from the Advanced Toroidal Facility (ATF) (Fusion Technol, 10, 179 (1986)) torsatron, the ZT-40M (Fusion Technol. 8, 1571 (1985)) reversed-field pinch, the Phaedrus-T (Nucl. Fusion 32, 2040 (1992)) tokamak, the Texas Experimental Tokamak (TEXT) (Nucl. Technol. Fusion 1, 479 (1981)), and the Tokamak Fusion Test Reactor (TFTR) (in Plasma and Controlled Nuclear Fusion Research, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol.1, p.9) have been obtained using a standardized Langmuir probe array and a consistent set of data analysis packages. Additional data from some other devices have also been furnished via private communications and incorporated from published results. Experimental results over a wide range of parameters are compared and the turbulence contribution to edge transport are assessed. Certain physical properties that are relevant to the modeling of edge turbulence are identified: namely, shear decorrelation of turbulence, the role of resistive dissipation and electron parallel thermal conduction, radial mode structure in sheared magnetic field, and electromagnetic contribution to the parallel Ohm's law. (47 refs.)
WARE, A. A. and J. C. WILEY (1993). "The electrostatic wake of a superthermal test electron in a magnetized plasma." Physics of Fluids B (Plasma Physics) 5(8): 2764-8.
The electrostatic potential is determined for a test electron with v{sub }//>>v{sub Te} in a uniform magnetized plasma ( omega {sub ce}>> omega {sub pe}). In the frame of the test electron, part of the spatially oscillatory potential has spherical symmetry over the hemisphere to the rear of the electron and is zero ahead of the electron. A second part of different character, which makes the potential continuous at the plane containing the electron, is oscillatory in the radial direction but decreases almost monotonically in the axial direction. (9 refs.)
YASAKA, Y., H. TAKENO, et al. (1993). "Nonlinear ion Bernstein wave heating experiment in the WT-3 tokamak." Plasma Physics and Controlled Fusion 35(3): 379-89.
Ion Bernstein wave (IBW) heating at ion cyclotron half-harmonic frequencies was performed in hydrogen (H) discharges in the WT-3 tokamak. The bulk ion temperature increase was measured by a charge exchange analyser and a spatially scanning polychrometer. The latter diagnostics can provide radial ion-temperature profiles. The incremental ion temperature has shown a peak at the exact location where the IBW frequency omega equals 3/2 of the ion cyclotron frequency, indicating the nonlinear direct acceleration of H ions. The wave electric field was measured by a Langmuir probe in the scrape-off layer. The electric field at the second harmonic of the excited IBW frequency was observed to vary depending on the power level and the toroidal magnetic field. These experimental results are compared with the theoretical prediction of nonlinear IBW damping. The measured dependences of the ion heating and the wave amplitude at 2 omega on the toroidal magnetic field are in reasonable agreement with the prediction. (16 refs.)
YOSHIDA, Z., A. HASEGAWA, et al. (1993). "Production of superthermal electrons and ion cyclotron waves in a reversed-field-pinch plasma." Physics of Fluids B (Plasma Physics) 5(9): 3261-6.
Production of superthermal electrons through kinetic interactions with electromagnetic fluctuations is studied to account for observations of fast electrons and ion cyclotron waves in reversed field pinch plasmas. Low-frequency Alfvenic (torsional) modes can interact with electrons through the Landau resonance when the wavelength perpendicular to the magnetic field is as small as the ion gyroradius. Such kinetic Alfven waves induce simultaneous diffusion (double diffusion) in the coordinate and velocity spaces and produce a field aligned superthermal electron beam in the edge region. Microinstabilities are driven by the electron-beam and ion cyclotron waves are excited. Through these processes the fluctuation energy in the low-frequency regime may be transported to the ion cyclotron frequency regime. (19 refs.)
BELLAN, P. M. (1994). "Alfven 'resonance' reconsidered: exact equations for wave propagation across a cold inhomogeneous plasma." Physics of Plasmas 1(11): 3523-41.
Previous discussions of Alfven wave propagation across an inhomogeneous plasma predicted that shear Alfven waves become singular (resonant) at the omega =k{sub z} upsilon {sub A} layer and that there is a strong wave absorption at this layer giving localized ion heating. In this paper the three standard derivations of the Alfven 'resonance' (incompressible magnetohydrodynamics, compressible magnetohydrodynamics, and two-fluid) are re-examined and shown to have errors and be mutually inconsistent. Exact two-fluid differential equations for waves propagating across a cold inhomogeneous plasma are derived; these show that waves in an ideal cold plasma do not become 'resonant' at the Alfven layer so that there is no wave absorption or localized heating. These equations also show that the real 'shear' Alfven wave differs in substance from both the ideal MHD and earlier two-fluid predictions and, in the low density, high field region away from the omega =k{sub z} upsilon {sub A} layer, is actually a quasielectrostatic resonance cone mode. For omega << omega {sub ci} and k{sub y}=0, the omega =k{sub z} upsilon {sub A} layer turns out to be a cutoff (reflecting) layer for both the 'shear' and compressional modes (and not a resonance layer). For finite omega / omega {sub ci} and k{sub y}=0 this layer becomes a region of wave inaccessibility. For omega << omega {sub ci} and finite k{sub y} there is strong coupling between shear and compressional modes, but still no resonance. (63 refs.)
BLACK, D. C., R. M. MAYO, et al. (1994). "2-DIMENSIONAL MAGNETIC-FIELD EVOLUTION MEASUREMENTS AND PLASMA-FLOW SPEED ESTIMATES FROM THE COAXIAL THRUSTER EXPERIMENT." Physics of Plasmas 1(9): 3115-3131.
Local, time-dependent magnetic field measurements have been made in the Los Alamos coaxial thruster experiment (CTX) [C. W. Barnes et al., Phys. Fluids B 2, 1871 (1990); J. C. Fernandez et al, Nucl. Fusion 28, 1555 (1988)] using a 24 coil magnetic probe array (eight spatial positions, three axis probes). The CTX is a magnetized, coaxial plasma gun presently being used to investigate the viability of high pulsed power plasma thrusters for advanced electric propulsion. Previous efforts on this device have indicated that high pulsed power plasma guns are attractive candidates for advanced propulsion that employ ideal magnetohydrodynamic (MHD) plasma stream flow through self-formed magnetic nozzles. Indirect evidence of magnetic nozzle formation was obtained from plasma gun performance and measurements of directed axial velocities up to upsilon(z) approximately 10(7) cm/s. The purpose of this work is to make direct measurement of the time evolving magnetic field topology. The intent is to both identify that applied magnetic field distortion by the highly conductive plasma is occurring, and to provide insight into the details of discharge evolution. Data from a magnetic fluctuation probe array have been used to investigate the details of applied magnetic field deformation through the reconstruction of time-dependent flux profiles. Experimentally observed magnetic field line distortion has been compared to that predicted by a simple one-dimensional (1-D) model of the discharge channel. Such a comparison is utilized to estimate the axial plasma velocity in the thruster. Velocities determined in this manner are in approximate agreement with the predicted self-field magnetosonic speed and those measured by a time-of-flight spectrometer.
Dubois, M. A., P. Ghendrih, et al. (1994). "Does magnetic turbulence explain anomalous electron energy transport?." Plasma Physics and Controlled Fusion
21st European Conference on Controlled Fusion and Plasma Physics 36(12B): B55-64.
The question of anomalous transport of the energy of the electrons has been puzzling tokamak physicists for nearly as long as tokamaks exist (and internal disruptions is the other riddle which has reached its twentieth anniversary...). Magnetic turbulence has been suggested as a possible answer for a long time, but linear theory gave negative results, and experimental evidence could not be obtained with existing diagnostics until recently. In the past few years, much progress has been made, and although no final conclusion can be reached yet, it is useful to regroup and compare the latest theoretical and experimental results. As recent and exhaustive reviews already exist, we will not try to cover all the existing literature, but rather adopt a more personal and maybe biased point of view. In the first part, we discuss the possible theoretical approaches, and point out the directions along which progress should be made, especially the need to obtain a dynamical description of the magnetic structure, taking into account the feedback which the induced transport has on the instabilities. In the second part, we review and discuss the experimental evidence in favour of magnetic turbulence, and discuss the relevance of the observations to the explanation of the measured transport coefficients. Although we try to cover most of the available data, many illustrations will be taken from the Tore Supra tokamak, which is well endowed with diagnostics of the fine scale magnetic structure. (60 refs.)
EJIRI, A., S. OHDACHI, et al. (1994). "Magnetic and velocity fluctuation measurements in the REPUTE-1 reversed-field pinch plasma." Physics of Plasmas 1(5): 1177-1185.
Magnetic and velocity fluctuations are studied in the REPUTE-1 (Plasma Phys. Controlled Fusion 28, 805 (1986)) reversed-field pinch (RFP). The first measurement of velocity fluctuation in an RFP plasma has been done using a Doppler shift of the O v(O{sup 4+}, 278.1 nm) line. The fluctuation level increases as the radius of the viewing chord increases. Magnetic fluctuation measurements by an insertable probe reveal that the radial cross correlation of toroidal field fluctuation changes its sign at the radius slightly inside the reversal surface. The level of magnetohydrodynamic dynamo term is estimated from magnetic fluctuations at the surface correlation changes and oxygen velocity fluctuations measured with the chord distance of 115 mm. The dynamo term and that due to resistivity are the same level. This fact is consistent with Ohm's law on which magnetohydrodynamic dynamo models are based. (39 refs.)
FALL, T., B. TERREAULT, et al. (1994). "Novel isotope exchange scenarios: investigation of particle recycling in the TdeV tokamak." Plasma Physics and Controlled Fusion 36(11): 1763-74.
The isotope ratio R{sub I}=D/(H+D) was measured during ohmic discharges in the magnetically diverted tokamak TdeV, using spectroscopy (H{sub alpha } and D{sub alpha }), gas analysis (H{sub 2}, HD and D{sub 2}), mass analysis of charge exchange neutrals, and detection of H and D in a collector probe, in the course of four different experimental phases: (I) 'dynamical' isotope switchover of the fuelling gas during a discharge, (II) shot-to-shot evolution with pure D{sub 2} fuelling, (III) same with walls conditioned by D{sub 2} glow discharge, and (IV) 'dynamical' switchover combined with divertor biasing and pumping. The different values obtained for R{sub I} are compared and analysed in terms of the particle recycling properties of the device. The dynamical scenario (phase I) revealed a lack of equilibrium between the edge and core isotopic composition that is an indication of the degree to which the particle flux is amplified by recycling in the vicinity of surfaces. The phase II results were fitted to a model which yielded an estimate of the magnitude of the wall inventory available for recycling. In phase III, the effectiveness of conditioning was assessed. Phase IV revealed an even more pronounced disequilibrium than phase I, and demonstrated the power of divertor biasing and pumping in exhausting the wall inventory and controlling the isotope ratio. (24 refs.)
FIKSEL, G., S. C. PRAGER, et al. (1994). "Measurement of magnetic fluctuation induced energy transport." Physical Review Letters 72(7): 1028-31.
The local electron energy flux produced by magnetic fluctuations has been measured directly in the MST reversed field pinch (over the radial range r/a>0.75). The flux, produced by electrons traveling parallel to a fluctuating magnetic field, is obtained from correlation between the fluctuations in the parallel heat flux and the radial magnetic field. The fluctuation induced flux is large (100 kW/cm{sup 2}) in the 'core' (r/a<0.85) and small (<10-30 kW/cm{sup 2}) in the edge. (14 refs.)
GAUSTER, W. B. (1994). "THE IMPACT OF MATERIALS SELECTION ON THE DESIGN OF THE INTERNATIONAL THERMONUCLEAR EXPERIMENTAL REACTOR (ITER)." Journal of Nuclear Materials
6th International Conference on Fusion Reactor Materials (ICFRM-6) 215: 3-10.
The design being developed for the International Thermonuclear Experimental Reactor (ITER) in the Engineering Design Activities (EDA) phase has to assure the achievement of controlled ignition and extended burn, and must provide a demonstration of technologies that are essential to a reactor. These requirements are being addressed by the application of present physical understanding of tokamaks, and by emphasis on simplicity in design and in maintenance concepts. The approach places severe demands on materials used in the design, and the concept developed so far requires earlier qualification of materials to more stringent specifications than was expected previously in the fusion program.
HAWRYLUK, R. J., H. ADLER, et al. (1994). "Confinement and heating of a deuterium-tritium plasma." Physical Review Letters 72(22): 3530-3.
The Tokamak Fusion Test Reactor has performed initial high-power experiments with the plasma fueled with nominally equal densities of deuterium and tritium. Compared to pure deuterium plasmas, the energy stored in the electron and ions increased by approximately 20%. These increases indicate improvements in confinement associated with the use of tritium and possibly heating of electrons by a particles created by the D-T fusion reactions. (13 refs.)
HEIDBRINK, W. W. and G. J. SADLER (1994). "The behaviour of fast ions in tokamak experiments." Nuclear Fusion 34(4): 535-615.
Fast ions with energies significantly larger than the bulk ion temperature are used to heat most tokamak plasmas. Fast ion populations created by fusion reactions, by neutral beam injection and by radiofrequency (RF) heating are usually concentrated in the centre of the plasma. The velocity distribution of these fast ion populations is determined primarily by Coulomb scattering; during wave heating, perpendicular acceleration by the RF waves is also important. Transport of fast ions is typically much slower than thermal transport, except during MHD events. Intense fast ion populations drive collective instabilities. Implications for the behaviour of alpha particles in future devices are discussed. (440 refs.)
HOSEA, J., J. H. ADLER, et al. (1994). "Deuterium-tritium experiments on the Tokamak Fusion Test Reactor." Fusion Technology
11th Topical Meeting on the Technology of Fusion Energy 26(3): 389-398.
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 fully explored. (37 refs.)
JI, H., A. F. ALMAGRI, et al. (1994). "Time-resolved observation of discrete and continuous magnetohydrodynamic dynamo in the reversed-field pinch edge." Physical Review Letters 73(5): 668-71.
We report the first experimental verification of the magnetohydrodynamic (MHD) dynamo in the reversed-field pinch (RFP). A burst of MHD dynamo electric field is observed during the sawtooth crash, followed by an increase in the local parallel current in the Madison Symmetric Torus RFP edge. By measuring each term, the parallel MHD mean-field Ohm's law is observed to hold within experimental error bars between and during sawtooth/crashes. (16 refs.)
KAMADA, Y., K. USHIGUSA, et al. (1994). "NONINDUCTIVELY CURRENT-DRIVEN H-MODE WITH HIGH BETA(N)-VALUES AND HIGH BETA(P)-VALUES IN JT-60U." Nuclear Fusion 34(12): 1605-1618.
Attainable beta(p) and beta(N) values were widely improved in JT-60U by peaking of the current profile and broadening of the pressure profile. In a quasi-steady state of ELMy H mode, beta(p) similar to 2.5-3, beta(N) similar to 2.5-3.1 and H factor similar to 1.8-2.2 were sustained for similar to 1 s simultaneously under the full current drive condition (bootstrap current 60%, beam driven current 48% at I-p = 0.5 MA). In a transient case, the maximum values of beta(p), epsilon beta(p) and beta(N) reached 4.7, 1.2 and 4.2, respectively. The achievable beta(p) value increases systematically with safety factor at the edge. To obtain these improved high beta plasmas, it is essential to control MHD activities; suppression of beta(p) collapse and medium m/n (poloidal mode number/toroidal mode number) modes by the modification in current and pressure profiles and control of ELM activity by suitable selection of the heating power.
KUBOTA, S., M. NAGATSU, et al. (1994). "Far-infrared laser interferometry measurements on the STP-3(M) reversed-field pinch." Japanese Journal of Applied Physics, Part 1 (Regular Papers & Short Notes) 33(4A): 2050-5.
Far-infrared laser interferometry at 432 mu m was carried out on the STP-3(M) reversed-field pinch. Measurements along two vertical chords showed a change from a parabolic-like to a flat-like electron density profile after field reversal. A density profile inversion and a correlated toroidal magnetic flux perturbation were also observed during the transition from the current rising to the current decay phase. Measurements of electron density fluctuations indicated relative fluctuation levels of approximately 10% for both chords during the current rising phase and approximately 5% and approximately 15% during the current decay phase for the central and outer chords, respectively. Spectral analysis showed a approximately 30 kHz mode consistent with poloidal mode number m=0 magnetic fluctuations, and a approximately 90 kHz mode localized to the outer region of the plasma, which was strongly excited during the current decay phase and may be connected to particle and energy transport in STP-3(M). (12 refs.)
MATSUOKA, A., Y. KONDOH, et al. (1994). "A Monte Carlo simulation of Langmuir probe methods for plasmas having RF amplitude oscillations." Japanese Journal of Applied Physics, Part 1 (Regular Papers & Short Notes) 33(5A): 2735-40.
A Monte Carlo simulation model to analyze Langmuir probe characteristics is developed for an ideal plane probe in plasmas having RF oscillations of space potential only. When the oscillation amplitude is larger than electron temperature, plasma parameters determined with this model agree with those predicted by the simple probe theory with space potential oscillations. Space potential shifts and deformed probe characteristics for an RF discharge are explained by profiles of space/time electron density, energy and potential gained from these probe simulations with RF space potential oscillations. (20 refs.)
MAYO, R. M., M. A. BOURHAM, et al. (1994). "Initial results on high enthalpy plasma generation in a magnetized coaxial source." Fusion Technology
11th Topical Meeting on the Technology of Fusion Energy 26(3): 1221-1225.
Initial investigation on high enthalpy plasma stream generation in the North Carolina State University Coaxial Plasma Source (1) (CPS) facility is presented. Tenuous, yet high enthalpy, flows are produced from this Magnetized Coaxial plasma Gun (MCG) which allow laboratory study of plasma streams with a wide variety of applications. The applicability includes, but is not limited to, advanced thrusters for electric space propulsion, astrophysical jets and critical ionization phenomena, magnetic fusion compact toroid devices and tokamak fueling, large scale plasma etching and deposition, etc. (13 refs.)
MAZUR, S. (1994). "Toroidal phase locking phenomena and nonaxisymmetric flux perturbations in the Extrap T1 reversed-field pinch." Physics of Plasmas 1(10): 3356-64.
The structure of the magnetic fluctuations in the Extrap T1 reversed-field pinch (Phys. Scr. 44, 358 (1991)) has been studied over the pinch parameter range 1.6< Theta =B{sub theta }(a)/(B{sub phi })<2.2. It is found that nonaxisymmetric perturbations of the toroidal flux and edge field dominate the global dynamics at all Theta . These perturbations are formed by phase locking of the dominant m=1 modes and m=0 modes due to strong nonlinear interaction. This m=1/m=0 mode locking rotates toroidally in the electron drift direction, with an angular rotation velocity of typically 3.3*10{sup 4} rad/s. The toroidal mode numbers of the m=0 modes are consistent with being nonlinearly driven by the m=1 modes. Evidence is found that the behavior of the leading m=1 mode governs the degree of phase locking. The role of this phenomenon in mean field maintenance is discussed. (34 refs.)
MAZUR, S. (1994). "MAGNETIC-FIELD FLUCTUATIONS IN THE HIGH-ASPECT-RATIO EXTRAP T1 REVERSED-FIELD PINCH." Physica Scripta 49(2): 233-238.
Magnetic field fluctuations have been investigated on the high aspect ratio (R/a = 8.8) Extrap TI reversed-field pinch. A global mode structure is found, characterised by poloidal mode numbers m = 0 and m = 1. The mode spectrum of m = 1 fluctuations is peaked around toroidal mode numbers n = - 19, ..., - 14, corresponding to modes with resonances on the q-profile inside the reversal surface. The mode amplitudes are such that magnetic island overlapping should occur in the central part of the plasma causing stochastisation of the magnetic field. Nearly rigid body rotation of the internally resonant modes is observed, with poloidal and toroidal angular velocities of omega(theta), = 1.4 x 10(4) +/- 0.2 x 10(5) rad/s and omega(phi) = 0.5 x 10(4) +/- 0.1 x 10(4) rad/s, respectively.
MORI, M., S. ISHIDA, et al. (1994). "Achievement of high fusion triple product in the JT-60U high beta {sub p} H mode." Nuclear Fusion 34(7): 1045-53.
Improvement of an enhanced confinement state in a high poloidal beta ( beta {sub p}) regime without sawtooth activity has been achieved in JT-60U. A confinement mode has been demonstrated where both the edge and the core confinement are improved. The attainable beta {sub p} was also extended to higher values in this improved mode, because of its broader pressure profile. As a result of the improvement in confinement and in attainable beta {sub p}, the highest value of the fusion triple product has been extended by a factor of 2.5 over that achieved in the 1992 experiments; it has reached (1.1+or-0.3)*10{sup 21} m{sup -3}.s.keV with a central ion temperature of about 37 keV. The D-D neutron emission rate has also been doubled in these experiments and has reached (5.6+or-0.6)*10{sup 16} s{sup -1}. (20 refs.)
MORI, A., M. IIDA, et al. (1994). "Fast electron behavior in comparison with dynamo models in a reversed-field pinch." Journal of the Physical Society of Japan 63(10): 3548-51.
On the basis of soft-X ray diagnostics, it is shown that the role of the edge fast electrons in the reversed-field pinch (RFP) dynamo mechanism is insignificant in the early phase of a discharge in the STE-2 RFP. In the high-fill-pressure region, this is also the case even with the current maximum phase. The pressure dependence of the edge magnetic fluctuations is possibly different from the expected dominance of the MHD dynamo activity in the high-pressure regime, indicating the importance of direct measurement of the phase relation between velocity and magnetic field fluctuations. (14 refs.)
NAGATA, A., K. I. SATO, et al. (1994). "A global dynamo in a reversed-field pinch plasma." Physics of Plasmas 1(9): 2876-81.
It is found that a reversed-field pinch (RFP) dynamo uniformly generates most of the toroidal flux in a RFP plasma with a high magnetic Reynolds (S) number. Hereafter, this is referred to as the global dynamo. The toroidal flux generated by the global dynamo is proportional to the plasma current. The effective inductance in a RFP plasma increases with S number and remains constant against the changing plasma current. This means that the pinch parameter Theta is held constant, that is, the global dynamo has a self-organizing effect sustaining a particular RFP configuration. By comparing simulation results and analysis of magnetic fluctuations, it is confirmed that a global dynamo is generated by the nonlinear evolution of the single-helicity of the m=1 mode alone. (24 refs.)
NORDLUND, P. (1994). "QUASI-LINEAR INTERACTION OF M=1 AND M=0 MODAL COMPONENTS WITH THE MEAN FIELDS IN THE EXTRAP-T1 REVERSED-FIELD PINCH." Physics of Plasmas 1(9): 2945-2955.
Experiments have been performed on the Extrap T1 reversed-field pinch [Phys. Scr. 44, 358 (1991)] where the detailed radial structure of the mean profiles, together with associated edge field fluctuation spectra, have been studied at different values of the pinch parameter 1.6 less-than-or-equal-to THETA less-than-or-equal-to 2.5. It is found that the parallel current density is strongly suppressed inside the resonances of the most active m = 1 modes, resulting in hollow mu profiles near the axis. This effect introduces a spectral shift in n space between the dominant m = 1 modes and the leading linear growth rates. At high THETA a local flattening of mu around the reversal (m =0 resonance) surface becomes apparent. As a result a second set of m = 1 modes resonant near the off-axis peak in mu becomes linearly unstable. In this case, the resulting m = 1 spectrum is double peaked for the internal modes. The experimental results show several characteristic similarities with recent numerical magnetohydrodynamic simulations of the reversed-field pinch dynamo.
PARK, H. K., M. G. BELL, et al. (1994). "Role of neutral-beam fuelling profile in energy confinement and neutron emission on TFTR." Nuclear Fusion 34(9): 1271-6.
The role of the neutral-beam fuelling profile in determining the stored energy and neutron emission in neutral-beam-heated discharges on TFTR is investigated. It is found that the neutral-beam fuelling profile is sensitive to both the magnitude and the shape of the electron density. The ion stored energy correlates well with the peaking of the neutral-beam fuelling profile, whereas the electron stored energy is insensitive to this parameter. The DD fusion neutron emission is also well correlated with the peaking of the neutral-beam fuelling profile. The derived empirical scalings of the ion stored energy and neutron yield include discharges from the L mode to the supershot regime. (20 refs.)
PARKS, P. B., G. A. GERDIN, et al. (1994). "Model of ablation flow near light-atom pellets with surface boundary conditions." Nuclear Fusion 34(3): 417-26.
A theory is formulated to describe the dynamics of the high density, partially ionized region of the ablation cloud in the vicinity of a spherical light-atom refractory pellet exposed to a tokamak plasma. Owing to the finite sublimation energy of refractory pellets, the surface boundary conditions require a more detailed treatment than in the case of frozen hydrogen. With the exception of conditions that lead to relatively cold ablation clouds (low plasma density, or small pellet radius), transonic flow solutions were found to exist for most parameters of interest. The shielding properties of the ablation cloud, degree of ionization at the M=1 surface, and the surface flow Mach number were found to be in fairly good agreement with the heuristic model of Parks, Leffler and Fisher. (29 refs.)
RICCARDI, C., M. FONTANESI, et al. (1994). "PLASMA DIAGNOSTIC THROUGH ION BERNSTEIN WAVES." Nuovo Cimento Della Societa Italiana Di Fisica D-Condensed Matter Atomic Molecular and Chemical Physics Fluids Plasmas Biophysics 16(5): 505-515.
The effectiveness of a diagnostic method for both ion temperature and ion minority concentrations, based on propagation of Ion Bernstein Waves (IBW), is analysed theoretically and verified experimentally. The method is based on a comparison between wave vector measurements, obtained with an interferometric system, and theoretical dispersion relations.
RICCARDI, C., M. FONTANESI, et al. (1994). "Experimental investigation on ion-Bernstein waves propagation." Plasma Physics and Controlled Fusion 36(11): 1791-803.
The IBW propagation is experimentally investigated in a steady-state toroidal plasma. IBW is indirectly excited by EPW conversion in lower hybrid resonance layer. The phenomenon is analysed in linear conditions, with respect to the relevant plasma parameters, as frequency and density profiles. (17 refs.)
Rostagni, G. (1994). "Confinement results in RFPs." Plasma Physics and Controlled Fusion
21st European Conference on Controlled Fusion and Plasma Physics 36(12B): B203-12.
A short presentation of the main characteristics and peculiar aspects of the RFP configuration is given. Confinement data from a number of experiments, which have been operated after 1980, are recalled; the main emphasis is on results from MST and RFX, the two largest existing devices, RFX, with a current capability of 2 MA, in the first two years has been operated at low currents (350+700) kA, mainly to compare different modes of creating the RFP configuration, to study field error effects and plasma wall interaction, to improve and optimize, in particular by current and field control, the discharge parameters. The reported results are in line with expectations. (48 refs.)
RYPDAL, K., E. GRONVOLL, et al. (1994). "Confinement and turbulent transport in a plasma torus with no rotational transform." Plasma Physics and Controlled Fusion 36(7): 1099-114.
In the BLAAMANN device a weakly ionized hydrogen plasma is produced by electrons accelerated from a hot, negatively biased tungsten filament and confined in a toroidal magnetic field of strength up to 0.4 T. The plasma is turbulent, with relative fluctuation levels in n{sub e}, phi and T{sub e} of 10% or more. The time-averaged state exhibits nested toroidal surfaces of constant potential and pressure, which requires an anomalous cross-field current to remove the space-charge injected by the cathode and the charge accumulated due to the Del B- and curvature drifts. Typical plasma parameters are n{sub e} approximately 10{sup 16} m{sup -3}, T{sub e} approximately 1-20 eV, T{sub i} approximately 1 eV. The cross-field diffusion coefficient is typically D{sub perpendicular to } approximately 30 m{sup 2} s{sup $} -{sup 1} approximately 10{sup 4}*D{sub perpendicular to }{sup classical} approximately 10{sup 1}*D{sub perpendicular to }{sup Bohm}. Evidence is presented in support of the hypothesis that the plasma goes turbulent because it needs to develop an anomalous current channel, and this turbulence in turn determines the plasma transport and the time-averaged state. (11 refs.)
SAILOR, W. C., C. W. BARNES, et al. (1994). "Conceptual design for a fast neutron ionization chamber for fusion reactor plasma diagnostics." Fusion Technology
11th Topical Meeting on the Technology of Fusion Energy 26(3): 945-948.
A conceptual design for a radiation-hard "pointing" fast neutron ionization chamber that is capable of delivering a 1 MHz countrate of T(D, n) events at ITER is given. The detector will use a approximately 1 cm{sup 3} volume of CO{sub 2} fill gas at 0.1 bar pressure in a approximately 500 V/cm electric field. The pulse widths will be approximately 10 ns, enabling it to operate in a flux of approximately 6*10{sup 13} DT n/cm{sup 2}/sec. A special collimator design is used, giving an estimated angular resolution of 4.5 degrees HWHM. (15 refs.)
SHARMA, R. P., A. KUMAR, et al. (1994). "Excitation of electron Bernstein and ion Bernstein waves by extraordinary electromagnetic pump: kinetic theory." Physics of Plasmas 1(3): 522-7.
In the present paper, parametric decay instability of an extraordinary electromagnetic (EM) wave into a high-frequency electron Bernstein wave and a low-frequency ion Bernstein wave is studied. The guiding center formalism is followed to obtain the coupling coefficients including kinetic effects. The momentum space integrals in the coupling coefficients are solved in the dipole approximation of the pump wave and in the limits, k{sub perpendicular to } rho {sub e}<1 for electron Bernstein wave and k{sub perpendicular to } rho {sub i}<1 for ion Bernstein wave. Analytical expressions for homogeneous growth rate and threshold are given explicitly. Applications of the present investigation are pointed out to ionospheric modification experiment and fusion plasmas. (11 refs.)
SHEFFIELD, J. and J. D. GALAMBOS (1994). "Prospects for toroidal fusion reactors." Fusion Technology
11th Topical Meeting on the Technology of Fusion Energy 26(3): 1122-1126.
Work on the International Thermonuclear Experimental Reactor (ITER) tokamak has refined understanding of the realities of a deuterium-tritium (D-T) burning magnetic fusion reactor. An ITER-like tokamak reactor using ITER costs and performance would lead to a cost of electricity (COE) of about 130 mills/kWh. Advanced tokamak physics to be tested in the Toroidal Physics Experiment (TPX), coupled with moderate extrapolation in engineering, technology, and unit costs (i.e., based on the ITER design), should lead to a COE comparable with best existing fission systems around 60 mills/kWh. However, a larger unit size, approximately 2000 MW{sub (e}), is favored for the fusion system. Alternative toroidal configurations to the conventional tokamak, such as the stellarator, reversed-field pinch, and field-reversed configuration, offer some potential advantage, but are less well developed, and have their own challenges. (13 refs.)
SHEFFIELD, J. (1994). "The physics of magnetic fusion reactors." Reviews of Modern Physics 66(3): 1015-103.
During the past two decades there have been substantial advances in magnetic fusion research. On the experimental front, progress has been led by the mainline tokamaks, which have achieved reactor-level values of temperature and plasma pressure. Comparable progress, when allowance is made for their smaller programs, has been made in complementary configurations such as the stellarator, reversed-field pinch and field-reversed configuration. The status of understanding of the physics of toroidal plasmas is reviewed. It is shown how the physics performance, constrained by technological and economic realities, determines the form of reference toroidal reactors. A comparative study of example reactors is not made, because the level of confidence in projections of their performance varies widely, reflecting the vastly different levels of support which each has received. Success with the tokamak has led to the initiation of the International Thermonuclear Experimental Reactor project. It is designed to produce 1500 MW of fusion power from a deuterium-tritium plasma for pulses of 1000 s or longer and to demonstrate the integration of the plasma and nuclear technologies needed for a demonstration reactor. (521 refs.)
SHIMADA, T., Y. YAGI, et al. (1994). "Large-amplitude oscillations of soft X-rays in a high current density plasma on TPE-1RM15 reversed field pinch." Plasma Physics and Controlled Fusion 36(3): 561-72.
In the high current density plasma on TPE-1RM15 reversed field pinch, large-amplitude oscillations with m=0, n=0 are observed in soft X-ray (SXR) emission for a wide range of theta values, even at low- theta ( theta <1.55) discharges. Another small-amplitude oscillation with m=1 is observed quasicontinuously for almost all the discharge. The burst of high-energy soft X-rays is observed in the channels of the SBD array looking at the electron drift side of a metal limiter and seems to correlate with the crash of the large-amplitude oscillation. This SXR burst indicates the existence of energetic electrons flowing along the field lines near the plasma surface in the dynamo phase. These large-amplitude SXR oscillations are mainly due to the variations in the electron temperature and are roughly in antiphase with the oscillation of ion temperature. (27 refs.)
Strachan, J. D., H. Adler, et al. (1994). "Deuterium and tritium experiments on TFTR." Plasma Physics and Controlled Fusion
21st European Conference on Controlled Fusion and Plasma Physics 36(12B): B3-15.
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. (8 refs.)
STRACHAN, J. D., H. ADLER, et al. (1994). "Fusion power production from TFTR plasmas fueled with deuterium and tritium." Physical Review Letters 72(22): 3526-9.
Peak fusion power production of 6.2+or-0.4 MW has been achieved in TFTR plasmas heated by deuterium and tritium neutral beams at a total power of 29.5 MW. These plasmas have an inferred central fusion alpha particle density of 1.2*10{sup 17} m{sup -3} without the appearance of either disruptive magnetohydrodynamics events or detectable changes in Alfven wave activity. The measured loss rate of energetic alpha particles agreed with the approximately 5% losses expected from alpha particles which are born on unconfined orbits. (22 refs.)
STRAIT, E. J. (1994). "Stability of high beta tokamak plasmas." Physics of Plasmas 1(5): 1415-1431.
Stability at high beta (the ratio of plasma pressure to magnetic field pressure) is an important requirement for a compact, economically attractive fusion reactor. It is also important in present large tokamak experiments, where the best performance is now often limited by instabilities rather than by energy transport. The past decade has seen major advances in our understanding of the stability of high beta tokamak plasmas, as well as in the achievement of high values of beta. Ideal magnetohydrodynamic (MHD) theory has been remarkably successful in predicting the stability limits, and the scaling of maximum stable beta with the normalized plasma current predicted by Troyon and others has been confirmed in many experiments, yielding a limit beta {sub max} approximately=3.5 (%-m-T/MA) I/aB (where I is the plasma current, a is the minor radius, and B is the toroidal field). The instabilities which are predicted to limit beta have been observed experimentally, in good agreement with theoretical predictions, including long-wavelength kink modes and short-wavelength ballooning instabilities. Advances in understanding of tokamak stability have opened several paths to higher values of beta. The use of strong discharge shaping, approaching the limits of axisymmetric stability, has allowed beta values as high as 12% to be reached in agreement with Troyon scaling. Recent experimental results and ideal MHD modeling have shown that the beta limit depends on the form of the pressure and current density profiles, and modification of the current density to create a centrally peaked profile has allowed beta values up to 6I/aB to be achieved experimentally. Recent experiments have also begun to explore both local and global access to the predicted second stable regime for ballooning modes, with the potential for very high values of beta /(I/aB). Preliminary experimental investigations of wall stabilization and radio-frequency (RF) current profile control hold the promise of further improvements in beta through passive and active control of instabilities. The developing understanding of high beta stability and the application of this understanding to present experiments and future fusion devices hold the potential for production of stable, steady state plasmas at high beta with good confinement. (167 refs.)
Anderson, J. L. and P. LaMarche (1995). "Tritium activities in the United States." Fusion Technology
5th Topical Meeting on Tritium Technology in Fission, Fusion, and Isotopic Applications 28(3): 479-490.
There have been many significant changes in the status of tritium activities in the US since the 4th Tritium Conference in October, 1991. The Replacement Tritium Facility (RTF) at Savannah River site and the Weapons Engineering Tritium Facility (WETF) at the Los Alamos National Laboratory are now operational with tritium. The Tokamak Fusion Test Reactor (TFTR) has initiated a highly successful experimental campaign studying DT plasmas, and has produced more than 10 megawatts (MW) of fusion power in a DT plasma. Sandia National Laboratory has ceased tritium operations at the Tritium Research Laboratory (TRL) and many of the activities previously performed there have been transferred to Los Alamos and Savannah River. The tritium laboratory at Lawrence Livermore National Laboratory has reduced the tritium inventory to <5 grams. The Tritium Systems Test Assembly (TSTA) at Los Alamos continues to be at the forefront of tritium technology and safety development for the fusion energy program. (18 refs.)
BAGATIN, M., A. BUFFA, et al. (1995). "RFX diagnostics." Fusion Engineering and Design 25(4): 425-60.
In this paper the diagnostic systems implemented on the reversed field pinch (RFP) experiment RFX are presented. The paper is mainly concerned with the general engineering aspects and with the development work addressed to meet specific requirements such as immunity to electrical noise, ultrahigh vacuum compatibility, reliability. A general guideline adopted in the development of diagnostics was to use, whenever possible, proven methods, at least for the systems dedicated to the measurement of the most significant parameters. Each section of the paper illustrates a different system, in which the diagnostic constraints and the specific choices derived from the peculiarities of the RFP configuration are discussed. Although the majority of the diagnostic instruments are at present fully developed and operational, some of them, i.e. far-IR polarimeter, microwave reflectometer, soft X-ray and bolometer tomographic system are either under prototype testing or under construction. (57 refs.)
Batha, S. H., F. M. Levinton, et al. (1995). "Experimental study of toroidicity induced Alfven eigenmode (TAE) stability at high q(0)." Nuclear Fusion
4th IAEA Technical Committee Meeting 35(12): 1463-8.
Experiments to destabilize the toroidicity induced Alfven eigenmode (TAE) by energetic alpha particles were performed on the Tokamak Fusion Test Reactor using deuterium and tritium fuel. To decrease the alpha particle pressure instability threshold, discharges with an elevated value of q(0)>1.5 were used. By raising q(0), the radial location of the low toroidal mode number TAE gaps moves towards the magnetic axis and into alignment with the region of maximum alpha pressure gradient, thereby (in theory) lowering the value of the central alpha particle beta {sub alpha } required for instability. No TAE activity was observed when beta {sub alpha }(0) reached 0.08% in a discharge with fusion power of 2.4 MW. Calculations predict that the alpha driven TAE was weakly unstable. (25 refs.)
Bell, M. G., K. M. McGuire, et al. (1995). "Overview of DT results from TFTR." Nuclear Fusion
4th IAEA Technical Committee Meeting 35(12): 1429-36.
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 {sub P} discharge following a current ramp-down. The fusion power density in the core of the plasma has reached 2.8 MW/m{sup 3}, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER). The energy confinement time tau {sub E} is observed to increase in DT, relative to D plasmas, by 20% and the n{sub 1}(0).T{sub 1}(0). tau {sub 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 {sub 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. (37 refs.)
Bergsaker, H., A. Moller, et al. (1995). "Edge plasma conditions and plasma-surface interactions in Extrap T1." Journal of Nuclear Materials
11th International Conference on Plasma Surface Interactions in Controlled Fusion Devices (PSI-1l) 222: 712-716.
The T1 device is a toroidal machine with R/a = 0.5 m/0.06 m and stainless steel wall. It has been used to study high current density reversed field pinch plasmas. The edge plasma has been extensively investigated with double and triple Langmuir probes, heat flux probes and passive probes. Discharges with 40-90 kA plasma current is characterized by an edge density n(e)(a) approximate to 10(19) m(-3), T-e(a) approximate to 10 eV and a high asymmetric parallel heat flux. A drift perpendicular to the magnetic field can be inferred from Langmuir probe measurements and passive probes and appears to be largely diamagnetic. Edge ion temperatures T-i greater than or equal to 100 eV are inferred from the saturation behaviour of hydrogen ions implanted in passive probes, from the high q(i parallel to) and from the inferred sputtering rate of metals.
BOHNET, M. A., J. P. GALAMBOS, et al. (1995). "The transient internal probe: a novel method for measuring internal magnetic field profiles." Review of Scientific Instruments 66(2): 1197-1200.
The transient internal probe (TIP) diagnostic is designed to permit internal magnetic field measurements in hot, high density plasmas. A small probe is fired through the plasma at high velocities and magnetic field measurements are accomplished using Faraday rotation within the Verdet glass probe. Magnetic field resolution of +or-40 G and spatial resolution of 5 mm have been achieved. System frequency response is 10 MHz. Ablative effects are avoided by minimizing both the probe size and the time the probe spends in the plasma. A two-stage light-gas gun is used to accelerate the probe (held by a sabot) to 2.2 km/s. The sabot is removed using gas dynamic forces and a gas interface system prevents the helium muzzle gas from entering the plasma chamber. Work is underway to integrate the TIP diagnostic with laboratory plasma experiments. (5 refs.)
Budny, R. V., M. G. Bell, et al. (1995). "Simulations of alpha parameters in a TFTR DT supershot with high fusion power." Nuclear Fusion
4th IAEA Technical Committee Meeting 35(12): 1497-508.
A TFTR supershot with a plasma current of 2.5 MA, a neutral beam heating power of 33.7 MW and a peak DT fusion power of 7.5 MW is studied using the TRANSP plasma analysis code. Simulations of alpha parameters such as the alpha heating, pressure and distributions in energy and v{sub 1}/v are given. The effects of toroidal ripple and mixing of the fast alpha particles during the sawteeth observed after the neutral beam injection phase are modelled. The distributions of alpha particles on the outer midplane are peaked near forward and backward v{sub 1}/v. Ripple losses deplete the distributions in the vicinity of v{sub 1}/v=-0.2. Sawtooth mixing of fast alpha particles is computed to reduce their central density and broaden their width in energy. (25 refs.)
Bush, C. E., S. A. Sabbagh, et al. (1995). "Deuterium-tritium high confinement (H-mode) studies in the tokamak fusion test reactor." Physics of Plasmas
36th Annual Meeting of the Division-of-Plasma-Physics of the American-Physical-Society 2(6): 2366-2374.
High or enhanced confinement (H-mode) plasmas have been obtained for the first time with nearly equal concentrations of deuterium and tritium in high-temperature, high poloidal beta plasmas in the tokamak fusion test reactor (TFTR) (McGuire, Phys. Plasmas 2, 2176 (1995)). Tritium fueling was provided mainly through high-power neutral beam injection (NBI) with powers up to 31 MW and beam energies of 90-110 keV. A transition to a circular limiter H-mode configuration has been obtained, following a programmed rapid decrease of the plasma current. Isotope effects, due to the presence of tritium, led to different behavior for deuterium-deuterium (DD) and deuterium-tritium (DT) H-modes relative to confinement, edge localized magnetohydrodynamic modes (ELMs), and ELM effects on fusion products. However, the threshold power for the H-mode transition was the same in DD and DT some of the highest values of the global energy confinement time, tau {sub E}, have been achieved on TFTR during the ELM-free phase of DT H-mode plasmas. Enhancements of tau {sub E} greater than four times the L-mode have been attained. (26 refs.)
BUSH, C. E., R. E. BELL, et al. (1995). "Improvements in the CHERS system for DT experiments on TFTR." Review of Scientific Instruments 66(2): 1193-1196.
Improvements in the charge exchange recombination spectroscopy (CHERS) system have resulted in accurate measurements of T{sub i} and V{sub phi } profiles during DT experiments. These include moving the spectrometer detector array and electronics further away from the tokamak to a low neutron flux location. This relocation has also improved access to all components of the system. Also, a nonplasma-viewing calibration fiber system was added to monitor the change in fiber transmission due to the high flux DT neutrons. Narrow band filtered light transmitted through the calibration fiber is now used as a reference for the V{sub phi } measurement. At a high neutron flux of approximately 2.5*10{sup 18} neutrons/s (peak fusion power approximately 6.2 MW) with total yield of 1.3*10{sup 18} neutrons a modest 5% decrease in fiber transmission was observed. Corrections for transmission loss are made and T{sub i}(r,t) and V{sub phi }(r,t) profiles are automatically calculated within four minutes of every shot. (8 refs.)
CANDY, J. and M. N. ROSENBLUTH (1995). "Mode structure and stability of toroidal Alfven eigenmodes in ITER and TFTR DT plasmas." Nuclear Fusion 35(9): 1069-97.
The small-inverse-aspect-ratio boundary layer approximation, which has been used previously to describe the analytic structure of a driven non-ideal toroidal Alfven eigenmodes (TAEs), is applied to a numerical stability calculation for the TAEs in ITER and TFTR plasmas. Away from TAE gaps (singular layers), zero beta cylindrical magnetohydrodynamics (MHDs) determines the generic structure of the outer solutions. Within each gap, a detailed kinetic treatment is used to include (i) modifications to the fluid equations arising from E{sub }// and finite Larmor radius, (ii) collisional damping from trapped electrons, (iii) collisionless (Landau) damping from passing ions and (iv) drive from finite-orbit-width fusion alpha particles and beam ions. The model is valid for arbitrary toroidal mode number and predicts the growth/damping rate of both the MHD-like TAE (that which is predicted by MHD theory) and the relevant kinetic TAEs. (26 refs.)
CHAKRABARTI, N. and P. K. KAW (1995). "Rayleigh-Taylor vortices and their stability in magnetized plasmas." Physics of Plasmas 2(5): 1460-5.
Recent experiments and computer simulations have shown that the probability distribution function (PDF) of density and potential fluctuations associated with two-dimensional Rayleigh-Taylor turbulence in collisionless plasmas, maybe exhibiting significant non-gaussian features. A possible interpretation of these results is that nonlinearly saturated coherent structures are formed which are however transient, because they are themselves unstable to fine scale secondary instabilities. An analytical investigation has been carried out to investigate some of these effects. The two-dimensional nonlinear equations describing the Rayleigh-Taylor turbulence are analytically solved; a dipole vortex solution has been obtained and its stability to secondary perturbations examined. Results show that there is a significant growth of the secondary perturbation due to ellipticity of the vortex. (15 refs.)
Chang, Z., E. D. Fredrickson, et al. (1995). "Alfven frequency modes at the edge of TFTR plasmas." Nuclear Fusion
4th IAEA Technical Committee Meeting 35(12): 1469-79.
An Alfven frequency mode (AFM) is very often seen in TFTR neutral beam heated plasmas as well as in ohmic plasmas. This quasi-coherent mode has so far only been seen on 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 on fast ion loss (e.g., the alpha particles in DT experiments) has been observed. This mode is not the conventional TAE (toroidicity induced Alfven eigenmode). The present TAE theory cannot explain this observation. Other possible explanations are discussed. (18 refs.)
Dendy, R. O., K. G. McClements, et al. (1995). "Ion cyclotron emission due to collective instability of fusion products and beam ions in TFTR and JET." Nuclear Fusion
4th IAEA Technical Committee Meeting 35(12): 1733-42.
Ion cyclotron emission (ICE) has been observed from neutral beam heated TFTR, and JET tritium experiments at sequential cyclotron harmonics of both fusion products and beam ions. The emission originates from the outer midplane plasma, where fusion products and beam ions are likely to have a drifting ring-type velocity-space distribution that is anisotropic and sharply peaked. Fusion product driven ICE in both TFTR and JET can be attributed to the magnetoacoustic cyclotron instability, which involves the excitation of obliquely propagating waves on the fast Alfven/ion Bernstein branch at cyclotron harmonics of the fusion products. Differences between ICE observations in JET and TFTR appear to reflect the sensitivity of the instability growth rate to the ratio v{sub birth}/c{sub A} where v{sub birth} is the fusion product birth speed and c{sub A} is the local Alfven speed for fusion products in the outer midplane edge of TFTR supershots, v{sub birth}<c{sub A} for alpha particles in the outer midplane edge of JET, the opposite inequality applies. If sub-Alfvenic fusion products are isotropic or have undergone even a moderate degree of thermalization, the magnetoacoustic instability cannot occur. In contrast, the super-Alfvenic alpha particles that are present in the outer midplane of JET can drive the magnetoacoustic cyclotron instability even if they are isotropic or have a relatively broad distribution of speeds. These conclusions may account for the observation that fusion product driven ICE in JET persists for longer than fusion product driven ICE in TFTR. Moreover, the time evolution of the maximum growth rate, obtained using the Sigmar model for the alpha particle distribution and TFTR data for the fusion product source rate, closely follows the observed time evolution of the ICE amplitude in TFTR supershot discharges. Other observed features of fusion product driven ICE that match the linear instability include the scaling with fusion product density, doublet splitting of spectral peaks, the relative strength of certain harmonics and source localization. A separate mechanism is proposed for the excitation of beam driven ICE in TFTR: electrostatic ion cyclotron harmonic waves, supported by strongly sub-Alfvenic beam ions, can be destabilized by a low concentration of such ions with a very narrow spread of velocities in the parallel direction. Sufficiently narrow distributions are likely to exist in the edge plasma, close to the point of beam injection. (25 refs.)
EFTHIMION, P. C., L. C. JOHNSON, et al. (1995). "Tritium particle transport experiments on TFTR during D-T operation." Physical Review Letters 75(1): 85-8.
The t(d,n) alpha and d(d,n){sup 3}He neutron emissivity profiles are measured in a deuterium neutral-beam-heated plasma where a small amount of tritium (T) gas has been puffed. The tritium density is inferred from the neutron emissivities, and transport coefficients (D, V) are determined. The particle diffusivities of T and {sup 4}He and the thermal diffusivity are similar in magnitude and profile shape. The convective velocity is small for r/a<0.6, and is anomalous for r/a>0.6. These are the first measurements of D and V for a hydrogen isotope in a tokamak plasma. (28 refs.)
FISHER, R. K., J. M. MCCHESNEY, et al. (1995). "Measurements of fast confined alphas on TFTR." Physical Review Letters 75(5): 846-9.
This paper reports the first measurements of the fast confined alpha -particle energy distribution in a fusion plasma. The pellet charge exchange technique shows the fusion generated alpha 's in the core of the Tokamak Fusion Test Reactor plasma slow down classically, and appear to be well confined. Preliminary indications are that stochastic ripple effects are responsible for steepening the energy distribution outside the plasma core (r/a>or approximately=0.35). Sawteeth mixing of fast alpha 's is suggested in data during the post-beam-heating plasma decay. (22 refs.)
FREDRICKSON, E. D., K. MCGUIRE, et al. (1995). "beta limit disruptions in the Tokamak Fusion Test Reactor." Physics of Plasmas 2(11): 4216-29.
A disruptive beta limit ( beta =plasma pressure/magnetic pressure) is observed in high-performance plasmas in the Tokamak Fusion Test Reactor (TFTR) (K.M. McGuire et al., Plasma Phys. Controlled Nuclear Fusion 1, 421 (1987)). The magnetohydrodynamic character of these disruptions differs substantially from the disruptions in high-density plasmas (density limit disruptions) on TFTR. The high beta disruptions can occur with less than a millisecond warning in the form of a fast growing precursor. The precursor appears to be an n=1 kink strongly coupled through finite p effects and toroidal terms to higher m components. It does not have the "cold bubble" structure found in density limit disruptions. The n=1 kink, in turn, appears to excite a ballooning-type mode that may contribute to the thermal quench. (34 refs.)
FUCHS, V., A. K. RAM, et al. (1995). "Mode conversion and electron damping of the fast Alfven wave in a tokamak at the ion-ion hybrid frequency." Physics of Plasmas 2(5): 1637-47.
Mode conversion of the fast Alfven wave (FAW) at the ion-hybrid frequency in the ion cyclotron range of frequencies (ICRF) is studied in the presence of ion cyclotron absorption and direct electron damping in a tokamak plasma. The usual Budden model is extended to include the effect of electron damping and of the high-field-side cutoff, and is solved analytically and numerically. The mode-conversion efficiency is given as a function of the Budden transmission coefficient and of a phase integral, which describes interference between the incoming and outgoing waves. In incidence from the low-field side, a discrete spectrum of phases exists for which complete absorption (i.e., combined mode conversion and direct electron damping) of the FAW for a single transit of the resonance region can be achieved. This permits efficient electron heating and/or current drive via mode conversion of FAWs. (28 refs.)
FUKUYAMA, A., K. ITOH, et al. (1995). "Transport simulation on L-mode and improved confinement associated with current profile modification." Plasma Physics and Controlled Fusion 37(6): 611-31.
A unified model of the L-mode confinement in tokamaks and the improved modes associated with current profile modification is investigated by means of a one-dimensional transport simulation. The thermal transport coefficient employed is based on the theory of self-sustained turbulence due to the current-diffusivity-driven modes. In the case of low beta {sub p} ( beta {sub p} being the ratio of the plasma pressure to the pressure of the poloidal magnetic field), the simulation results show fairly good agreement with empirical L-mode scaling laws of the thermal energy confinement time tau {sub E}, indicating favourable dependence on the plasma current. When beta {sub p} exceeds about unity, however, the transport in the core region is strongly reduced. This confinement improvement is attributed to the weak or negative magnetic shear due to the bootstrap current and the Shafranov shift of the magnetic surface. The enhancement factor of tau {sub E} scales as beta {sub p}{sup 0.76} and is consistent with experimental observation. The effect of current profile modification due to the current ramp down and the lower hybrid current drive is also studied. (36 refs.)
Garzotti, L., P. Innocente, et al. (1995). "A pellet ablation diagnostic system for the RFX reversed field pinch." Review of Scientific Instruments
10th Topical Conference on High Temperature Plasma Diagnostics 66(1): 616-618.
In view of the installation of a multiple pellet injector on the RFX experiment, an integrated diagnostic system has been designed to monitor the pellet behavior in the plasma. The evidence in the reversed field pinches (RFPs) of large poloidal and toroidal pellet deflections along magnetic field lines due to asymmetric ablation, calls for a 3D system with good space and time resolution. Two arrays of H{sub alpha } detectors will view the trajectories of the pellets from below and from behind and a digital charged coupled device (CCD) camera will view the injection poloidal section from a tangential viewpoint. The two systems will allow the complete determination of the time-resolved trajectory of each pellet. The space integrated H{sub alpha } emission rate will be measured by a large angle detector and by a space integration of the CCD camera measurement. (7 refs.)
Gentle, K. W. (1995). "Diagnostics for magnetically confined high-temperature plasmas." Reviews of Modern Physics 67(4): 809-36.
During the last 20 years, magnetically confined laboratory plasmas of steadily increasing temperatures and densities have been obtained, most notably in tokamak configurations, and now approach the conditions necessary to sustain a fusion reaction. Even more important to the goal of understanding the physics of such systems; remarkable advances in plasma diagnostics, the techniques for determining the properties of such plasmas, have accompanied these developments. More parameters can: be determined with greater accuracy and finer spatial and temporal resolution. The magnetic configuration, the primary local thermodynamic quantities (density, temperature, and drift velocity), and other necessary quantities can now be measured with sufficient accuracy to determine particle and energy fluxes within the plasma and to characterize the basic transport processes. These plasmas are far from thermodynamic equilibrium. This deviation manifests itself in a variety of instabilities on several spatial and temporal scales, many of which are aptly described as turbulence. Many aspects of the turbulence can also de characterized. This article reviews the current state of diagnostics from an epistemological perspective: the capabilities and limitations for measuring each important physical quantity are presented. (92 refs.)
Goloborodko, V. Y., S. N. Reznik, et al. (1995). "Fokker-Planck 3-D modelling of axisymmetric collisional losses of fusion products in TFTR." Nuclear Fusion
4th IAEA Technical Committee Meeting 35(12): 1523-35.
The results of a 3-D (in constants of motion space) Fokker-Planck simulation of the collisional losses of fusion products in axisymmetric DT and DD discharges in TFTR are presented. The distributions of escaped ions over poloidal angle and pitch angle, and their energy spectra, are obtained. Axisymmetric collisional losses of fusion products are found to be less than 2 to 5%. The distribution of confined fusion products is shown to be strongly anisotropic and non-uniform in the radial co-ordinate, mainly for slowed-down fusion products with small longitudinal energy. A comparison between these modelling results and experimental data is made. (22 refs.)
Grisham, L. R., J. H. Kamperschroer, et al. (1995). "Operations with tritium neutral beams on the tokamak fusion test reactor." Nuclear Instruments & Methods in Physics Research Section B-Beam Interactions With Materials and Atoms
13th International Conference on the Application of Accelerators in Research and Industry 99(1-4): 353-356.
In November of 1993 the tokamak fusion test reactor began operating with a deuterium-tritium fuel mixture instead of the pure deuterium which it had used heretofore. The major portion of this tritium has been supplied as energetic neutral particles injected by the neutral beams. After an initial run in which some ion sources used a mixture of 2% T and 98% D to test tokamak systems, full tritium beam operations commenced, with some of the ion sources run on pure tritium and some on deuterium to optimize the fuel mixture in the core plasma. Hundreds of tritium source shots have now occurred, with reliability which is better than that typical of deuterium operation. The maximum power injected with deuterium and tritium beams was 39.6 MW. D-T fusion power levels of up to 10.7 MW have been produced. Energy confinement in D-T plasmas of the ''supershot'' variety appears to be better than in similar deuterium plasmas.
Gronvoll, E., J. Trulsen, et al. (1995). "Electrostatic ion-Bernstein waves in the Blaamann device." Plasma Physics and Controlled Fusion 37(11): 1263-75.
This article is based on wave experiments performed in the toroidal plasma device Blaamann, and a comparison between experimental and theoretical values for the dispersion relation. A motorized probe, which could be moved in two axial directions, gave us the opportunity to investigate spatial dependencies, in our case wavefronts of ion-Bernstein waves. The article demonstrates how wave measurements of this kind can be used for obtaining ion temperature and density relations between the ion species H{sup +}, H{sub 2}{sup +} and H{sub 3}{sup +} in a weakly ionized hydrogen gas. The resulting relations are compared with what was obtained in a similar experimental set-up in the plasma device Thorello in Italy. (17 refs.)
GUREVICH, A. V., A. V. LUKYANOV, et al. (1995). "Kinetic theory of the anomalous transport of superthermal electrons in toroidal devices." Nuclear Fusion 35(7): 827-42.
An investigation has been made of the behaviour of fast electrons in toroidal discharges. The kinetic equation, describing the evolution of the fast particle distribution has been derived and analysed. A semi-analytical solution of the kinetic equation has been obtained in the superthermal energy region. The external electric field, turbulence, the non-uniformity of the mean magnetic field and collisions are shown to be important factors affecting the distribution function. The role of the ambipolar electric field has been established and identified as an essential factor in the process of fast electron diffusion. The strong influence of the density profile on the diffusion of fast particles is clearly demonstrated. A comparison has been made with experimental data obtained on the ZT-40M device. Agreement with the results of these experiments is observed. (37 refs.)
Hattori, K., Y. Hirano, et al. (1995). "Power balance in a reversed-field-pinch plasma with partial poloidal current drive by neutral beams." Fusion Technology 28(4): 1619-33.
Zero-dimensional power balance is analyzed, and an operation boundary is deduced in a "beam-assisted reversed-field pinch"; the latter utilizes partial poloidal current drive by neutral beams so that transport losses arising from magnetohydrodynamics (MHD)-dynamo, i.e., tearing mode instability are reduced. Changes of power flow and heat conductivity due to a beam driven current are treated by considering an MHD-dynamo-based power balance model that assumes linear dependence of magnetic fluctuation level on the externally driven current. It is shown that a ratio of a beam driven current to a dynamo current must not exceed approximately 40% regarding a beta-limit in the next generation of plasma experiments (minor radius/major radius=0.6 m/1.8 m, plasma current=1 MA, poloidal beta=0.1). At that point, the energy confinement time is predicted to increase by a multiple or so of that estimated from the MHD dynamo model without a current drive. (41 refs.)
HATTORI, K., K. HAYASE, et al. (1995). "First results of the toroidal divertor experiment on the TPE-2M reversed field pinch." Nuclear Fusion 35(8): 981-7.
A reversed field pinch configuration with a quadra-null toroidal divertor has been sustained for a much longer time than the magnetic diffusion time on the TPE-2M device. By applying toroidal divertor fields, the plasma resistance has been reduced by 40% in low plasma current operations (40-55 kA), along with a decrease in the line emission of light impurities and limiter materials (molybdenum). Direct measurement of internal magnetic fields by inserted magnetic probes has shown that a separatrix exists within the limiter radius. Though the behaviour of the slow magnetic fluctuations has not deteriorated, fast and spiky m=0 mode fluctuations have been excited, especially near the reversal surface. (21 refs.)
Hoffman, A. L. (1995). "Reactor prospects and present status of field-reversed configurations." Fusion Technology
6th International Toki Conference on Plasma Physics and Controlled Nuclear Fusion - Research for Advanced Concepts (ITC-6) 27: 91-6.
Field-reversed configurations (FRC) have an ideal geometry for a reactor, combining high beta toroidal confinement, with a linear external geometry. Present small diameter FRCs are thought to be stabilized by kinetic effects, but recent experiments in the Large s Experiment (LSX) have demonstrated stability well into the MHD regime. Present empirical transport coefficients are already sufficient for a small pulsed reactor, but small steady state reactors will require about an order of magnitude reduction in plasma diffusivity. (13 refs.)
HOYT, R. P., J. T. SCHEUER, et al. (1995). "Magnetic nozzle design for coaxial plasma accelerators." IEEE Transactions on Plasma Science 23(3): 481-94.
Magnetic nozzles have great potential for improving the efficiency and performance of coaxial plasma accelerators in applications such as space propulsion and advanced manufacturing. Proper design of magnetic field geometry can improve coaxial accelerator performance in three ways. First, the applied field which intercepts the anode surface without directly connecting the two electrodes can minimize anode fall inefficiencies by improving electron conduction across the anode sheath and by opposing the Hall-induced starvation effect. Second, a properly designed magnetic geometry can provide a nozzling mechanism to permit the plasma to accelerate smoothly and efficiently from sub to super-magnetosonic flow. Third, a magnetic nozzle provides control over the flow of plasma from the accelerator. For applications such as surface modification and etching, magnetic nozzles can maximize the treatable surface area and tailor the downstream plasma energy distribution. For thrust generation, proper design of a magnetic nozzle can enable efficient detachment of the plasma from the magnetic field. (41 refs.)
ISHIJIMA, D., M. IIDA, et al. (1995). "Heat flux studies in a poloidal-divertor reversed-field pinch." Plasma Physics and Controlled Fusion 37(6): 657-66.
Heat flux measurements have been performed with calorimeter probes in closed poloidal-divertor reversed-held pinch (RFP) discharges in the STE-2. A high heat flux with large anisotropy has been observed at the edge of the main plasma, indicating the existence of the fast edge electrons in divertor RFP discharges. The spatial distribution measurements show that a fraction of the fast edge electrons is also exhausted into the divertor chamber as fast (superthermal) electrons. It is also shown that most of the parallel heat flux in the divertor chamber is carried by the fast electrons. (18 refs.)
JI, H., Y. YAGI, et al. (1995). "Effect of collisionality and diamagnetism on the plasma dynamo." Physical Review Letters 75(6): 1086-9.
Fluctuation-induced dynamo electric fields are measured over a wide range of electron collisionality in the edge of TPE-1RM20 reversed-field pinch (RFP). In the collisionless region the magnetohydrodynamic dynamo alone can sustain the parallel current, while in the collisional region a new dynamo mechanism resulting from the fluctuations in the electron diamagnetic drift becomes dominant. A comprehensive picture of the RFP dynamo emerges by combining with earlier results from MST and REPUTE RFPs. (23 refs.)
Johnson, D. W., V. Arunasalam, et al. (1995). "Recent D-T results on TFTR." Plasma Physics and Controlled Fusion
22nd European-Physical-Society Conference on Controlled Fusion and Plasma Physics 37: A69-A85.
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.
Kamperschroer, J. H., L. R. Grisham, et al. (1995). "Observation of Doppler shifted T alpha emission from TFTR tritium neutral beams." Review of Scientific Instruments
10th Topical Conference on High Temperature Plasma Diagnostics 66(1): 632-634.
195 tritium ion source shots were injected into Tokamak Fusion Test Reactor (TFTR) high power plasmas during December 1993-March 1994. In addition, four highly diagnosed pulses were fired into the calorimeter. Analysis of the Doppler shifted T alpha emission of the beam in the neutralizer has revealed that the extracted ion compositions for deuterium and tritium are indistinguishable: 0.72+or-0.04 D{sup +}; 0.22+or-0.02 D{sub 2}{sup +}; 0.07+or-0.01 D{sub 3}{sup +} compared to 0.72+or-0.04 T{sup +}; 0.23+or-0.02 T{sub 2}{sup +}; 0.05+or-0.01 T{sub 3}{sup +}. The resultant tritium full-energy neutral fraction is higher than for deuterium due to the increased neutralization efficiency at lower velocity. To conserve tritium, it was used only for injection and a few calorimeter test shots, never for ion source conditioning. When used, the gas species were switched to tritium only for the shot in question. This resulted in an approximately 2% deuterium contamination of the tritium beam and vice versa for the first deuterium pulse following tritium. Data from the calorimeter shots indicate that tritium contamination of the deuterium beam cleans up in five to six beam pulses, and is reduced to immeasurable quantities prior to deuterium beam injection. (17 refs.)
KANNO, R., A. ISHIDA, et al. (1995). "Ideal-magnetohydrodynamic-stable tilting in field-reversed configurations." Journal of the Physical Society of Japan 64(2): 463-78.
The tilting mode in field-reversed configurations (FRC) is examined using ideal-magnetohydrodynamic stability theory. Tilting, a global mode, is the greatest threat for disruption of FRC confinement. Previous studies uniformly found tilting to be unstable in ideal theory: the objective here is to ascertain if stable equilibria were overlooked in past work. Solving the variational problem with the Rayleigh-Ritz technique, tilting-stable equilibria are found for sufficiently hollow current profile and sufficient racetrackness of the separatrix shape. Although these equilibria were not examined previously, the present conclusion is quite surprising. Consequently checks of the method are offered. Even so it cannot yet be claimed with complete certainty that stability has been proved: absolute confirmation of ideal-stable tilting awaits the application of more complete methods. (47 refs.)
Kudo, H. (1995). "Radioactivity and fusion energy." Radiochimica Acta 70/71: 403-12.
Nuclear fusion is expected to give an ultimate solution to energy problems over the long term. From recent progress in developing technology for fusion reactors, we can anticipate a prototype fusion reactor by 2030. This review article describes the present status of nuclear fusion research, including muon catalyzed fusion ( mu CF) which attracts quite new physical interest. Tritium is an essential component of fusion reactors, because the first-stage fusion reactors will utilize a mixture of deuterium and tritium as their fuel. The knowledge about tritium as well as the fusion-neutron induced radioactivity is summarized in terms of nuclear fusion research. (75 refs.)
LINDL, J. (1995). "Development of the indirect-drive approach to inertial confinement fusion and the target physics basis for ignition and gain." Physics of Plasmas 2(11): 3933-4024.
Inertial confinement fusion (ICF) is an approach to fusion that relies on the inertia of the fuel mass to provide confinement. To achieve conditions under which inertial confinement is sufficient for efficient thermonuclear burn, a capsule (generally a spherical shell) containing thermonuclear fuel is compressed in an implosion process to conditions of high density and temperature. ICF capsules rely on either electron conduction (direct drive) or X-rays (indirect drive) for energy transport to drive an implosion. In direct drive, the laser beams (or charged particle beams) are aimed directly at a target. The laser energy is transferred to electrons by means of inverse bremsstrahlung or a variety of plasma collective processes. In indirect. Drive, the driver energy (from laser beams or ion beams) is first absorbed in a high-Z enclosure (a hohlraum), which surrounds the capsule. The material heated by the driver emits X-rays, which drive the capsule implosion. For optimally designed targets, 70%-80% of the driver energy can be converted to X-rays. The optimal hohlraum geometry depends on the driver. Because of relaxed requirements on laser beam uniformity, and reduced sensitivity to hydrodynamic instabilities, the U.S. ICF Program has concentrated most of its effort since 1976 on the X-ray or indirect-drive approach to ICF. As a result of years of experiments and modeling, we are building an increasingly strong case for achieving ignition by indirect drive on the proposed National Ignition Facility (NIF). The ignition target requirements for hohlraum energetics, radiation symmetry, hydrodynamic instabilities and mix, laser plasma interaction, pulse shaping, and ignition requirements are all consistent with experiments. The NIF laser design, at 1.8 MJ and 500 TW, has the margin to cover uncertainties in the baseline ignition targets. In addition, data from the NIF will provide a solid database for ion-beam-driven hohlraums being considered for future energy applications. In this paper we analyze the requirements for indirect drive ICF and review the theoretical and experimental basis for these requirements. Although significant parts of the discussion apply to both direct and indirect drive, the principal focus is on indirect drive. (266 refs.)
MANSFIELD, D. K., J. D. STRACHAN, et al. (1995). "Enhanced performance of deuterium-tritium-fueled supershots using extensive lithium conditioning in the Tokamak Fusion Test Reactor." Physics of Plasmas 2(11): 4252-6.
In the Tokamak Fusion Test Reactor (TFTR) (K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)) a substantial improvement in fusion performance has been realized by combining the enhanced confinement due to tritium fueling with the enhanced confinement due to extensive conditioning of the limiter with lithium. This combination has resulted in not only significantly higher global energy confinement times than have previously been obtained in high current supershots, but also in the highest central ratio of thermonuclear fusion output power to input power observed to date. (21 refs.)
Martin, P., V. Antoni, et al. (1995). "Target emission probe for suprathermal electron detection." Review of Scientific Instruments
10th Topical Conference on High Temperature Plasma Diagnostics 66(1): 431-433.
The design and preliminary results of a target emission probe (TEP), applied to the continuous detection of suprathermal electrons in the edge of fusion plasmas, is presented in this paper. This diagnostic, developed for the RFX reversed field pinch experiment, is based on the measurement of the X-ray emission from a material target, exposed to the edge plasma, due to the bremsstrahlung of the electrons. The diagnostic has been successfully tested in RFX and a clear indication of the presence of a mostly unidirectional fast electron flow in the RFX plasma edge has been found. (19 refs.)
MAYO, R. M., R. A. GERWIN, et al. (1995). "Theory for a direct magnetic-field measurement of microturbulence enhanced electron collisionality." Physics of Plasmas 2(1): 337-9.
A theoretical framework is established for the description of anchored field-line distortion by a flowing resistive magnetofluid. The Hall effect is included and significantly influences field-line deformation in two dimensions when the electron magnetization parameter, omega {sub e}/ nu {sub e}>or approximately=1. Moreover, the formalism described herein provides for a direct and local magnetic-held determination of the effective electron collisionality, which may then be gauged against microturbulence descriptions of enhanced transport. (6 refs.)
MAYO, R. M., M. A. BOURHAM, et al. (1995). "A magnetized coaxial source facility for the generation of energetic plasma flows." Plasma Sources, Science and Technology 4(1): 47-55.
The design features and characterization of a new magnetized coaxial plasma source facility for the generation of energetic plasma stream flows are presented. Careful, yet simple design of a low-inductance (<or approximately=200 nH) external power delivery system provides high-current (>or approximately=130 kA) pulses to the plasma source with fast rise ( approximately 26 mu s). This allows for the production and acceleration of dense ( approximately 2*10{sup 15} cm{sup -3}), high-quality (10-40 eV), self-field plasmas to approximately 10{sup 5} m s{sup -1}. We are then afforded the laboratory study of plasma streams with a wide variety of applications including, but not limited to, advanced thrusters for electric propulsion, astrophysical jets, critical ionization velocity, magnetic fusion, large-scale plasma etching and deposition, etc. Comparison to an ideal MHD model is made indicating reasonably good agreement in these self-field discharges, while elucidating the advantages of strong applied magnetization to provide nozzling in a magnetically constricted flow. (13 refs.)
McGuire, K. M., H. Adler, et al. (1995). "Review of deuterium-tritium results from the Tokamak Fusion Test Reactor." Physics of Plasmas
36th Annual Meeting of the Division-of-Plasma-Physics of the American-Physical-Society 2(6): 2176-2188.
After many years of fusion research, the conditions needed for a D-T fusion reactor have been approached on the Tokamak Fusion Test reactor (TFTR) (Fusion Technol. 21, 1324 (1992)). For the first time the unique phenomena present in a D-T plasma are now being studied in a laboratory plasma. The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high- beta {sub p} discharge following a current rampdown. The fusion power density in a core of the plasma is approximately=2.8 MW m{sup -3}, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) (Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), vol. 3, p. 239) at 1500 MW total fusion power. The energy confinement time, tau {sub E}, is observed to increase in D-T, relative to D plasmas, by 20% and the n{sub i}(0) T{sub i}(0) tau {sub 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 {sub p} discharges. Ion cyclotron range of frequencies (ICRF) heating of a D-T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP (Nucl. Fusion 34, 1247 (1994)) simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of 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. D-T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor. (40 refs.)
MCKEE, G., R. FONCK, et al. (1995). "Confined alpha distribution measurements in a deuterium-tritium tokamak plasma." Physical Review Letters 75(4): 649-52.
Fusion-produced alpha particles with energy <or=0.7 MeV have been spectroscopically observed in the core of a deuterium-tritium plasma in the TFTR tokamak at alpha densities of 3*10{sup 16} m{sup -3}. During a sawtooth-free discharge, the measured energy spectra at r/a=0.3 are in good agreement with those predicted on the basis of collisional transport. Time-resolved measurements during the alpha thermalization after alpha source turn-off show decay of the distribution function to lower energies consistent with the classical slowing-down time of 0.5 s. (17 refs.)
Meade, D. M. (1995). "TFTR experience with D-T operation." Fusion Engineering and Design
3rd International Symposium on Fusion Nuclear Technology 27: 17-26.
Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, alpha confinement, alpha heating and possible alpha-driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of approximate to 20 MW of tritium and 14 MW of deuterium neutral beams into the TFTR produced a plasma with a T/D density ratio of approximate to 1 and yielded a maximum fusion power of approximate to 9.2 MW. The fusion power density in the core of the plasma was approximate to 1.8 MW m(-3), approximating that expected in a D-T fusion reactor. A TFTR plasma with a TID density ratio of approximate to 1 was found to have approximate to 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass, A, of tau(E) approximate to A(0.6). The core ion temperature increased from 30 keV to 37 keV due to a 35% improvement in ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 keV to 10.6 keV can be attributed to electron heating by the alpha-particles. The similar to 5% loss of alpha-particles, as observed on detectors near the bottom edge of the plasma, was consistent with classical first orbit loss without anomalous effects. Initial measurements have been made of the confined energetic alphas and the resultant alpha ash density. At fusion power levels of 7.5 MW, fluctuations at the toroidal Alfven eigenmode frequency were observed by the fluctuation diagnostics. However, no additional alpha loss due to the fluctuations was observed. These D-T experiments will continue over a broader range of parameters and higher power levels.
Meade, D. M. (1995). "Tokamak Fusion Test Reactor D-T results." Fusion Engineering and Design
18th Symposium on Fusion Technology (SOFT-18) 30(1/2): 13-23.
Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, alpha confinement, alpha heating and possible alpha -driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of about 20 MW of tritium and 14 MW of deuterium neutral beams into the TFTR produced a plasma with a T-to-D density ratio of about 1 and yielding a maximum fusion power of about 9.2 MW. The fusion power density in the core of the plasma was about 1.8 MWm{sup -3}, approximating that expected in a D-T fusion reactor. A TFTR plasma with a T-to-D density ratio of about 1 was found to have about 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass A of tau {sub E} approximately A{sup 0.6}. The core ion temperature increased from 30 to 37keV owing to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 to 10.6 keV can be attributed to electron heating by the alpha particles. The approximately 5% loss of alpha particles, as observed on detectors near the bottom edge of the plasma, was consistent with classical first orbit loss without anomalous effects. Initial measurements have been made of the confined high energy alpha particles and the resultant alpha ash density. At fusion power levels of 7.5 MW, fluctuations at the toroidal Alfven eigen-mode frequency were observed by the fluctuation diagnostics. However, no additional alpha loss due to the fluctuations was observed. These D-T experiments will continue over a broader range of parameters and higher power levels. (21 refs.)
MILORA, S. L., W. A. HOULBERG, et al. (1995). "Pellet fuelling." Nuclear Fusion 35(6): 657-754.
Steady progress has been made worldwide in the application and development of hydrogen pellet injection as a method for fuelling magnetically confined plasmas. The theoretical, experimental and technological aspects of this field of research are reviewed, emphasizing developments over the past decade. (411 refs.)
MINOO, H., A. ARSAOUI, et al. (1995). "An analysis of the cathode region of a vortex-stabilized arc plasma generator." Journal of Physics D (Applied Physics) 28(8): 1630-48.
In an arc plasma generator in air at atmospheric pressure, stabilized by a magnetic field and a vortex flowing gas, with a copper tubular cathode cavity, based on the analysis of forces acting on the cathode region, the shape of the arc column, the direction of the arc current density J near the cathode surface, the position of the cathode spot and its 'jumping' process are studied. The cathode spot's movement over the cathode surface is the result of a repetitious process of birth of new spots and disappearance of old spots provoked by the magnetic field and the gas flow. The spot lifetime is estimated at around (3-4)*10{sup -4} s. Moreover, following this study, the position of the plane (Z=Z{sub c}) in which the arc cathode spot rotates is determined as a function of arc current I, components of the magnetic field B{sub r} and B{sub Z}, flowing gas velocities U{sub Z} and U{sub theta }, the velocity of the arc cathode spot rotation V{sub theta }{sup arc} and the radius of the arc cathode spot a{sub 0}. The results of this study are in satisfactory agreement with experimental determinations. In this analysis cylindrical coordinates r, theta and Z are used with the symmetrical axis of the cathode tube oriented in the Z direction. (25 refs.)
MONTEIRO, L. H. A., M. Y. KUCINSKI, et al. (1995). "Coupling of modes in RFPs: an analytical approach." Plasma Physics and Controlled Fusion 37(5): 541-50.
The magnetic structure of a cylindrical RFP plasma in which the equilibrium is perturbed by two resonant modes is determined analytically by using an averaging method to solve the differential field-line equations. The primary modes couple themselves and excite others, also creating magnetic islands in secondary resonance regions. Average magnetic surfaces are compared with Poincare maps obtained by direct integration of field-line equations. (19 refs.)
Moreau, D., G. Agarici, et al. (1995). "Theory and experiments on rf plasma heating, current drive and profile control in Tore Supra." Fusion Engineering and Design
5th International Toki Conference on Plasma Physics and Controlled Nuclear Fusion - Physics and Technology of Plasma Heating and Current Drive 26(1-4): 29-48.
The combination of r.f. waves in the lower hybrid (LH) and ion cyclotron frequency ranges offers a versatile and efficient way of heating tokamak plasmas while controlling their transport properties and magnetohydrodynamic stability through the control of the current density profile. Experimental and theoretical studies on the applications of such plasma waves have been carried out on Tore Supra during recent years and are reported here. The LH system coupled up to 6.5 MW during 2 s at 3.7 GHz through two multijunction launchers. In the longest plasma shot, the total injected LH energy reaches a record value of 170 MJ during a 62 s LH pulse, at a power level of 2.8 MW, corresponding to an average power density of 17 MW m(-2). The ion cyclotron resonant frequency (ICRF) system (35-80 MHz) is composed of three resonant double-loop antennae. Up to 4 MW have been coupled with a single antenna, allowing a record power density through the Faraday screen of 16 MW m(-2) to be reached. 30 s steady-state r.f. pulses have been obtained with up to 54 MJ delivered to the plasma. One of the major observations has been the transition to the so-called ''stationary lower hybrid enhanced performance (LHEP) regime'' (I-p = 0.8 MA; n(e0) = 2.8 x 10(19) m(-3); P-LH = 3.2 MW) in which the (flat) central current density (q(0), approximate to 2) and (peaked) electron temperature profiles (T-e0 approximate to 6-8 keV) are fully decoupled. This regime exhibits a significant improvement of the global confinement (40%) owing to the increase in l(i), i.e. in the magnetic shear in the outer half of the discharge, supplemented by a large reduction in the electron thermal diffusivity in the central zone where the magnetic shear vanishes because of the slight off-axis character of the LH power and current deposition. TRANSP analyses show that LHEP plasmas provide access to the second ballooning stability regime. At higher current and density (I-p = 1.5 MA; n(e0) = 6 x 10(19) m(-3)), ICRH stabilization of sawteeth (4 MW) combined with lower hybrid current drive (LHCD) current profile modifications has allowed to extend the stabilized phase for up to 1 a with 3.4 MW of LH power, the duration of sawtooth-free periods increasing with increasing LH power. The dynamical properties of fast electrons during LHCD have been investigated recently on Tore Supra through power modulation experiments. It is shown that slowing down always predominates and, from the long-time evolution of the hard X-ray emission, the radial diffusion rate of the fast electrons is estimated to be 0.1-0.3 m(2) s(-1). Theoretical developments have focused on the modelling of LHCD and also on fast wave current drive (FWCD) and fast wave heating. The effect of intrinsic stochasticity on the propagation of LH waves is discussed and a fully developed statistical theory of stochastic wave diffusion and multipass absorption, with applications to Tore Supra through a wave diffusion-Fokker-Planck (WDFP) numerical code, is briefly presented. This model provides a simple explanation for the temperature dependence of the LHCD efficiency in small tokamaks. The ion cyclotron resonant heating (ICRH) full-wave code ALCYON has been upgraded to compute the power and current deposition profiles from direct electron absorption of the fast wave (electron Landau damping-transit time magnetic pumping). The code has been used to study FWCD in Tore Supra, the Joint European Torus and the International Thermonuclear Experimental Reactor. Finally a new concept of an efficiently cooled reflector antenna for LHCD applications in a steady-state reactor is briefly described.
NAGATA, A., H. SAKAMOTO, et al. (1995). "A field-reversal mechanism in a reversed-field pinch." Physics of Plasmas 2(4): 1182-91.
A field-reversal mechanism in a reversed-field pinch (RFP) is studied through a three-dimensional resistive compressible magnetohydrodynamic (MHD) simulation of the single- and multiple-helicity modes. As the magnetic Reynolds number increases, the m=1 fluctuating magnetic field grows exponentially and extends radially, and then the flow begins to form the vortex structure around the core of the plasma. This radial flow acts such as to push out the toroidal magnetic field. As the dynamo electric field induced by this interaction increases near the core of the plasma, the toroidal magnetic field at the wall decreases toward the negative value and the toroidal magnetic field reverses. As a result, it is found that the field reversal is achieved by the single-helicity evolution of the m=1 mode alone, without the (m=0; n not=0) modes, and the interaction of the radial flow and the toroidal magnetic field is the most dominant source for the dynamo action on the field-reversal process. (22 refs.)
Neilson, G. H., D. B. Batchelor, et al. (1995). "The Tokamak Physics Experiment: tokamak improvement through advanced steady state control." Fusion Engineering and Design
5th International Toki Conference on Plasma Physics and Controlled Nuclear Fusion - Physics and Technology of Plasma Heating and Current Drive 26(1/4): 563-74.
The achievement of a long-pulse ignited discharge with over 1000 MW of fusion power in the International Thermonuclear Experimental Reactor will be an important goal for the next phase of the world magnetic fusion program. However, improvements in the physics are needed to design a more economically attractive tokamak power reactor than the present data base would support. Advanced, steady state plasma controls are the key to realizing these improvements. The Tokamak Physics Experiment has a flexible heating and current drive system for profile control; a flexible poloidal field system that supports a strongly shaped double-null poloidal divertor plasma configuration over a wide range of profiles; and a divertor designed for dispersive operation, flexibility, and remote handling. The machine performance in deuterium is sufficient to produce a reactor-like bootstrap current profile and to confine fast electrons for localized current profile control. A conducting structure, plasma rotation, field error compensation coils, and profile control are used to provide stable plasma configurations with beta up to twice the Troyon limit and bootstrap current fraction approaching unity. The facility will be designed for 1000 s pulses initially to minimize the influence of initial transients on system behavior, but the pulse length can be extended through upgrades of external systems if necessary. (14 refs.)
OMELCHENKO, Y. A. and R. N. SUDAN (1995). "Formation of field-reversed ion rings in a magnetized background plasma." Physics of Plasmas 2(7): 2773-83.
In typical field-reversed ion ring experiments, an intense annular ion beam is injected across a magnetic cusp into neutral gas immersed in a solenoidal magnetic field. In anticipation of a new experimental thrust to create strong field-reversed ion rings the beam evolution is investigated in a preformed background plasma on a time scale greater than an ion cyclotron period, using a new two and a half-dimensional (2 1/2 -D) hybrid, particle-in-cell (PIC) code FIRE, in which the beam and background ions are treated as macro-particles and the electrons as a massless fluid. It is shown that under appropriate conditions axial beam bunching occurs in the downstream applied field and a compact field-reversed ring is formed. It is observed that the ring is reflected in a ramped magnetic field. Upon reflection its axial velocity is very much less than that expected from a single particle model due to the transfer of the mean axial momentum to the background ions. This increases the time available to apply a pulsed mirror for trapping the ring experimentally. (25 refs.)
PHILLIPS, J. A., D. A. BAKER, et al. (1995). "A global analysis of the behaviour of the ZT-40M reversed field pinch." Nuclear Fusion 35(8): 935-58.
Experimental data from the reversed field experiment, ZT-40M, have been re-examined in an attempt to determine the scaling behaviour of the physical plasma quantities and their fluctuations. A subset of the data is defined, allowing a reduced number of independent variables to describe the behaviour. For flat-top ZT-40M discharges the independent variables are chosen as being the toroidal current, I{sub phi }, and the dimensionless pinch parameter, Theta , which is proportional to the ratio of the toroidal current to the toroidal magnetic flux. The amplitudes of the dependent variables, including the electron temperature, plasma resistance, toroidal flux, the ratio of I{sub phi } to the mean electron density and their fluctuation amplitudes, exhibit minima as functions of Theta for constant I{sub phi }. These minima move towards lower Theta values with increasing I{sub phi }. Over the range of conditions for acceptable operation, the scaling of variables with I{sub phi } is not unique but depends on the variation of Theta as I{sub phi } increases. The Theta variation is governed by the specific conditions (such as constant poloidal beta, beta {sub p}) chosen to set the desired RFP operational constraints. Contour plots of the dependent variables versus the two independent variables, I{sub phi } and Theta , allow the determination of the I{sub phi }- Theta trajectory that corresponds to discharges that meet the chosen condition. The analysis shows that the amplitude of the low frequency fluctuations correlates with the mean beta {sub p} and energy confinement time of ZT-40M. By modifying the external circuits on ZT-40M, low frequency fluctuations were reduced. Comparing the designs of different RFP experiments and their operating behaviour, these modifications suggest design changes for present and future RFP experiments that will benefit their performance. (90 refs.)
Redi, M. H., R. V. Budny, et al. (1995). "Modelling TF ripple loss of alpha particles in TFTR DT experiments." Nuclear Fusion
4th IAEA Technical Committee Meeting 35(12): 1509-16.
Modelling of TF ripple loss of alphas in DT experiments on TFTR now includes neoclassical calculations of first orbit loss, stochastic ripple diffusion, ripple trapping and collisional effects. A rapid way to simulate experiment has been developed which uses a simple stochastic domain model for TF ripple loss within the TRANSP analysis code, with the ripple diffusion threshold evaluated by comparison with more accurate but computationally expensive Hamiltonian co-ordinate guiding centre code simulations. Typical TF collisional ripple loss predictions are 6-10% loss of alphas for TFTR DT experiments at Ip=1.0-2.0 MA and R=2.52 m. (22 refs.)
REDI, M. H., M. C. ZARNSTORFF, et al. (1995). "Collisional stochastic ripple diffusion of alpha particles and beam ions on TFTR." Nuclear Fusion 35(10): 1191-211.
Predictions for ripple loss of fast ions from TFTR are investigated with a guiding centre code including both collisional and ripple effects. A synergistic enhancement of fast ion diffusion is found for toroidal field ripple with collisions. The total loss is calculated to be roughly twice the sum of ripple and collisional losses calculated separately. Discrepancies between measurements and calculations of plasma beta at low current and large major radius are resolved when both effects are included for neutral beam ions. A 20 to 30% reduction in alpha particle heating is predicted for q{sub a}=6-14, R=2.6 m DT plasmas on TFTR owing to first orbit and collisional stochastic ripple diffusion. (48 refs.)
RICCARDI, C., P. CANTU, et al. (1995). "A DIAGNOSTIC METHOD OF ION MINORITY CONCENTRATIONS BASED ON IBW PROPAGATION." Plasma Physics and Controlled Fusion 37(8): 883-889.
A diagnostic method for ion minority concentrations, based on the propagation of ion Bernstein waves (IBW), is analysed theoretically and verified experimentally. The results obtained are compared with those provided by the neutralized ion Bernstein waves (NIBW) based diagnostic, as used in previous experiments.
RICCARDI, C., E. AGOSTINI, et al. (1995). "Study of plasma impedance for electrostatic waves coupling." Plasma Physics and Controlled Fusion 37(7): 763-71.
The analysis of the plasma impedance with respect to the coupling of electrostatic waves is described in this paper. We consider the impedance due to bulk (EPW, IBW) and surface waves (SPW). The impedance is calculated using the hypothesis of a magnetized homogeneous plasma and is evaluated with respect to a slow-wave antenna, for the main plasma and wave parameters. (14 refs.)
RICCARDI, C., E. AGOSTINI, et al. (1995). "Measurements of plasma loading in the presence of electrostatic waves." Physics of Plasmas 2(10): 3588-94.
An experimental analysis of the plasma impedance with respect to the coupling of ES (electrostatic) waves is described in this paper. The waves are excited through a slow-wave antenna and the experiment performed in a toroidal device (C. Riccardi et al., Plasma Phys. 36, 1791 (1994)). The measured impedance is compared with a simple theoretical model for magnetized homogeneous plasma, in order to establish the presence of bulk or surface waves and of some nonlinear effects when power is raised. (22 refs.)
ROTH, J. R. (1995). "BALL-LIGHTNING - WHAT NATURE IS TRYING TO TELL THE PLASMA RESEARCH COMMUNITY." Fusion Technology 27(3): 255-270.
Ball lightning has been extensively observed in atmospheric air, usually in association with thunderstorms, by untrained observers who were not in a position to make careful observations. These chance sightings have been documented by polling observers, who constitute perhaps 5 % of the adult U.S. population. Unfortunately, ball lightning is not accessible to scientific analysis because it cannot be reproduced in the laboratory under controlled conditions. Natural ball lightning has been observed to last longer than 90 s and to have diameters from 1 cm to several metres. The energy density of a few lightning balls has been observed to be as high as 20 000 J/cm(3), well above the limit of chemical energy storage of, for example, TNT at 2000 J/cm(3). Such observations suggest a plasma-related phenomenon with significant magnetic energy storage. If this is the case, ball lightning should have very interesting implications for fusion research, industrial plasma engineering, and military applications, as well as being of great theoretical and practical interest to the plasma research community.
Scott, S. D., M. C. Zarnstorff, et al. (1995). "Isotopic scaling of confinement in deuterium-tritium plasmas." Physics of Plasmas
36th Annual Meeting of the Division-of-Plasma-Physics of the American-Physical-Society 2(6): 2299-2307.
The confinement and heating of supershot plasmas are significantly enhanced with tritium beam injection relative to deuterium injection in the Tokamak Fusion Test Reactor (Plasma Phys. Controlled Fusion 26, 11 (1984)). The global energy confinement and local thermal transport are analyzed for deuterium and tritium fueled plasmas to quantify their dependence on the average mass of the hydrogenic ions. Radial profiles of the deuterium and tritium densities are determined from the D-T fusion neutron emission profile. The inferred scalings with average isotopic mass are quite strong with tau {sub E} varies as (A){sup 0.85+or-0.20}, tau {sub E}{sup thermal} varies as (A){sup 0.89+or-0.20}, chi {sub i}{sup tot} varies as (A){sup -2.6+or-0.5}, and D{sub e} varies as (A){sup -1.4+or-0.2} at fixed P{sub inj}. For fixed local plasma parameters chi {sub i}{sup tot} varies as (A){sup -1.8+or-0.4} is obtained. The quoted 2 sigma uncertainties include contributions from both diagnostic errors and shot irreproducibility, and are conservatively constructed to attribute the entire scatter in the regressed parameters to uncertainties in the exponent on plasma mass. (33 refs.)
SCOTT, S. D., D. R. ERNST, et al. (1995). "ISOTOPIC SCALING OF TRANSPORT IN DEUTERIUM-TRITIUM PLASMAS." Physica Scripta 51(3): 394-401.
Both global and thermal energy confinement improve in high-temperature supershot plasmas in the Tokamak Fusion Test Reactor (TFTR) when deuterium beam heating is partially or wholly replaced by tritium beam heating. For the same heating power, the tritium-rich plasmas obtain up to 22% higher total energy, 30% higher thermal ion energy, and 20-25% higher central ion temperature. Kinetic analysis of the temperature and density profiles indicates a favorable isotopic scaling of ion heat transport and electron particle transport, with tau(Ei)(a/2) proportional to [A](0.7-0.8) and tau(pe)(a) proportional to [A](0.8).
SHARMA, P. K. and D. BORA (1995). "Experimental study of a toroidal magnetized plasma in the presence of a weak vertical magnetic field." Plasma Physics and Controlled Fusion 37(9): 1003-14.
Detailed measurements of density and floating potential in the poloidal cross section of a toroidal plasma with an extended source at the minor axis are presented in the first half of the paper. The second part deals with the effect of k{sub }// on curvature-induced low frequency (< omega {sub i}, the ion gyro frequency) fluctuations, where k{sub }// is the wavenumber parallel to the ambient magnetic field. The variation in k{sub }// is introduced with the help of a weak vertical magnetic field, which helps in varying the length of the toroidal magnetic field lines. We report a reduction in the fluctuation level in the presence of a weak vertical field. (12 refs.)
SKINNER, C. H., H. ADLER, et al. (1995). "First measurements of tritium recycling in TFTR." Nuclear Fusion 35(2): 143-51.
The first spectroscopic measurements of tritium Balmer alpha (T{sub alpha }) emission from a fusion plasma were made on TFTR using a Fabry-Perot interferometer. The tritium alpha (T{sub alpha }) emission line is partially blended with the deuterium alpha (D{sub alpha }) line, commonly used in edge plasma diagnostics, and the contributions of H{sub alpha }, D{sub alpha } and T{sub alpha } are separated by spectral analysis. The data are a measure of the fuelling of the plasma by tritium accumulated in the TFTR limiter, as well as the amount of neutral tritium generated by charge exchange of plasma ions. The T{sub alpha } line first became detectable in a high power, tritium only, neutral beam injection discharge at the level of T{sub alpha }/(H{sub alpha }+D{sub alpha }+T{sub alpha })=2%. Subsequently this ratio has increased to as high as 7.5%. Data on the time evolution of the T{sub alpha } emission during a single discharge and over a series of tritium and deuterium discharges are presented. (33 refs.)
Slough, J. T., A. L. Hoffman, et al. (1995). "Transport, energy balance, and stability of a large field-reversed configuration." Physics of Plasmas
36th Annual Meeting of the Division-of-Plasma-Physics of the American-Physical-Society 2(6): 2286-2291.
Experiments have been conducted on the Large s Experiment (LSX) (Phys. Rev. Lett. 69, 2212 (1992)) field-reversed theta pinch, where plasmas confined in a field-reversed configuration (FRC) have exhibited record energy, particle, and configuration lifetimes. By careful control of the formation process, it was possible to form symmetric, quiescent FRCs with s values (the number of ion gyroradii from the field null to the separatrix of the FRC) as large as 5. LSX particle confinement showed a strong scaling with s. The inferred particle diffusivity, D{sub s}, at large s approached approximately 2 m{sup 2}/s, which, along with previous experimental results, indicate a favorable D{sub 2} approximately s{sup -1/2} scaling. At large s, both electron and ion cross-field thermal conduction losses become negligible compared to convective losses, with the inferred chi {sub perpendicular to e} approximately 4 m{sup 2}/s, which was near classical values. Data from several diagnostics employed on the LSX device were analyzed to seek correlation between distortions in the plasma shape and the confinement properties of the FRCs formed. No clear correlation between the quality of confinement and distortion was observed. Experiments conducted over a large range of s (1<s<8) appeared to be grossly stable to low-order modes, such as the internal tilt. (18 refs.)
Spong, D. A., C. L. Hedrick, et al. (1995). "Strategies for modifying alpha driven TAE thresholds through q profile and ion temperature control." Nuclear Fusion
4th IAEA Technical Committee Meeting 35(12): 1687-96.
Two methods are examined for influencing the stability thresholds of alpha driven toroidal Alfven eigenmode (TAE) instabilities in the deuterium-tritium (DT) operation of the Tokamak Fusion Test Reactor (TFTR). These are: (a) increasing the TAE drive through raising the central q(0) value (in order to centre the peak alpha pressure gradient on the lower m gap locations) and (b) decreasing the TAE damping rate on the background ions through cooling perturbations such as helium puffs or injected pellets. A gyrofluid model of the TAE is used that is inherently non-perturbative and includes continuum damping, ion/electron/neutral beam Landau damping, ion FLR effects and a generalized resistivity (finite E{sub 1} effects). Specific TFTR DT cases are analysed using a combination of measured and inferred profiles. The resulting predicted thresholds are generally consistent with an observed lack of strong TAE activity. However, it is found that special techniques, especially increasing q(0), can significantly lower the alpha driven TAE threshold into ranges that should be experimentally accessible. (37 refs.)
Sugai, H., H. Toyoda, et al. (1995). "Wall conditioning with lithium evaporation." Journal of Nuclear Materials
11th International Conference on Plasma Surface Interactions in Controlled Fusion Devices (PSI-1l) 222: 254-258.
Thin lithium layers are deposited on metal or graphite walls by evaporation in vacuum. Clean lithium surfaces have wide chemical activities on such gases as O-2, CO and CH4. In particular, a strong gettering effect on oxygen is observed; the maximum number of O atoms gettered by the lithium layer is about one half the total number of Li atoms deposited on the wall, thus suggesting the formation of Li2O. On the other hand, H-2 gas hardly reacts with the lithium surface which, however, displays a large pumping effect in a hydrogen glow discharge. The maximum number of H atoms pumped by the lithium layer below 180 degrees C is approximately equal to the number of Li atoms in the vessel, probably due to the formation of LiH. Helium glow conditioning of the used lithium layer allows the partial recovery of lithium from losses to hydrogen pumping or oxygen gettering. This lithium evaporation method has been applied to JIPP T-IIU. 100-250 mg of lithium was deposited onto a limited vessel area of similar to 1 m(2). The lithium coating leads to 20-50% reduction in oxygen and carbon impurities with less hydrogen recycling in ohmic and NBI discharges.
SYNAKOWSKI, E. J., R. E. BELL, et al. (1995). "Measurements of the production and transport of helium ash in the TFTR tokamak." Physical Review Letters 75(20): 3689-92.
Helium ash production and transport have been measured in TFTR deuterium-tritium plasmas using charge-exchange recombination spectroscopy. The helium ash confinement time, including recycling effects, is 6-10 times the energy confinement time and is compatible with sustained ignition in a reactor. The ash confinement time is dominated by edge pumping rates rather than core transport. The measured evolution of the local thermal ash density is consistent with modeling based on previously measured helium transport coefficients and classical slowing down of the alpha particles. (28 refs.)
Tennfors, E. (1995). "Alfven-wave ion heating in the reversed field pinch." Physica Scripta
International Workshop on Alfven Waves T60: 65-70.
Ion energies in reversed-field pinch plasmas are observed to be considerably higher than expected from ohmic heating and collisional transfer from electrons. Other mechanisms must be involved and seem to be related to tearing-mode fluctuations. Observations of ion heating and fluctuation spectra are reviewed here as well as various mechanisms proposed to be responsible for the ion heating. A mechanism is proposed, where induced currents due to relating tearing-mode perturbations passing by irregularities in the wall structure excite plasma waves, which in turn are absorbed at the Alfven-wave resonance. The partition of energy between electrons and ions is discussed.
TURNBULL, A. D., T. S. TAYLOR, et al. (1995). "HIGH-BETA AND ENHANCED CONFINEMENT IN A 2ND STABLE CORE VH-MODE ADVANCED TOKAMAK." Physical Review Letters 74(5): 718-721.
Villard, L., S. Brunner, et al. (1995). "Alfven waves heating and stability." Physica Scripta Volume T
International Workshop on Alfven Waves T60: 44-56.
Alfven waves in fusion plasmas play an important role in a number of situations. First, in Alfven wave heating schemes. Second, both theory and experiment have demonstrated the existence of global Alfven eigenmodes (GAEs). GAEs have been observed in different tokamaks (PRETEXT, TCA, TEXTOR, etc.) and, more recently, in a stellarator (Wendelstein 7-AS) where they were shown to become unstable under intense neutral beam injection. Third, the existence and possible destabilization by fast ions of toroidicity induced Alfven eigenmodes (TAEs) has been shown both theoretically and experimentally. This destabilization could hamper the operation of a magnetically confined fusion reactor by setting a limit on the number of fusion alpha particles in the plasma. It is therefore crucial to understand the mechanisms leading to the occurrence of the instability and also those that can stabilize the TAEs by increasing the strength of the damping. The aim is to be able to devise possible ways to avoid the instability of Alfven eigenmodes in a region of parameter space that is compatible with the functioning of a fusion reactor. A global perturbative approach is presented to tackle the problem of the linear stability of TAEs. Our model computes the overall wave particle power transfers to the different species and thus could also be applied to the study of alpha power extraction in the presence of Alfven waves. We indicate also how to go beyond the perturbative approach. (38 refs.)
VILLARD, L., S. BRUNNER, et al. (1995). "Global marginal stability of TAEs in the presence of fast ions." Nuclear Fusion 35(10): 1173-90.
The global stability of toroidicity induced Alfven eigenmodes (TAEs) in the presence of fast ions in realistic tokamak fusion-grade plasmas is analysed with a global, perturbative approach. Volume averaged fast particle betas for marginal stability are obtained and analysed for a wide range of plasma parameters such as the fast ion radial density profile width, the ratio of birth velocity to the Alfven velocity on axis and the bulk plasma beta. There is evidence of different stability behaviour of two types of TAEs ('internal' or 'external'). (29 refs.)
WANG, A. K., X. M. SONG, et al. (1995). "Fluctuation-induced current and energetic electrons in reversed field pinch plasmas." Plasma Physics and Controlled Fusion 37(6): 647-55.
In this paper, two sets of equation are derived from a set of fundamental MHD equations with a dynamo field. The first set of equations, which governs the evolution of the energies of the mean magnetic field and the mean plasma velocity with time, couples not only the mean motions with turbulent motion but also the mean motions with each other. It is pointed out that the reversed field pinch (RFP) plasma described by this set of equations can spontaneously relax to the Taylor state. The second set of equations, which governs the evolution of the helicity of the mean field and the mean helicity of the fluctuating field, depicts the helicity transfer between the mean and fluctuating field. It is shown that in the RFP plasma there exists a fluctuation-induced mean current consisting of energetic electrons, the generation of which is dominated by the dynamo field and completely determined by the dynamo field at the edge, and the energetic electron flow sustains the field-reversal configuration and reduces the anomaly in the loop voltage. (33 refs.)
Wen, Y. Z. and R. V. Bravenec (1995). "High-sensitivity, high-resolution measurements of radiated power on TEXT-U." Review of Scientific Instruments
10th Topical Conference on High Temperature Plasma Diagnostics 66(1): 549-551.
WERLEY, K. A. (1995). "A mean field description of reversed field pinch: equilibrium, dynamo, plasma heating and confinement." Nuclear Fusion 35(4): 455-66.
Three simple physics constraints are applied to describe the reversed field pinch (RFP) in a mean field sense. These constraints are: (a) a steady state equilibrium, (b) a negative radial pressure gradient, and (c) a net volumetric source of mean field magnetic energy that is always locally negative. Using these constraints, realistic RFP equilibria are calculated, plasma heating profiles are determined, and the role of the dynamo in rearranging profiles can be described by a mean field one dimensional picture. The RFP exists in a near minimum energy state characterized by a mu profile (where mu identical to mu {sub 0}aJ.B/B{sup 2}) that decreases to a small value at the plasma edge. The resulting RFP confinement picture has three regions: a poorly confined plasma core characterized by parallel (radial) transport and flux surfaces destroyed by m=1 tearing mode activity associated with dynamo relaxation, a medium confined edge limited by ideal pressure gradient driven modes, and a good confinement region located near the reversal layer. This good confinement region determines the global confinement characteristics of the RFP, and, if limited by resistive interchange modes, is consistent with the Conner-Taylor scaling that has provided a good fit to international RFP results. (25 refs.)
Wurden, G. A., R. E. Chrien, et al. (1995). "Scintillating-fiber 14 MeV neutron detector on TFTR during DT operation." Review of Scientific Instruments
10th Topical Conference on High Temperature Plasma Diagnostics 66(1): 901-903.
A compact 14 MeV neutron detector using an array of scintillating fibers has been tested on the TFTR tokamak under conditions of a high gamma background. This detector uses a fiber-matrix geometry, a magnetic field-insensitive phototube with an active HV base and pulse-height discrimination to reject low-level pulses from 2.5 MeV neutrons and intense gammas. Laboratory calibrations have been performed at EG&G Las Vegas using a pulsed DT neutron generator and a 30 kCi {sup 60}Co source as background, at PPPL using DT neutron sources, and at LANL using an energetic deuterium beam and target at a tandem van de Graaff accelerator. During the first high-power DT shots on TFTR in December 1993, the detector was 15.5 m from the torus in a large collimator. For a rate of 1*10{sup 18} n/s from the tokamak, it operated in an equivalent background of 1*10{sup 10} gammas/cm{sup 2}/s ( approximately 4 mA current drain) at a DT count rate of 200 kHz. (6 refs.)
ZWEBEN, S. J., D. S. DARROW, et al. (1995). "Alpha particle loss in the TFTR DT experiments." Nuclear Fusion 35(8): 893-917.
Alpha particle loss was measured during the TFTR DT experiments with a scintillator detector located at the vessel bottom in the ion Del B drift direction. The DT alpha particle loss to this detector was consistent with the calculated first orbit loss over the whole range of plasma current I=0.6-2.7 MA. In particular, the alpha particle loss rate per DT neutron at a given plasma current did not increase significantly with fusion power up to 10.7 MW, indicating the absence of any new 'collective' alpha particle loss processes in these experiments. (36 refs.)
Antoni, V., E. Martines, et al. (1996). "Stochastic magnetic and ambipolar electric fields in the plasma edge region of RFX." Nuclear Fusion 36(4): 435-42.
The edge region of the reversed field pinch experiment RFX has been investigated with Langmuir and calorimetric probes. The energy flux measurements reveal a spatial structure that is consistent with the presence of a superthermal tail in the energy distribution function of the electrons, as expected according to the kinetic dynamo theory (KDT). In the framework of this model, the value of the magnetic field line diffusion coefficient in the edge region has been derived. The radial electric field obtained from the plasma potential gradient is opposite in sign to the ambipolar electric field expected in a stochastic magnetic field. The discrepancy is discussed in terms of particle recycling at the wall. (44 refs.)
Barnes, C. W., S. D. Scott, et al. (1996). "Confinement analysis in low-confinement mode of hydrogen isotope experiments on the Tokamak Fusion Test Reactor." Physics of Plasmas 3(12): 4521-35.
The effect of isotope on confinement in high-recycling, L-mode plasmas is studied on the Tokamak Fusion Test Reactor (TFTR) [see D.M. Meade, J. Fusion Energy 7, 107 (1988)] by comparing hydrogen and deuterium plasmas with the same magnetic field and similar electron densities and heating power, with both Ohmic and deuterium-neutral-beam heating. Following a long operational period in deuterium, nominally hydrogen plasmas were created through hydrogen glow discharge and hydrogen gas puffing in Ohmic plasmas, which saturated the exposed limiter surface with hydrogen and raised the H/(H+D) ratio from 10+or-3% to 65+or-5%. Ohmic deuterium discharges obtained higher stored energy and lower loop voltage than hydrogen discharges with similar limiter conditions. Neutral-beam power scans were conducted in L-mode plasmas at minor radii of 50 and 80 cm, with plasma currents of 0.7 and 1.4 MA. To minimize transport differences from the beam deposition profile and beam heating, deuterium neutral beams were used to heat the plasmas of both isotopes. Total stored energy increased approximately 20% from nominally hydrogen plasmas to deuterium plasmas during auxiliary heating. Of this increase about half can be attributed to purely classical differences in the energy content of unthermalized beam ions. Kinetic measurements indicate a consistent but small increase in central electron temperature and total stored electron energy in deuterium relative to hydrogen plasmas, but no change in total ion stored energy. No significant differences in particle transport, momentum transport, and sawtooth behavior are observed. Overall, only a small improvement (~10%) in global energy confinement time of the thermal plasma is seen between operation in hydrogen and deuterium. (69 refs.)
Batha, S. H., F. M. Levinton, et al. (1996). "Confinement and the safety factor profile." Physics of Plasmas 3(4): 1348-55.
The conjecture that the safety factor profile, q(r), controls the improvement in tokamak plasmas from poor confinement in the low- (L) mode regime to improved confinement in the supershot regime has been tested in two experiments on the tokamak fusion test reactor (TFTR) [Plasma Phys. Controlled Nucl. Fusion Res. 1, 51 (1987)]. First, helium was puffed into the beam-heated phase of a supershot discharge, which induced a degradation from supershot to l-mode confinement in about 100 ms, far less than the current relaxation time. The q and shear profiles measured by a motional Stark effect polarimeter showed little change during the confinement degradation. Second, rapid current ramps in supershot plasmas altered the q profile, but were observed not to change significantly the energy confinement. Thus, enhanced confinement in supershot plasmas is not due to a particular q profile, which has enhanced stability or transport properties. The discharges making a continuous transition between supershot and l-mode confinement were also used to test the critical-electron-temperature-gradient transport model. It was found that this model could not reproduce the large changes in electron and ion temperature caused by the change in confinement. (40 refs.)
Bettenhausen, M. H. and J. E. Scharer (1996). "Simulation of parasitic edge ion Bernstein wave coupling during fast wave heating." Nuclear Fusion 36(7): 881-90.
Parasitic coupling to the ion Bernstein wave during fast wave heating is investigated using the WIGS computer code. The WIGS code models ion cyclotron range of frequencies (ICRF) antenna coupling to tokamak plasmas for phased coil antenna arrays in a recessed cavity. The antenna radiation resistance and the spectrum of the ICRF power coupled to the plasma are calculated. Parasitic coupling to the ion Bernstein wave is shown to be important for low parallel wavenumbers during minority ion heating with large gradients in the plasma density near the edge of the plasma. Results are presented for TFTR antenna and plasma parameters using experimental data. (17 refs.)
Black, D. C. and R. M. Mayo (1996). "High sensitivity, inductively coupled miniature magnetic probe array for detailed measurement of time varying magnetic field profiles in plasma flows." Review of Scientific Instruments 67(4): 1508-16.
A modified B circuit design has been implemented as part of a miniature magnetic probe array for the Coaxial Plasma Source experiment [R.M. Mayo et al., Plasma Sources Sci. Technol. 4, 47 (1995)] at the North Carolina State University. This facility is currently being used for the generation of energetic plasma flows to allow laboratory study of magnetogasdynamics with particular emphasis on the importance of the Hall effect [D.C. Black et al., Phys. Plasma 1, 3115 (1994)] and plasma microinstabilities [R.M. Mayo et al., Phys. Plasma 2, 337 (1995)] to plasma transport in coaxial plasma sources. The miniature magnetic probe array consists of ten spatially separated coils wound on an Acetal form of dimensions 2.38 cm by 0.32 cm by 0.32 cm. At five positions, with roughly 0.32 cm separation, two mutually perpendicular coils are wound to measure the magnetic field in the theta and z directions. The modification to the signal processing circuitry consists of the use of a step-up transformer to boost the probe signal prior to filtering and acquiring the signal at the data acquisition system. This additional means of amplifying the signal allows for reduction in the size of the probe, and thus helps minimize the perturbing effect of the magnetic probe on the plasma. An additional advantage of using a signal transformer is that it provides electrical isolation between the experiment and the data acquisition system. (9 refs.)
Bosch, H. S., V. Erckmann, et al. (1996). "Summary of the workshop on technological aspects of steady-state devices." Plasma Physics and Controlled Fusion 38(3): 415-449.
The major contributions to this workshop came from TORE-SUPRA, TPX, LHD and W7-X (two tokamaks, one heliotron and a stellarator). All four devices are of similar size and designed for a similar range of parameters and, in particular, they address the same target-steady-state plasma operation. TORE-SUPRA is a circular cross section tokamak, LHD is designed with optimized continuous coils and provides a helical divertor, TPX and W7-X incorporate strongly-shaped geometries to improve confinement and stability and optimize the bootstrap current. In TPX, the system is optimized for a large bootstrap current, in W7-X for basically no bootstrap current. Also the heating systems are designed for the specific purpose of steady-state operation. TORE-SUPRA is in operation, LHD is near completion; W7-X has recently (after the workshop) been approved; TPX is still in the approval phase. Superconducting coil material is NbTi for TORE-SUPRA, LHD and W7-X and Nb3Sn for TPX. The TF-coils of TORE-SUPRA (1.8 K, superfluid He) and the helical coil of LHD (4.2 K for the first part of experiments with 3 T) are bath-cooled. Cable-in-conduit conductors are used in the other SC coils. Largely different solutions are selected for the SC cable composition. Except for TPX, which has a double-null poloidal field divertor matched to the strong plasma shaping, most of the steady-state devices have a very flexible plasma edge configuration: islands or ergodic boundaries (W7-X, TORE-SUPRA, LHD), limiter (TORE-SUPRA), helical divertor (LHD, W7-X). The status of plasma equilibrium control, high heat flux materials and activation problems were also discussed. Power plant studies based on different steady-state confinement concepts were presented and confronted with a summary on the PULSAR study based on pulsed systems.
Chakrabarti, N. and P. K. Kaw (1996). "Velocity shear effect on Rayleigh-Taylor vortices in nonuniform magnetized plasmas." Physics of Plasmas 3(10): 3599-603.
In this paper, the effect of velocity shear on Rayleigh-Taylor vortices has been demonstrated. An inhomogeneous plasma is considered with a density profile such that the diamagnetic drift velocity V{sub n}=(cT{sub e}/eB)dn{sub 0}/dx is a constant and includes the effect of an ambient poloidal shear flow V{sub eq}(x)=V'{sub perpendicular to 0}(x-x{sub 0})y. The final equation describing the stationary Rayleigh-Taylor vortex is shown to have the structure of a nonlinear Poisson equation, where the nonlinearity arises essentially because of the velocity shear term. This equation has been solved numerically and it has been shown that qualitatively new two-dimensional monopole vortex solutions may be obtained in the appropriate parameter space. Therefore, a new important nonlinear effect related to equilibrium shear flow has been identified in the calculations of Rayleigh-Taylor vortices which results in monopole-like solutions in plasmas. (19 refs.)
Chaube, N. R. and K. K. Jain (1996). "Radial profile of plasma potential with various biased electrode ring configurations in a toroidal plasma." Physics of Plasmas 3(7): 2626-30.
An experimental study on behavior of radial profile of the floating potential with different biased electrode ring configurations has been carried out in a currentless magnetized toroidal plasma. Radial profile of the floating potential has been measured by biasing single ring of various sizes and two rings. It is observed that floating potential profile of a well shaped with controllable depth, hill-cum-well shaped, and almost flat positive potential can be obtained. Results on parameter dependence studies of floating potential on the bias voltage, magnetic field, and gas pressure are presented. (16 refs.)
Darrow, D. S., S. J. Zweben, et al. (1996). "Alpha particle losses from Tokamak Fusion Test Reactor deuterium-tritium plasmas." Physics of Plasmas
37th Annual Meeting of the Division-of-Plasma-Physics of the American-Physical-Society 3(5): 1875-1880.
Because alpha particle losses can have a significant influence on tokamak reactor viability, the loss of deuterium-tritium alpha particles from the tokamak fusion test reactor (TFTR) [K.M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] has been measured under a wide range of conditions. In TFTR, first orbit loss and stochastic toroidal field ripple diffusion are always present. Other losses can arise due to magnetohydrodynamic instabilities or due to waves in the ion cyclotron range of frequencies. No alpha particle losses have yet been seen due to collective instabilities driven by alphas. Ion Bernstein waves can drive large losses of fast ions from TFTR, and details of those losses support one element of the alpha energy channeling scenario. (40 refs.)
Fredrickson, E. D., K. M. McGuire, et al. (1996). "Ballooning instability precursors to high beta disruptions on the Tokamak Fusion Test Reactor." Physics of Plasmas 3(7): 2620-5.
Toroidally localized ballooning modes have been found as precursors to high beta disruptions in many regimes on the Tokamak Fusion Test Reactor (TFTR) [D. Meade et al., proceedings of the international conference on plasma physics and controlled nuclear fusion, Washington, DC, 1990 (International Atomic Energy Agency, Vienna, 1991), vol. 1, pp. 9-24]. Lower frequency, global magnetohydrodynamic (MHD) activity, typically an ideal n=1 kink mode, causes the toroidal localization. Larger-amplitude n=1 modes result in stronger toroidal localization of the ballooning modes. The modes are typically localized to a region spanning about 90 degrees -120 degrees in the toroidal direction. (15 refs.)
Fu, G. Y., C. Z. Cheng, et al. (1996). "Analysis of alpha particle-driven toroidal Alfven eigenmodes in Tokamak Fusion Test Reactor deuterium-tritium experiments." Physics of Plasmas 3(11): 4036-45.
The toroidal Alfven eigenmodes (TAE) are calculated to be stable in the presently obtained deuterium-tritium plasmas in the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Nucl. Fusion Res. 26, 11 (1984)]. However, the core localized TAE mode can exist and is less stable than the global TAE modes. The beam ion Landau damping and the radiative damping are the two main stabilizing mechanisms in the present calculation. In future deuterium-tritium experiments, the alpha-driven TAE modes are predicted to occur with a weakly reversed shear profile. (37 refs.)
Fundamenski, W. R. and A. A. Harms (1996). "Evolution and status of D-He-3 fusion: A critical review." Fusion Technology 29(3): 313-349.
Advanced fuels for nuclear fusion - of which deuterium and He-3 mixture is the leading candidate - could reduce tritium inventory, neutron fluence, structural damage, and activation in future reactors as well as allow for direct energy conversion. The feasibility of D-He-3 fusion is assessed based on recent developments in the areas of fuel resources, fusion and plasma physics, magnetic and inertial reactors, space propulsion, reactor safety, and waste disposal. It appears that D-He-3 fusion is not well suited to the conventional tokamak design (beta similar to 10%) because of excessive synchrotron loss and closed field topology. High-beta and/or non-Maxwellian plasma configurations are promising but at present lack a sufficient experimental database to predict reactor-relevant behavior. Space propulsion appears to be a most advantageous application of D-He-3 fusion.
Galambos, J. P., M. A. Bohnet, et al. (1996). "Magnetic field measurements using the transient internal probe (TIP)." Review of Scientific Instruments 67(2): 469-72.
The transient internal probe (TIP) is a novel diagnostic technique for measuring magnetic fields in hot plasmas. The concept involves shooting a diamond clad magneto-optic probe through the plasma at high velocity allowing measurement of the local magnetic field before ablation occurs. Magnetic field measurements are obtained by illuminating the probe with an argon laser and measuring the amount of Faraday rotation in the reflected light. The diagnostic was tested by measuring a permanent magnetic field inside a vacuum chamber with a probe traveling at 2 km/s using an unclad probe. The purpose of this experiment was to demonstrate the capability of the TIP diagnostic and to verify compatibility with plasma vacuum requirements. Magnetic field resolutions of 20 G and 1 cm spatial resolution were achieved. The response time of the detection system is 10 MHz. Introduction of a helium muzzle gas into the plasma chamber was limited to less than 0.4 Torr l. (16 refs.)
Goldston, R. J. (1996). "Reactor compatibility of the H-mode." Plasma Physics and Controlled Fusion
5th IAEA Technical Committee Meeting/US-Japan Workshop on H-Mode Physics 38(8): 1161-72.
H-mode operation presents challenges for ITER and for reactor application in the areas of performance, accessibility, sustainability and divertor compatibility. Some issues are more severe for ITER, while others are more problematic for an economic fusion reactor producing 3 MW m{sup -2} of neutron flux to its walls. The problems that are more severe for ITER include the power requirements for the H-mode, and confinement uncertainties. Problems which are more severe for a reactor include capability for high beta * operation, the role of the Greenwald density limit, and divertor compatibility. Within the present ITER concept, it would be very helpful to have a well defined upgrade path for the auxiliary heating systems, including options for strong profile control. The world fusion research programme should maintain a major focus on improving the reactor compatibility of the H-mode, with an eye to the special advantages of H-mode plasmas with strong shaping. (14 refs.)
Hawryluk, R. J., C. W. Barnes, et al. (1996). "Review of D-T results from TFTR." Fusion Technology
12th Topical Meeting on the Technology of Fusion Energy 30(3): 648-659.
During the D-T campaign on TFTR, safe and successful operation has been demonstrated with tritium fuel enabling a broad range of physics studies. Transport studies have focused on the formation of internal transport barriers in the enhanced reversed shear regime. Current profile modification has been employed to study MHD stability in both reversed shear and high l{sub i} discharges. Several important alpha physics topics have been studied including the confinement and loss of alpha particles in both quiescent and MHD active discharges and the effect of alpha-particle heating and alpha-particle destabilization of TAE modes. Plans for future experiments are being discussed. (56 refs.)
Hoek, M., T. Nishitani, et al. (1996). "Triton burnup measurements by neutron activation at JT-60U." Nuclear Instruments & Methods in Physics Research, Section A (Accelerators, Spectrometers, Detectors and Associated Equipment) 368(3): 804-14.
This paper describes measurements on triton burnup in a deuterium plasma by the detection of the 2.5 MeV neutrons (from DD fusion) and the 14 MeV neutrons (from DT fusion). The 2.5 MeV neutrons have been measured by fission chambers and activation of indium foils while the 14 MeV neutrons have been detected by activation of silicon, aluminum, and copper foils. The measured yields of the 2.5 MeV neutrons utilizing In foils are 20-40% higher than the yields obtained from fission chambers depending on what calibration factors are used. The deviation decreases with the plasma major radius (or increasing plasma volume). When the triton burnup is measured by utilizing neutron threshold reactions (E{sub n}>2.5 MeV) and In foils, then systematic errors in the calibration factors cancel and the maximum deviation between the measured triton burnup for different calibration factors is reduced to 5%. The measurements indicate that triton burnup increases with the 14 MeV neutron yield, indicating that the relative yield of 14 MeV neutrons increases depending on the time duration of the deuterium neutral beam injection (NBI). Furthermore, the triton burnup decreases with an increased plasma major radius, indicating increased triton ripple losses, and increases with plasma current, indicating reduced banana orbit losses. (14 refs.)
Hoffman, A. L. (1996). "An ideal compact fusion reactor based on a field-reversed configuration." Fusion Technology
12th Topical Meeting on the Technology of Fusion Energy 30(3): 1367-1371.
Field-reversed configurations (FRC) have been recognized as possessing almost ideal fusion reactor characteristics from the point of view of engineering simplicity and maintainability. The external geometry is cylindrical while the internal magnetic field configuration is toroidal, allowing for both a simple magnetic confinement design and the possibility of good plasma confinement. FRCs are unique among all toroidal confinement concepts in not possessing any significant toroidal field. This necessitates a very high plasma beta, which provides for extreme compactness, but imposes very non-standard requirements for basic stability. Recent experimental results have gone far toward demonstrating this stability, and new experiments are underway toward developing other aspects along the FRC reactor development path. If successful, these experiments could represent a breakthrough in fusion reactor attractiveness.
Jain, K. K. (1996). "Poloidal plasma rotation and its effect on fluctuations in a toroidal plasma." Nuclear Fusion 36(12): 1661-8.
An experimental study on poloidal plasma rotation induced by electrode biasing and its effect on fluctuations is carried out in a purely toroidal plasma having no poloidal magnetic field. Poloidal plasma rotation and, concomitantly, significant reduction in the fluctuation level are observed. The radial profiles of poloidal flows and fluctuations are measured and compared. The dependence of the poloidal plasma rotation on the bias voltage and the toroidal magnetic field is presented. It is found that the large poloidal plasma rotation gives rise to a transition of the plasma state from a turbulent to a coherent mode. (33 refs.)
Janos, A. C., E. D. Fredrickson, et al. (1996). "Characterization of alpha particle loss during disruptions in TFTR during deuterium-tritium operation." Nuclear Fusion 36(4): 475-94.
The timing of alpha losses, with respect to the various phases of a disruption, and the impact location of the losses are characterized during high fusion power operation of TFTR with deuterium and tritium fuels. Characterization of alpha losses is important for the design of future fusion devices such as ITER. In addition, characterization of the alpha losses with respect to the disruption evolution helps in the understanding of the dynamics of the disruption process and related MHD events such as sawtooth crashes. Disruptions are characterized as having several standard phases, applicable to most disruptions in all tokamaks: precursor, thermal quench(es) and current quench. Most of the losses are observed to occur during the thermal quench phase(s). In high beta disruptions, alpha losses start abruptly during the growth of MHD precursors, just before the onset of the thermal quench. The initial burst of losses, lasting as little as 100 mu s, can release a significant fraction (one third) of the total disruption induced losses during the thermal quench. An inventory of alpha particles suggests that the alpha loss distribution during disruptions might be quite different from that expected during non-disruptive discharges. There are no obvious differences between fast fusion product losses during deuterium-deuterium (DD) and deuterium-tritium (DT) disruptions, aside from the large alpha loss component in DT discharges. (29 refs.)
Juliano, D. R., D. N. Ruzic, et al. (1996). "Sensitivity of neutral throughput to geometry and plasma position to the Tokamak Physics Experiment divertor region." Fusion Technology 29(2): 269-76.
Sufficient neutral atom and molecular throughput is essential for the steady-state operation of the proposed Tokamak Physics Experiment tokamak. To predict the throughput, the B2 edge-plasma fluid code and the DEGAS Monte Carlo neutral transport code were coupled globally. For the day 1 low-power (17.5-MW) operation condition, the recycling coefficient for both codes matched at 0.985, implying that for every 1000 ions striking the divertor plate, 15 are ultimately removed down the pump duct. The neutral molecular density was 2.52+or-0.15*10{sup 19}/m{sup 3}, giving a throughput of 92.6+or-5.6 Torr.l/s. Varying the scrape-off length for the plasma extending into the gap between the baffle and the plate from 0 to 2 on decreased the throughput by a factor of >2. Moving the strike point away from the gap at first increases the throughput by lessening the pumping efficiency of the plasma in the gap. As the plasma is moved even farther array, the throughput drops due to a lack of source term for neutrals entering the pumped region. Illustrating the importance of moving the source term, moving the strike point away from the gap but retaining the original plasma in the gap lowers the throughput by a factor of 10. Altering the curvature of the baffle has little effect on the neutral solution. (7 refs.)
Kamada, Y. (1996). "Characteristics of and issues regarding combined H-modes." Plasma Physics and Controlled Fusion
5th IAEA Technical Committee Meeting/US-Japan Workshop on H-Mode Physics 38(8): 1173-88.
Owing to the recent progress of `profile controls' utilizing the freedom of profiles of plasma parameters, various types of combined (advanced) H-modes have been obtained; PEP H-mode, high- beta {sub p} H-mode, high-T{sub i} H-mode, high-l{sub i} H-mode, reversed shear H-mode, CH-mode, VH-mode, etc. These combined H-modes with high potential for confinement and stability are characterized by improved transport in the core region in addition to the edge or deep inward penetration of the edge confinement pedestal. Such additional improvements seem to be related to E*B flow shear, magnetic shear, safety factor, density profile or T{sub i}/T{sub e}. The improved confinement region can propagate from the core to the edge or from the edge to the core, or appear almost simultaneously over the whole radius. However, there are some critical issues to be solved concerning reactor conditions, that is, under electron heating at high edge density with small impurity accumulation in the steady-state. Most of the NB heated combined H-modes have been obtained with ion heating, where ion thermal diffusivity and particle diffusivity can be reduced to the neoclassical level. However electron thermal diffusivity has so far not been clearly reduced except in the H-mode and in the VH-mode. Therefore, improvement of electron transport is one of the main issues. The second issue is that some improved modes are accompanied by strong density peaking which may result in large impurity accumulation. The third issue is the difficulty in achieving the improved modes at a high edge density with the high particle recycling essential for a dense and cold divertor. The fourth issue is the stability at high B in the steady state. It is not certain that pressure and current profiles including the bootstrap current can be sustained stably. The requirement of ELMs for heat and particle exhaust is also an important issue in stability design as concerns the extent of the second regime access for the high-n ballooning mode. (62 refs.)
Kruger, S. E., R. V. Budny, et al. (1996). "Investigations of tritium recycling in TFTR using the DT neutron rate." Nuclear Fusion 36(8): 1053-6.
During deuterium-only neutral beam injected discharges, tritium from earlier deuterium-tritium (DT) discharges is released from the vessel limiters and walls to cause a DT neutron count rate comparable to the DD neutron count rate. A measure of the tritium density in the plasma based on neutron rate measurements is defined and used to determine which parameters influence tritium influx to the plasma core. The tritium density is observed to decrease in a sequence of deuterium-only supershots and to depend on the amount of tritium injected in prior DT shots and on the amount of tritium present in the limiter. A weak correlation is also observed with the plasma current, but not with beam power, hydrogen influx, carbon influx, visible bremsstrahlung, lithium pellet injection, blooms or disruptions. (7 refs.)
Kugel, H. W., G. Ascione, et al. (1996). "Status of tokamak fusion test reactor neutron activation." Fusion Technology
12th Topical Meeting on the Technology of Fusion Energy 30(3): 1065-1068.
Measurements have been made following TFTR D-T campaigns to characterize the behavior of D-T fusion reactor neutron activation using Ionization Chamber, Geiger Mueller, and Ge detector gamma-ray spectroscopy measurements. The results exhibit decay rates characteristic of the materials and geometries of the Test Cell hardware, and allow extrapolation to higher fusion power yields. The results can be used for benchmarking D-T fusion reactor activation simulations for accurate determinations of low activation long-lived cooling.
Kurnaev, V. A. (1996). "Plasma-surface interaction." Izvestiya Akademii Nauk Seriya Fizicheskaya
12th Conference on Interaction of Ions with Surfaces 60(7): 82-99.
Mansfield, D. K., K. W. Hill, et al. (1996). "Enhancement of Tokamak Fusion Test Reactor performance by lithium conditioning." Physics of Plasmas
37th Annual Meeting of the Division-of-Plasma-Physics of the American-Physical-Society 3(5): 1892-1897.
Wall conditioning in the Tokamak Fusion Test Reactor (TFTR) [K.M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] by injection of lithium pellets into the plasma has resulted in large improvements in deuterium-tritium fusion power production (up to 10.7 MW), the Lawson triple product (up to 10{sup 21} m{sup -3} s keV), and energy confinement time (up to 330 ms). The maximum plasma current for access to high-performance supershots has been increased from 1.9 to 2.7 MA, leading to stable operation at plasma stored energy values greater than 5 MJ. The amount of lithium on the limiter and the effectiveness of its action are maximized through (1) distributing the Li over the limiter surface by injection of four Li pellets into ohmic plasmas of increasing major and minor radius, and (2) injection of four Li pellets into the ohmic phase of supershot discharges before neutral-beam heating is begun. (9 refs.)
McClements, K. G., R. O. Dendy, et al. (1996). "Interpretation of ion cyclotron emission from sub-Alfvenic fusion products in the Tokamak Fusion Test Reactor." Physics of Plasmas 3(2): 543-53.
Ion cyclotron emission (ICE) has been observed during neutral beam-heated supershots in the Tokamak Fusion Test Reactor (TFTR) [Phys. Rev. Lett. 72, 3526 (1994)] deuterium-tritium campaign at fusion product cyclotron harmonics. The emission originates from the outer midplane edge plasma, where fusion products initially have an anisotropic velocity distribution, sharply peaked at a sub-Alfvenic speed. It is shown that the magnetoacoustic cyclotron instability, resulting in the generation of obliquely propagating fast Alfven waves at fusion product cyclotron harmonics, can occur under such conditions. The time evolution of the growth rate closely follows that of the observed ICE amplitude. Instability is suppressed if the fusion products undergo a moderate degree of thermalization, or are isotropic. In contrast, the super-Alfvenic fusion products present in the outer midplane of the Joint European Torus (JET) [Nucl. Fusion, 33, 1365 (1993)] can drive the instability if they are isotropic or have a broad speed distribution. This may help to account for the observation that fusion product-driven ICE in JET persists for longer than fusion product-driven ICE in TFTR supershots. (33 refs.)
Meade, D. M. (1996). "Recent progress on the Tokamak Fusion Test Reactor." Journal of Fusion Energy 15(3/4): 163-7.
The deuterium-tritium (D-T) experiments on the Tokamak Fusion Test Reactor (TFTR) have yielded unique information on the confinement, heating and alpha particle physics of reactor scale D-T plasmas as well as the first experience with tritium handling and D-T neutron activation in an experimental environment. The D-T plasmas produced and studied in TFTR have peak fusion power of 10.7 MW with central fusion power densities of 2.8 MW m{sup -3} which is similar to the 1.7 MW m{sup -3} fusion power densities projected for 1500 MW operation of the International Thermonuclear Experimental Reactor (ITER). Detailed alpha particle measurements have confirmed alpha confinement and heating of the D-T plasma by alpha particles as expected. Reversed shear, high l{sub i} and internal barrier advanced tokamak operating modes have been produced in TFTR which have the potential to double the fusion power to ~20 MW which would also allow the study of alpha particle effects under conditions very similar to those projected for ITER. TFTR is also investigating two new innovations, alpha channeling and controlled transport barriers, which have the potential to significantly improve the standard advanced tokamak. (13 refs.)
Medley, S. S., D. K. Mansfield, et al. (1996). "Design and operation of the pellet charge exchange diagnostic for measurement of energetic confined alpha particles and tritons on the Tokamak Fusion Test Reactor." Review of Scientific Instruments 67(9): 3122-35.
Radially resolved energy and density distributions of the confined alpha particles in D-T experiments on the Tokamak Fusion Test Reactor (TFTR) are being measured with the pellet charge exchange (PCX) diagnostic. Other energetic ion species can be detected as well, such as tritons produced in D-D plasmas and H, He{sup 3}, or tritium rf-driven minority ion tails. The ablation cloud formed by injected low-Z impurity pellets provides the neutralization target for this active charge exchange technique. Because the cloud neutralization efficiency is uncertain, the PCX diagnostic is not absolutely calibrated so only relative density profiles are obtained. A mass and energy resolving E||B neutral particle analyzer (NPA) is used which has eight energy channels covering the energy range of 0.3-3.7 MeV for alpha particles with energy resolution ranging from 5.8% to 11.3% and a spatial resolution of ~5 cm. The PCX diagnostic views deeply trapped ions in a narrow pitch angle range around a mean value of upsilon ||/ upsilon =-0.048+or-10{sup -3}. For D-T operation, the NPA was shielded by a polyethylene-lead enclosure providing 100* attenuation of ambient gamma radiation and 14 MeV neutrons. The PCX diagnostic technique and its application on TFTR are described in detail. (23 refs.)
Medley, S. S., R. V. Budny, et al. (1996). "Measurements of confined alphas and tritons in the MHD quiescent core of TFTR plasmas using the pellet charge exchange diagnostic." Plasma Physics and Controlled Fusion 38(10): 1779-89.
The energy distributions and radial density profiles of the fast confined trapped alpha particles in DT experiments on TFTR are being measured in the energy range 0.5-3.5 MeV using the pellet charge exchange (PCX) diagnostic. A brief description of the measurement technique which involves active neutral particle analysis using the ablation cloud surrounding an injected impurity pellet as the neutralizer is presented. This paper focuses on alpha and triton measurements in the core of MHD quiescent TFTR discharges where the expected classical slowing-down and pitch angle scattering effects are not complicated by stochastic ripple diffusion and sawtooth activity. In particular, the first measurement of the alpha slowing-down distribution up to the birth energy, obtained using boron pellet injection, is presented. The measurements are compared with predictions using either the TRANSP Monte Carlo code and/or a Fokker-Planck Post-TRANSP processor code, which assumes that the alphas and tritons are well confined and slow down classically. Both the shape of the measured alpha and triton energy distributions and their density ratios are in good agreement with the code calculations. We can conclude that the PCX measurements are consistent with classical thermalization of the fusion-generated alphas and tritons. (27 refs.)
Messiaen, A. M., J. Ongena, et al. (1996). "High confinement and high density with stationary plasma energy and strong edge radiation in the TEXTOR-94 tokamak." Physical Review Letters 77(12): 2487-90.
Stationary high energy confinement is observed on TEXTOR-94 for times limited only by the flux swing of the transformer using strong edge radiation cooling. Necessary tools are the feedback control of the radiated power and of the plasma energy content. At the highest densities obtained (up to 1.2 times the Greenwald limit), energy confinement exceeds the edge-localized-mode-free H-mode scaling ITERH93-P by more than 20%. beta limits of TEXTOR-94 are reached with f{sub H89}/q{sub a} approximately=0.6. No detrimental effect of the seeded impurity is seen. These high confinement discharges meet many conditions necessary for a fusion reactor regime. (16 refs.)
Mueller, D., W. Blanchard, et al. (1996). "Removal of tritium from TFTR." Fusion Technology
12th Topical Meeting on the Technology of Fusion Energy 30(3): 840-844.
Operation of the Tokamak Fusion Test Reactor (TFTR) with a mixture of deuterium and tritium fueling has permitted the opportunity to measure the retention of tritium in the graphite limiter and other internal hardware. The use of discharge cleaning techniques and venting to remove the tritium was investigated. The tritium was introduced into TFTR by neutral beam injection and by gas puffing. The graphite limiter is subject to erosion and codeposition. While short term retention was high, the retention averaged over the 1993-1995 D-T campaign was 52%+/-15%. The tritium removal techniques resulted in lowering the in-vessel inventory from 16.4 kCi at the end of 1995 operation to 7.2 kCi at the start of the 1996 experimental program. (14 refs.)
Najmabadi, F., J. Drake, et al. (1996). "Alternative concepts: A report to the Fusion Energy Sciences Advisory Committee." Journal of Fusion Energy 15(3-4): 249-280.
Nishitani, T., M. Hoek, et al. (1996). "Triton burn-up study in JT-60U." Plasma Physics and Controlled Fusion 38(3): 355-64.
Time-resolved triton burn-up measurements have been performed using a new type of 14 MeV neutron detector based on scintillating fibres. The shot-integrated triton burn-up ratio was measured to be in the range 0.3-2.0%. Time histories of 14 MeV emission after NB turn-off have been analysed based on the classical slowing-down theory. Assuming the loss of fast tritons can be represented as a diffusivity, values increasing with increasing toroidal ripple were determined between 0.05 and 0.15 m{sup 2} s{sup -1}, from the modelling of the time histories of the 14 MeV emission after the NB turn-off. (15 refs.)
Porcelli, F., D. Boucher, et al. (1996). "Model for the sawtooth period and amplitude." Plasma Physics and Controlled Fusion
22nd European-Physical-Society Conference on Controlled Fusion and Plasma Physics 38(12): 2163-86.
A model for sawtooth oscillations in tokamak experiments is outlined. A threshold criterion for the onset of internal kink modes and a prescription for the relaxed profiles immediately after the sawtooth crash have been implemented in a transport code that evolves the relevant plasma parameters. In this paper, applications of this model to the prediction of the sawtooth period and amplitude in projected ITER discharges are discussed. It is found that sawteeth can be stabilized transiently by the fusion alpha particles in ITER for periods that are long on the energy confinement timescale ( tau {sub E} approximately=5 s). The sawtooth period depends on the amount of reconnected flux at the preceding sawtooth crash. When Kadomtsev's full reconnection model is used, the period can exceed 100 s. The sawtooth mixing radius following long duration sawtooth ramps can easily exceed half the plasma minor radius, raising questions about the desirability of transient sawtooth suppression. (46 refs.)
Pozzo, M. and M. Tessarotto (1996). "Construction of the weakly-relativistic Fokker-Planck kinetic equation in the Darwin approximation." Physics of Plasmas 3(6): 2255-64.
The explicit form of the weakly-relativistic Vlasov and the Fokker-Planck collision operators for a multispecies plasma are evaluated by using a systematic expansion in beta = upsilon _/c and retaining corrections up to the second order properties of these operators are investigated. (17 refs.)
Rogister, A. L. (1996). "Anomalous transport models for fusion plasmas: review and perspectives." Fusion Technology 29(2T): 170-82.
The basic physical concepts underlying the theories of anomalous transport in magnetic confinement devices are reviewed. Anomalous transport is a consequence of electric and/or magnetic fluctuations driven by various linear and/or nonlinear instability mechanisms. The latter saturate by inducing a relaxation of the profiles towards a marginally stable state or/and by nonlinear coupling of the various modes. Specific theoretical models are described, together with their successes and drawbacks in the light of observed characteristics of plasma confinement, a non exhaustive list of which is given. A rough estimate of the nuclear heating power required to balance the anomalous losses in the International Tokamak Experimental reactor (ITER) is calculated on the basis of the electrostatic drift wave instability model. (58 refs.)
Skinner, C. H., W. Blanchard, et al. (1996). "Measurements of tritium retention and removal on the Tokamak Fusion Test Reactor." Journal of Vacuum Science & Technology A (Vacuum, Surfaces, and Films) 14(6): 3267-74.
Recent experiments on the Tokamak Fusion Test Reactor have afforded an opportunity to measure the retention of tritium in a graphite limiter that is subject to erosion, codeposition, and high neutron flux. The tritium was injected by both gas puff and neutral beams. The isotopic mix of hydrogenic recycling was measured spectroscopically and the tritium fraction T/(H+D+T) transiently increased to as high as 75%. Some tritium was pumped out during the experimental run and some removed in a subsequent campaign using various clean-up techniques. While the short term retention of tritium was high, various conditioning techniques were successful in removing approximately=8000 Ci and restoring the tritium inventory to a level well below the administrative limit. (33 refs.)
Stacey, W. M. (1996). "Radial thermal stability of the radiative mantle." Physics of Plasmas 3(3): 1012-21.
Sufficient conditions for the thermal stability against radial excursions of a cylindrical equilibrium temperature distribution in a plasma with a strongly radiative edge, or mantle, are established in terms of the minimum allowable ratio of the core heating power density to the edge radiation power density. A positive dependence of the thermal conductivity upon temperature is found to be stabilizing, and a divertor separatrix boundary condition for the temperature is found to be destabilizing relative to a fixed separatrix temperature condition. The stability conditions are consistent with the conditions extant in a stable radiative mantle experiment and in a stable radiative mantle simulation. Expressions are developed for a radiation edge density limit and for the maximum fraction of the plasma heating power which can be radiated from the mantle. (20 refs.)
Stotler, D. P., C. H. Skinner, et al. (1996). "Modeling of neutral hydrogen velocities in the Tokamak Fusion Test Reactor." Physics of Plasmas 3(11): 4084-94.
Monte Carlo neutral transport simulations of hydrogen velocities in the Tokamak Fusion Test Reactor (TFTR) [K.M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] are compared with experiment using the Doppler-broadened Balmer- alpha spectral line profile. Good agreement is obtained under a range of conditions, validating the treatment of charge exchange, molecular dissociation, surface reflection, and sputtering in the neutral gas code DEGAS [D. Heifetz et al., J. Comput. Phys. 46, 309 (1982)]. A residual deficiency of 10-100 eV neutrals in most of the simulations indicates that further study of the energetics of H{sub 2}{sup +} dissociation for electron energies in excess of 100 eV is needed. (51 refs.)
Strachan, J. D., J. S. McCauley, et al. (1996). "Triton burnup profile measurements." Nuclear Fusion 36(9): 1189-200.
The TFTR helium proportional counters measured the 14 MeV neutron emission profiles from the 1 MeV tritons produced by d(d,p)t fusion reactions in deuterium plasmas. The magnitude and profile of the 1 MeV triton burnup indicate that there is no energetic triton transport (D<or=0.1 m{sup 2}/s) as the tritons slow down from 1 MeV to ~0.09 MeV. (24 refs.)
Stratton, B. C., R. J. Fonck, et al. (1996). "Observation of sawtooth redistribution of non-thermal, confined alpha particles in TFTR DT discharges." Nuclear Fusion 36(11): 1586-90.
Radial profiles of the density of confined alpha particles with energies in the 0.15 to 0.6 MeV range are spectroscopically observed before and after a sawtooth crash in a TFTR deuterium-tritium plasma. A large drop in the core alpha density is seen, indicating expulsion of alphas from the core to the plasma periphery. The measured changes in the alpha density profiles are consistent with predictions based on the Kolesnichenko sawtooth model in this case. (18 refs.)
Sugai, H., M. Ohori, et al. (1996). "Lithium wall conditioning for fuel and impurity control." Vacuum
13th International Vacuum Congress/9th International Conference on Solid Surfaces (IVC-13/ICSS-9) 47(6/8): 981-4.
Small-scale laboratory experiments on wall conditioning by thin lithium layer deposition are carried out. Suppression of carbon impurities by lithium deposition onto graphite walls are clearly demonstrated by a glow discharge in 1% oxygen in helium. Strong gettering effects of clean lithium surfaces on oxygen and hydrogen atoms are found, supporting recent findings in fusion devices. The maximum number of H atoms pumped by the lithium layer at room temperature is approximately equal to the number of Li atoms in the vessel Thus, the effective solubility of hydrogen in lithium is as high as H/(Li+H) approximately=50% which is orders of magnitude larger than the values in the thermal equilibrium state. Furthermore, most of the hydrogen pumped by the lithium layer are desorbed as H{sub 2} molecules at relatively low temperatures such as 200 degrees C. This thermal desorption study suggests a very slow formation of lithium hydride whose thermal decomposition takes place at approximately=700 degrees C. (6 refs.)
Tasso, H. and W. Horton (1996). "Statistical properties of the drift wave fluctuations." Plasma Physics Reports 22(9): 701-713.
The nature of turbulence in macroscopically confined plasmas is reviewed and contrasted with turbulence in hydrodynamics. The statistical properties of the fluctuations are analyzed both from the Gibb's distribution for the soliton gas model of the electrostatic field and for the Hamiltonian held theory equilibrium statistics. For that purpose, nonlinear drift wave equations are derived from the two-fluid theory. From the reduced nonlinear drift wave equations, we construct continuous plasma models with simple Hamiltonians, which allow canonical distributions to be defined explicitly. Partition functions and correlation functions can be calculated analytically in the one-dimensional case as functional integral averages over canonical distributions. The relation of the k-space fluctuation spectrum obtained from canonical distributions with those inferred from the electromagnetic scattering experiments is given. The open problem of saturation levels of fluctuations is discussed in the conclusions.
Taylor, G., J. D. Strachan, et al. (1996). "Fusion heating in a deuterium-tritium tokamak plasma." Physical Review Letters 76(15): 2722-5.
Evidence for fusion heating in the core of a deuterium-tritium (D-T) tokamak plasma is reported for the first time. Electron temperature profile data were analyzed for differences between D-T, D, and T plasmas in the Tokamak Fusion Test Reactor. Data from D and D-T plasmas with similar plasma parameters were averaged to minimize isotopic effects. The electron temperature in D-T plasmas was systematically higher than in D or T plasmas. The temperature difference between D-T and D plasmas with similar confinement times is consistent with alpha-particle heating of electrons. (17 refs.)
Wang, A. K. and X. M. Qiu (1996). "Model for relaxation in reversed-field pinch plasmas." Physics of Plasmas 3(6): 2316-30.
In this article, a model for anomalous ion heating, a dynamo current-sustained edge toroidal field, and a sawtooth oscillation during the relaxation in the reversed-field pinch (RFP) plasma is presented. The dynamo ( alpha ), the turbulent resistivity ( beta ) and viscosity ( chi ), dependent on the magnetohydrodynamics (MHD) fluctuations, are incorporated into the model. Turbulent viscous dissipation of the fluctuation energy is proposed as the mechanism of the anomalous ion heating. This is a straightforward corollary of the turbulent viscosity heating of ions in that the temperature of the heavier ions is higher than that of the lighter ions and that the ion temperature increases with the MHD fluctuation level. Correspondingly, the turbulent resistivity heats electrons anomalously. It is shown that the dynamo, current, generated by the back-transfer of fluctuating magnetic field helicity to mean magnetic field, sustains the RFP magnetic configuration. In the edge the total current density is approximately equal to the dynamo current density, while at the core the dynamo current opposes the applied electric-field-driven current, flattening the current profile. Provided the alpha dynamo has a periodic behavior in time, the physical quantities of the RFP plasma have a sawtooth time dependence. The local poloidal current density in the edge increases during the sawtooth crash and peaks at the end of the crash, as do the ion and electron temperatures. In contrast, the toroidal current density at the core decreases during the crash and arrives at its minimum at the end of the crash. Qualitatively, the conclusions drawn from the present model are in good agreement with many of the experimental results [Scime et al., Phys. Rev. Lett. 68, 2165 (1992); Ji et al., ibid. 73, 668 (1994)] and the numerical simulations. (90 refs.)
Werley, K. A., J. N. DiMarco, et al. (1996). "Energy confinement and future reversed field pinches." Nuclear Fusion 36(5): 629-42.
Energy confinement within the reversed field pinch (RFP) is governed by three plasma regions: a poorly confined plasma core characterized by parallel (radial) transport and flux surfaces destroyed by m=1 tearing mode activity associated with dynamo relaxation; a medium confined edge limited by ideal pressure gradient driven modes; and a good confinement region located near the reversal layer. The good confinement region determines the global confinement characteristics of the RFP, and, if limited by resistive interchange modes, would be consistent with the Connor-Taylor scaling that has provided a good fit to international RFP results. After establishing a two parameter fit of confinement scaling to the RFP database, the scaling relation is used to project the physical characteristics of, and costs associated with, next step and ignition experiments for ohmically heated RFPs. The RFP projects to smaller and less expensive machines than the tokamak with comparable performance. (40 refs.)
Wong, K. L., G. L. Schmidt, et al. (1996). "First evidence of collective alpha particle effect on toroidal Alfven eigenmodes in the TFTR D-T experiment." Physical Review Letters 76(13): 2286-9.
The alpha particle effect on the excitation of toroidal Alfven eigenmodes (TAE) was investigated in deuterium-tritium (D-T) plasmas in the Tokamak Fusion Test Reactor. RF power was used to position the plasma near the instability threshold, and the alpha particle effect was inferred from the reduction of RF power threshold for TAE instability in D-T plasmas. Initial calculations indicate that the alpha particles contribute 10%-30% of the total drive in a D-T plasma with 3 MW of peak fusion power. (18 refs.)
Wurden, G. A. and D. O. Whiteson (1996). "High-speed plasma imaging: a lightning bolt." IEEE Transactions on Plasma Science 24(1): 83-4.
Using a gated intensified digital Kodak Ektapro camera system, we captured a lightning bolt at 1000 frames per second, with 100- mu s exposure time on each consecutive frame. As a thunder storm approached while darkness descended (7:50 pm) on July 21, 1994, we photographed lightning bolts with an f22 105-mm lens and 100% gain on the intensified camera. This 15-frame sequence shows a cloud to ground stroke at a distance of about 1.5 km, which has a series of stepped leaders propagating downwards, followed by the upward-propagating main return stroke. (3 refs.)
Yagi, Y., Y. Hirano, et al. (1996). "Measurement of fast electron velocity distribution function in a reversed field pinch plasma." Japanese Journal of Applied Physics, Part 1 (Regular Papers, Short Notes & Review Papers) 35(7): 4064-72.
Fast electrons in the edge region of the reversed field pinch (RFP) device, TPE-1RM20 are studied using an electron energy analyzer (EEA) in the energy range below 1 keV. Results show that the fast electron velocity distribution function can be approximated either by a drift Maxwellian distribution function or two asymmetric half-Maxwellian distribution functions. The fast electron temperature is comparable to or slightly less than the central electron temperature depending on the fitting functions. The temperature obtained using the EEA is more than one order of magnitude lower than the characteristic fast electron energy estimated from soft X-ray emissions from the inserted materials at 2-8 keV. These observations indicate that fast electrons mainly have a Maxwellian distribution function with a temperature comparable to the central electron temperature and that the distribution function may have a higher energy tail in the keV region. (20 refs.)
Zweben, S. J., R. V. Budny, et al. (1996). "Search for alpha driven TAEs at lowered ion temperature in TFTR DT discharges." Nuclear Fusion 36(8): 987-1008.
An experiment was performed in TFTR DT plasmas in an attempt to destabilize the alpha particle driven toroidicity-induced Alfven eigenmodes (TAEs) by reducing their thermal ion Landau damping. The thermal ion Landau damping was reduced by transiently lowering the ion temperature using either helium (He) gas puffs or deuterium (D) or lithium (Li) pellet injection during neutral beam injection (NBI) into DT supershots. The ion temperature was successfully lowered from T{sub i}(0) approximately=20 keV to T{sub i}(0) approximately=10 keV in about 0.2 s; however, no alpha driven TAEs were observed. Theoretical analyses of the TAE instability of these DT discharges indicate that the alpha pressure required for TAE stability still remained greater than that actually obtained in this experiment, mainly because of the effects of beam ion Landau damping. (45 refs.)
Antoni, V. (1997). "Edge physics in reversed-field pinch and tokamak: similarities and differences." Plasma Physics and Controlled Fusion
24th European-Physical-Society Conference on Controlled Fusion and Plasma Physics 39: B223-B235.
The edge-region properties of a reverse-field pinch (RFP) configuration are reviewed and a comparison with those of tokamaks is attempted. Despite the different magnetic configurations, many similarities have been observed. Among the similarities is the electrostatic nature of the particle transport and the structure of the plasma potential. In particular the radial electric field changes sign across the last closed flux surface and gives rise to an E x B velocity shear layer. The anomalous transport driven by electrostatic and magnetic fluctuations is addressed and the regimes of improved confinement recently observed in RFPs are reviewed. The role of the velocity shear on turbulence stabilization in these regimes is discussed.
Antoni, V., D. Desideri, et al. (1997). "Plasma potential well and velocity shear layer at the edge of reversed field pinch plasmas." Physical Review Letters 79(24): 4814-17.
The first evidence of an electrostatic potential well at the edge of plasmas confined in a reversed field pinch configuration is reported, based on measurements made on the RFX experiment. The radial width of the well decreases with plasma current, whereas the potential drop is a few times the electron temperature at the edge, almost independent of plasma current. The resulting radial electric field points inward at the edge and this behavior has been related to finite Larmor radius losses as in tokamaks. Because of the spatial structure of the plasma potential, a naturally occurring double velocity shear layer has been identified at the edge with a shear value comparable to that of tokamaks and stellarators. (19 refs.)
Arista, N. R. and E. M. Bringa (1997). "Interaction of ion clusters with fusion plasmas: scaling laws." Physical Review A (Atomic, Molecular, and Optical Physics) 55(4): 2873-81.
The interaction between large ion clusters or very intense ion beams with fusion plasma is studied using the dielectric function formalism with appropriate quantum corrections. The contributions from individual and collective modes to the energy loss are calculated. The general properties of the interference effects are characterized in terms of the relevant parameters, and simple scaling laws are obtained. In particular, the conditions for a maximum enhancement in the energy deposition are derived. The study provides a unified view and a general formulation of collective effects in the energy loss for low and high velocities of the beam particles. (22 refs.)
Armstrong, R., A. Fredriksen, et al. (1997). "Measurements of ion temperature in weakly ionized, steady-state plasmas." Physics Essays 10(3): 483-91.
We describe and analyze the different methods of measuring the ion temperature in low-temperature, steady-state plasmas, that is, Katsumata probes, energy analyzers, and wave diagnostics. The experiment's were performed on two toroidal machines: the "Blaaman" at the University of Tromso and "Thorello" at the University of Milan, which produce very similar plasmas. The limitations and advantages of each diagnostic are compared and briefly discussed. (24 refs.)
Arunasalam, V. (1997). "Ion cyclotron emission due to fast Alfven-wave radiative instabilities in tokamaks induced by newly born fusion products." Journal of Plasma Physics 58: 287-313.
The velocity distribution functions of newly born (t=0) charged fusion products (protons in DD and alpha particles in DT plasmas) of tokamak discharges can be approximated by a monoenergetic ring distribution with a finite nu {sub perpendicular to } approximately= nu {sub |} approximately=V{sub j}, where 1/2 M{sub j}V{sub j}{sup 2}=E{sub j}, the directed birth energy of the charged fusion-product species j of mass M{sub j}. As the time t progresses, these distribution functions will evolve into a Gaussian in velocity (i.e. a drifting Maxwellian type), with thermal spreads given by the perpendicular and parallel temperatures T{sub perpendicular to j}(t)=T{sub |j}(t), with T{sub j}(t) increasing as t increases and finally reaching an isotropic saturation value of T{sub perpendicular to j}(t approximately= tau {sub j})=T{sub j}(t approximately= tau {sub j}) approximately=[M{sub j}T{sub d}E{sub j}/(M{sub j}+M)]{sup 1/2}. Here T{sub d} is the temperature of the background deuterium plasma ions, M is the mass of a triton or a neutron for j=protons and alpha particles respectively, and tau {sub j} approximately= 1/4 tau {sub sj} is the thermalization time of the fusion product species j in the background deuterium plasma, with tau {sub sj} the slowing-down time. For times t of the order of tau {sub j}, the distributions call be approximated by a Gaussian in the total energy (i.e. a Brysk type). Then, for times t>or= tau {sub sj}, the velocity distributions of the fusion products will relax towards their appropriate slowing-down distributions. We shall examine the radiative stability of all these (i.e. a monoenergetic ring, a Gaussian in velocity, a Gaussian in energy, and the slowing-down) distributions. (48 refs.)
Barnes, C. W., M. J. Loughin, et al. (1997). "Neutron activation for ITER." Review of Scientific Instruments
11th Topical Conference on High-Temperature Plasma Diagnostics 68(1): 577-580.
There are three primary goals for the Neutron Activation system for ITER: to maintain a robust relative measure of fusion power with stability and wide dynamic range (seven orders of magnitude), allow an absolute calibration of fusion power production, and provide a flexible and reliable system for materials testing. The nature of the activation technique is such that stability and wide dynamic range can be intrinsic properties of the system. It has also been the technique that demonstrated (on JET and TFTR) the most accurate neutron measurements in DT operation. Since the detectors for assaying the radioactivity are not located on the tokamak and are therefore amenable to accurate characterization, and if the activation samples are placed very close to the ITER plasma with minimal scattering or attenuation, high overall accuracy in the fusion energy production (7%-10%) should be achievable on ITER. In the paper, a conceptual design is presented. A system is shown to be capable of meeting these three goals, and unresolved design issues are identified. (7 refs.)
Bell, M. G., S. Batha, et al. (1997). "Deuterium-tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor." Physics of Plasmas
38th Annual Meeting of the Division-of-Plasma-Physics of the American-Physical-Society 4(5): 1714-1724.
Experiments in the Tokamak Fusion Test Reactor (TFTR) [Phys. Plasmas 2, 2176 (1995)] have explored several novel regimes of improved tokamak confinement in deuterium-tritium (D-T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high l{sub i}). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q{sub a} approximately=4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l{sub i} plasmas produced by rapid expansion of the minor cross section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D-T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D-T plasmas with q{sub 0}>1 and weak magnetic shear in the central region, a toroidal Alfven eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions. (39 refs.)
Bolvin, R. L., M. Koltonyuk, et al. (1997). "Time-of-flight neutral particle analyzer for Alcator C-Mod." Review of Scientific Instruments
11th Topical Conference on High-Temperature Plasma Diagnostics 68(1): 982-985.
A neutral particle analyzer, based on the time-of-flight technique is scheduled to be installed on the Alcator C-Mod tokamak. The instrument was originally designed and used in the ZT-40 experiment, and later, briefly used on the CTX experiment at Los Alamos National Laboratory. The design consists of a chopper wheel mounted on a turbomolecular pump, a similar to 2 m long flight tube, a Cu-Be secondary electron emitting surface, and an electron multiplier, which together are used as the neutral particle detector. Among the changes introduced in the system from the original design, of particular interest is the location of the instrument. The instrument is mounted behind a magnetic-type neutral particle analyzer and shares the same Line of sight. Both of them can be scanned, poloidally down to the X point and tangentially to R/R(0) = 0.7. This will enable us to compare neutral flux through two different techniques with an overlap in energies from similar to 0.5-4 keV, with the instrument capable of detecting neutral particles with energies as low as similar to 20 eV. This should aid in characterizing neutral and ion behavior at the edge of the tokamak plasma especially near the X point. We will also describe changes made to the controls for improved versatility and ease of operation. (C) 1997 American Institute of Physics.
Bush, C. E., R. E. Bell, et al. (1997). "Core V-phi and T-i profiles and transport in TFTR DD and DT plasmas with lithium conditioning." Journal of Nuclear Materials
12th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices 241: 892-896.
High performance DT plasmas have been obtained using neutral beam heating with lithium (Li) conditioned graphite walls in TFTR. Values of tauE > 300 ms have been obtained with neutron source rates of > 10(18) n/s and n tau T approximate to 10(21). Also, ion temperature (T-i)> 40 keV and toroidal velocity (V-phi)> 800 km/s have been obtained. The T-i(R, t) and V-phi(R, t) profiles show strong gradients near the plasma core with del V-phi > 35 X 10(6)/s and E X B shearing rate > 2 X 10(5)/s realized. This strong E X B flow shear is consistent with formation of a 'transport barrier' in the plasma core. Measured V-phi, T-i, and carbon density, n(c), profiles from charge-exchange recombination spectroscopy (CHERS) and neoclassical calculations of poloidal velocity, V-phi, are used to assess the roles of the pressure and velocity contributions to E-r (or E X B) with varying Li conditioning. The profiles and gradients and resulting confinement and transport are found to vary with the amount of Li applied and the Li deposition technique. Correlations between the V-phi and T-i profiles and rccycling and impurity behavior as implied from edge carbon and D-alpha light and Li deposition are also observed.
Carreras, B. A. (1997). "Progress in anomalous transport research in toroidal magnetic confinement devices." IEEE Transactions on Plasma Science 25(6): 1281-321.
Transport is the outstanding physics issue in the quest for fusion by magnetic confinement. In spite of the intrinsic difficulty, a great deal of progress has been made in the past 25 years. Experiments have gone from being dominated by high anomalous losses, of the order of Bohm diffusion losses, to operation with no anomalous transport. This success is due to a combination of improved experimental infrastructure and the high degree of knowledge on how to control plasma discharges, Both have made it possible to access enhanced confinement regimes and to unravel new effects in confinement physics. Although there is not yet a complete understanding of the dynamical mechanisms underlying the anomalous transport process, there is some understanding of important components such as the ion transport loss mechanism at the plasma core and of the main mechanism for turbulence suppression in the enhanced confinement regimes. (461 refs.)
Chance, M. S. (1997). "Vacuum calculations in azimuthally symmetric geometry." Physics of Plasmas 4(6): 2161-80.
A robustly accurate and effective method is presented to solve Laplace's equation in general azimuthally symmetric geometry for the magnetic scalar potential in the region surrounding a plasma discharge which may or may not contain external conductors. These conductors can be topologically toroidal or spherical, and may have toroidal gaps in them. The solution is incorporated into the various magnetohydrodynamic stability codes either through the volume integrated perturbed magnetic energy in the vacuum region or through the continuity requirements for the normal component of the perturbed magnetic field and the total perturbed pressure across the unperturbed plasma-vacuum boundary. The method is based upon using Green's second identity and the method of collocation. As useful by-products, the eddy currents and the simulation of Mirnov loop measurements are calculated. (22 refs.)
Conn, R. W., R. P. Doerner, et al. (1997). "Beryllium as the plasma-facing material in fusion energy systems-experiments, evaluation, and comparison with alternative materials." Fusion Engineering and Design 37(4): 481-513.
The history of plasma-facing materials in fusion systems is reviewed to explain the evolution of needs and requirements that has led to the interest in beryllium today. Critical fusion-related characteristics of beryllium, including outgassing properties, plasma-driven erosion, surface morphology modification, and post-bombardment deuterium retention have been investigated under high-flux, steady-state deuterium plasma bombardment conditions. In experiments with beryllium, hydrogen and water molecules are observed to be the major desorbed species of as-received beryllium. Different beryllium samples manufactured using sintered powder metallurgical methods show similar outgassing behavior. However, cast beryllium samples desorb 20 times less gas than sintered beryllium. Simulating the divertor plasma conditions of the International Thermonuclear Experimental Reactor (ITER), the PISCES-B plasma generating facility is used to bombard targets with a deuterium plasma. The sputtering yield of beryllium is measured using a weight loss method. The target temperature and the partial pressure of residual gases during plasma bombardment are observed to be the critical parameters affecting erosion behavior. Data on sputtering yield for room temperatures (about 40 degrees C) and high temperatures (above 250 degrees C) samples is reported. It is also observed that at the higher temperatures, impurities from the plasma having low atomic number (such as carbon, nitrogen, and oxygen) are deposited on the beryllium surface. An impurity layer forms which is not eroded by extended plasma exposure. In-situ emission spectroscopy shows that the Be II line intensity drops by an order of magnitude as the deposited layer begins to form, and remains low during plasma exposure. (104 refs.)
delaCal, E. and E. Gauthier (1997). "First-wall cleaning and isotope control studies by D{sub 2} ICRF conditioning in Tore Supra with a permanent magnetic field." Plasma Physics and Controlled Fusion 39(7): 1083-99.
The use of a permanent magnetic field in superconducting magnetic fusion devices impedes conditioning by glow discharges as actually applied in most pulsed machines. An alternative has been studied in Tore Supra for this purpose with a 3.8 T permanent field: a discharge produced by ion cyclotron range of frequency wave injection (ICRF). Helium ICRF discharge conditioning (ICRF-DC) has already been shown to desaturate the deuterium-loaded carbon first wall efficiently. In this paper, we describe how D{sub 2} ICRF-DC can be applied to clean the wall or change its hydrogen isotopic ratio. This is achieved by pumping wall-desorbed molecules induced by particle bombardment from the ICRF plasma. The conditioning efficiency is optimized as a function of two input parameters: the gas pressure and the applied power. A plasma characterization is also given as a function of these parameters and a global model has been used to compare the experimental results and to interpret the main processes to which the neutral and charged species are subjected. The key to optimizing ICRF-DC is the wave ion heating to produce energetic particle bombardment of the first wall, while still achieving a low electron density, low-temperature plasma (n{sub e}<or=10{sup 17} m{sup -3}, T{sub e}<6 eV) to avoid reionization and redeposition of wall-desorbed neutrals before they can be evacuated by the torus pumping system. (16 refs.)
Fasoli, A., D. Borba, et al. (1997). "Alfven eigenmode experiments in tokamaks and stellarators." Plasma Physics and Controlled Fusion
24th European-Physical-Society Conference on Controlled Fusion and Plasma Physics 39: B287-B301.
In tokamaks and stellarators, measurements of electromagnetic fluctuations in the presence of resonant particle drive, including fusion-produced alpha's, reveal the excitation of Alfven eigenmodes (AE), related under certain conditions to a degradation in the fast-particle confinement. The balance between the drive and the background damping is investigated using active diagnostic systems to excite and measure the AE spectrum in terms of frequencies and damping rates. At JET, saddle-coil antennae drive low toroidal mode number (n < 4) AE in the range 30-500 kHz, including toroidal AE, kinetic AE, elliptical AE and global AE. Conditions for weak damping (gamma/omega(damp) < 1%) are identified. Low-n AE appear to be strongly damped (gamma/omega(damp) > 1%) during the creation of the magnetic X-point. In the presence of resonant fast particles, information on the instability drive is obtained: low-n modes are found to be stable in the presence of NBI with upsilon(parallel to)/upsilon(A) < 1. Fast ions generated by ICRH are observed to produce a drive for 4(ICRH) > P-thresh, with 2.5 MW < P-thresh < 5 MW; under these conditions, intrinsically driven TAE and EAE are clearly observed in the magnetic fluctuation spectra, with no measurable effect on the plasma performance.
Fredrickson, E. D., S. A. Sabbagh, et al. (1997). "The stability of advanced operational regimes on the Tokamak Fusion Test Reactor." Physics of Plasmas
38th Annual Meeting of the Division-of-Plasma-Physics of the American-Physical-Society 4(5): 1589-1595.
The performance of the Tokamak Fusion Test Reactor [D. Meade and the TFTR Group, in Plasma Physics and Controlled Nuclear Fusion Research, Washington, D.C. 1990 (International Atomic Energy Agency, Vienna, 1991), vol. I, pp. 9-24], as defined by the maximum fusion power production, has been limited, not by confinement, but by stability to pressure-driven modes. Two classes of current profile modification have been investigated to overcome this limit. A new technique has been developed to increase the internal inductance of low-q (q approximately=4), high-current (Ip>2 MA) plasmas. As was the case at higher edge q, the disruptive beta limit has been found to increase roughly linearly with the internal inductance, l{sub i}. Plasmas with hollow current profiles, i.e., reversed shear, are also predicted and experimentally observed to have increased stability in the negative shear region to ballooning and kink modes. However, performance of these plasmas is still limited by pressure-driven modes in the normal shear region. (29 refs.)
Gauthier, E., E. delaCal, et al. (1997). "Wall conditioning technique development in Tore Supra with permanent magnetic field by ICRF wave injection." Journal of Nuclear Materials
12th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices 241: 553-558.
Wall conditioning techniques tokamaks with a permanent magnetic field have been performed in Tore Supra by using an ion cyclotron range frequency (ICRF) facility. Plasmas have been produced by injection of ICRF power in the range from 40 kW to 350 kW either in helium or deuterium gas. Electron density in the range of 1.10(17) to 6.10(17) m(-3) and electron temperatures from 1.5 to 8 eV have been measured depending on the gas pressure and injected power. Energetic neutral atoms of hydrogen and deuterium with energies up to 50 keV have been produced. High hydrogen removal rates have been obtained in helium discharges, either in a continuous or pulsed operation mode.
Gerstel, U., L. Horton, et al. (1997). "Quantitative simulation of non-thermal charge-exchange spectra during helium neutral beam injection." Plasma Physics and Controlled Fusion 39(5): 737-56.
Non-thermal He II spectra for discharges with helium beam fuelling are analysed. Simulated spectra are used to study the effects of plasma temperature, plasma density and Z{sub eff} on observed charge-exchange (CX) spectra. Differences in modelling the non-thermal velocity distribution function with a numerical Fokker-Planck code or alternatively using analytical expressions are investigated. The intensities and spectral shapes of both active, localized CX spectra and competing, non-localized, passive electron-impact excitation components are simulated and compared with observations. The `plume' contributions of electron-impact excited He{sup +,*} particles are found to be quite appreciable and uncertainties in the plume calculation lead to non-negligible errors in the extraction of the active signal from the total spectrum. However, for experimental conditions with magnetic field configurations minimizing the plume effect good agreement can be found between fast-particle densities derived from the numerical calculations and the experimental observations. Significant problems in deriving absolute He{sup 2+} densities are encountered when a helium beam also acts as a CX diagnostic beam. For the case of dominant passive emission components, simulated fast spectral intensities for the core lines of sight agree within a factor of two with experimental data. (39 refs.)
Idzorek, G. C. and H. Oona (1997). "Properties of plasma radiation diagnostics." Review of Scientific Instruments
11th Topical Conference on High-Temperature Plasma Diagnostics 68(1): 1065-1068.
A number of diagnostics utilizing the radiation emitted from high-temperature plasmas have been developed at Los Alamos. Photoemissive X-ray diodes with photon energy bandpass filters provide time-resolved rough spectral data from about 6 eV to >10 keV photon energy. Filtered silicon photodiodes can be used down to 1 eV and offer the advantages of nominally flat response and ability to operate in poor vacuum conditions. Both types of diodes will provide a rough time-resolved spectrum and both are relatively inexpensive, reliable, and passive (i.e., no synchronization problems). For higher-energy resolution, bent crystal spectrographs are used in the X-ray region. With the addition of streak cameras or gated microchannel plates these systems provide data with high energy and high time resolution. To measure the total energy output, a thin foil bolometer is used that measures the change in foil resistance as it is heated by the plasma radiation. By combining these diagnostics into a complementary set good diagnostic information can be guaranteed on any plasma experiment. (19 refs.)
Ishida, S., T. Fujita, et al. (1997). "Achievement of high fusion performance in JT-60U reversed shear discharges." Physical Review Letters 79(20): 3917-21.
Fusion performance of reversed shear discharges with an L-mode edge has been significantly improved in a thermonuclear dominant regime with up to 2.8 MA of plasma current in the JT-60U tokamak. The core plasma energy is efficiently confined due to the existence of persistent internal transport barriers formed for both ions and electrons at a large minor radius of r/a~0.7 near the boundary of the reversed shear region. In an assumed deuterium-tritium fuel, the peak fusion amplification factor defined for transient conditions involving the dW/dt term would be in excess of unity. (16 refs.)
Kaneko, J., M. Katagiri, et al. (1997). "A directional neutron detector based on a recoil proton telescope for neutron emission profile monitor." Nuclear Instruments & Methods in Physics Research, Section A (Accelerators, Spectrometers, Detectors and Associated Equipment) 385(1): 157-60.
A directional neutron detector based on a recoil proton telescope was investigated by experimental approach. The detector showed a directionality not only for 14-MeV neutrons but also for fission and 2.4-MeV neutrons. Due to a good rise-time pulse-shape discrimination capability of a CsI(Tl) scintillator, a recoil proton energy spectrum was clearly observed in high gamma-ray background. At a detector angle of 21 degrees with respect to the neutron incident direction, the counts corresponding to the high energy neutron became a half of those at 0 degrees . A major source for reducing the signal-to-noise ratio was identified to be the energetic protons via the (n,p) reaction in the CsI(Tl) scintillator. (11 refs.)
Larsson, D., H. Bergsaker, et al. (1997). "Surface modification by solid target boronisation in the Extrap T2 experiment." Vacuum
5th European Vacuum Conference (EVC 5) / 10th International Conference on Thin Films (ICTF 10) 48(7/9): 693-5.
A solid target boronisation has been performed in the Extrap T2 reversed field pinch. A boron carbide rod was inserted in the edge plasma during hydrogen RFP-discharges. The amount of material released has been measured and the boron carbide film created on the first wall analysed by scanning Auger spectrometry, X-ray photoelectron spectroscopy, and ion beam techniques. The article reports on the effect the boronisation has upon plasma performance, and contains a quantitative discussion on the oxygen reduction due to oxidation of boron. (10 refs.)
Lengyel, L. L. (1997). "Theory of pellet cloud oscillation striations - Comment." Plasma Physics and Controlled Fusion 39(4): 647-649.
The analogy between the striations seen in pellet plasmas and those observed in some extraterrestrial plasma scenarios is discussed. With regard to the origin of the striations, some alternatives, complementary to Parks' proposal (1996 Plasma Phys. Control. Fusion 38 571-91), are outlined.
Loughlin, M. J., F. E. Cecil, et al. (1997). "Evaluation of an ITER compatible, thin foil Faraday collector as a lost alpha particle diagnostic for high yield D-T fusion plasmas." Review of Scientific Instruments
11th Topical Conference on High-Temperature Plasma Diagnostics 68(1): 361-364.
We have examined the concept of a thin foil Faraday collector as a lost alpha particle detector capable of operating under ITER-like conditions. A prototype detector consisting of a single set of four 2.5 mu m Ni foils was installed on the JET first wall and operated during a variety of deuterium plasma conditions during the 1995 JET run period. Although there was no significant production of alpha particles during these plasmas, the prototype demonstrated the expected resistance to the high temperature and X-ray backgrounds, as well as moderate neutron and gamma ray backgrounds characteristic of these plasmas. In addition, this prototype showed no significant response to neutral beam, RF, or lower hybrid plasma heating. The device did pick up a low level signal when neutral beams were injected simultaneously with heavy gas puffing. Strong intermittent correlations were seen with excursions in the H alpha edge brightness signal. In addition, the detector produced a significant signal in response to a roughly 250 ms disruption precursor. A similar prototype detector was installed immediately outside the TFTR vacuum vessel during the 1994 D-T run period to test the expected insensitivity to neutron backgrounds. No signal was seen above background during D-T plasmas for which the fast neutron production was in excess of 2*10{sup 18} n/s. (6 refs.)
Mahajan, S., R. Singh, et al. (1997). "Theory of plasma confinement in devices with pure toroidal field." Physics of Plasmas 4(7): 2612-18.
Equilibrium, stability and confinement in currentless toroidal device is studied in the terms of the flow-fluctuation cycle. In the initial seed equilibrium provided by the limiter, Rayleigh-Taylor (RT) fluctuations grow appreciably. These fluctuations are additional source of rotational transform in two ways. First, they directly drive a poloidal flow via Reynolds stress which improves the equilibrium. Second, the flow modifies the RMS level profile in such a way that the ponderomotive force due to the fluctuations impedes the free fall. Detailed linear theory of Rayleigh-Taylor fluctuations with poloidal flow is presented and criteria for flow stabilization are identified. Using the exact eigenfunction of fluctuations, an exact ordinary differential equation for poloidal flow is derived and solved using an ansatz. Finally, the relevance of this analysis to the recently proposed low to high confinement mode transition theories is discussed. (19 refs.)
McKee, G. R., R. J. Fonck, et al. (1997). "Transport measurements for confined non-thermal alpha particles in TFTR DT plasmas." Nuclear Fusion 37(4): 501-16.
Fusion produced non-thermal alpha particle radial profile measurements are obtained with the alpha -CHERS diagnostic in deuterium-tritium (DT) supershot plasmas on the Tokamak Fusion Test Reactor (TFTR). Alpha particles in the energy range 0.15<or=E{sub alpha }<or=0.6 MeV are observed spectroscopically over a five point radial profile. The extracted non-thermal alpha signal is <or=1% of the background bremsstrahlung intensity for typical total fast alpha densities in the range (0.5-1.0)*10{sup 17} m{sup -3}. The profiles obtained in two sets of discharges vary slightly, and are best described by a slowing down alpha distribution subject to neoclassical diffusion plus a small anomalous cross-field diffusion. The data are consistent with an effective anomalous diffusion coefficient in the range 0.00<or=D{sub alpha ,a}<0.10 m{sup 2}/s, where D{sub alpha ,a} is constant with alpha energy and with radius. (34 refs.)
Messiaen, A. M., J. Ongena, et al. (1997). "High confinement and high density with stationary plasma energy and strong edge radiation cooling in the upgraded Torus Experiment for Technology Oriented Research (TEXTOR-94)." Physics of Plasmas
38th Annual Meeting of the Division-of-Plasma-Physics of the American-Physical-Society 4(5): 1690-1698.
An overview of the results obtained so far for the radiative I-mode regime on the Upgraded Torus Experiment for Technology Oriented Research (TEXTOR-94) [Proceedings of the 16th IEEE Symposium on Fusion Engineering (Institute of Electrical and Electronics Engineers, Piscataway, NJ, 1995) vol. 1, p. 470] is given. This regime is obtained under quasistationary conditions with edge neon seeding in a pumped limiter tokamak with circular cross section. It combines high confinement and high beta (up to a normalized beta, beta {sub n}=2) with low edge q values (down to q{sub a}=2.8) and high density even above the Greenwald limit together with dominant edge radiative heat exhaust, and therefore shows promise for the future of fusion research. Bulk and edge properties of these discharges are described, and a detailed account is given of the energy and particle confinement and their scaling. Energy confinement scales linearly with density as for the nonsaturated ohmic Neo-Alcator scaling, but the usual degradation with total power remains. No deleterious effects of the neon seeding on fusion reactivity and plasma stability have been observed. (33 refs.)
Mikkelsen, D. R., J. Manickam, et al. (1997). "Deuterium-tritium simulations of the enhanced reversed shear mode in the Tokamak Fusion Test Reactor." Physics of Plasmas 4(5): 1316-1325.
The potential performance, in deuterium-tritium plasmas, of a new enhanced confinement regime with reversed magnetic shear [enhanced reversed shear (ERS) mode] is assessed. The equilibrium conditions for an ERS mode plasma are estimated by solving the plasma transport equations using the thermal and particle diffusivities measured in a short duration ERS mode discharge in the Tokamak Fusion Test Reactor [F.M. Levinton, et al., Phys. Rev. Lett. 75, 4417 (1995)]. The plasma performance depends strongly on Z{sub eff} and neutral beam penetration to the core. The steady-state projections typically have a central electron density of ~2.5*10{sup 20} m{sup -3} and nearly equal central electron and ion temperatures of ~10 keV. In time-dependent simulations the peak fusion power, ~25 MW, is twice the steady-state level. Peak performance occurs during the density rise when the central ion temperature is close to the optimal value of ~15 keV. The simulated pressure profiles can be stable to ideal magnetohydrodynamic instabilities with toroidal mode number n=1,2,3,4 and infinity for beta {sub norm} up to 2.5; the simulations have beta {sub norm}<or=2.1. The enhanced reversed shear mode may thus provide an opportunity to conduct alpha physics experiments in conditions similar to those proposed for advanced tokamak reactors. (40 refs.)
Muller, H. W., P. T. Lang, et al. (1997). "Improvement of q-profile measurement by fast observation of pellet ablation at ASDEX upgrade." Review of Scientific Instruments 68(11): 4051-60.
The q-profile measurement presented in this article is based on the observation of pellets injected into the plasma and viewed with a fast-framing camera. The pellets sublimate in the plasma and the ablated material is ionized. The ions move along the magnetic field lines, while an ablation cloud is forming. This ablation cloud contains a small fraction of neutral particles which emit visible light after collisional excitation. It is therefore possible to visualize the magnetic field lines along which the pellet passes. The inclination angle of the magnetic field lines with respect to the torus midplane can be determined from this observation. The results are compared with the inclination angles delivered by an equilibrium code. Further analysis shows that it is not meaningful to determine the q-profile from the data of only the pellet ablation, in the case of an elongated plasma as in ASDEX Upgrade. However, the accuracy of the q-profile determined by an equilibrium code, especially in the plasma center, can be greatly improved by using the pellet measurements as additional input data. (25 refs.)
Nazikian, R., G. Y. Fu, et al. (1997). "Alpha-particle-driven toroidal Alfven eigenmodes in the tokamak fusion test reactor." Physical Review Letters 78(15): 2976-9.
Alpha-particle-driven toroidal Alfven eigenmodes (TAEs) have been observed for the first time in deuterium-tritium (D-T) plasmas on the tokamak fusion test reactor (TFTR). These modes are observed 100-200 ms following the end of neutral beam injection in plasmas with reduced central magnetic shear and elevated central safety factor [q(0)>1]. Mode activity is localized to the central region of the discharge (r/a<0.5) with magnetic fluctuation level B{sub perpendicular to }/B{sub ||}~10{sup -5} and toroidal mode numbers in the range n=2-4, consistent with theoretical calculations of alpha -TAE stability in TFTR. (25 refs.)
Nishimura, K., R. Horiuchi, et al. (1997). "Tilt stabilization by cycling ions crossing magnetic separatrix in a field-reversed configuration." Physics of Plasmas 4(11): 4035-42.
The stabilization of the tilt disruption in a field-reversed configuration is investigated by means of a three-dimensional particle simulation. The growth rate of tilting instability decreases as the plasma beta value at magnetic separatrix beta {sub sp} increases, while it is slightly affected by the finite ion Larmor radius parameter s and the hollowness parameter of an equilibrium current profile D for low beta {sub sp} and moderately kinetic (2<or=s<or=5) plasmas. It is found that the number flux of ions crossing the separatrix repeatedly increases with increasing beta {sub sp} and the crossing motion of ions plays a role in leading to the tilt stabilization by disturbing the unstable tilting motion. (23 refs.)
Nishitani, T., M. Isobe, et al. (1997). "Triton burnup measurements using scintillating fiber detectors on JT-60U." Fusion Engineering and Design
7th International Toki Conference on Plasma Physics and Controlled Nuclear Fusion Research (ITC-7) 34/35: 563-6.
A new type of 14 MeV neutron detector, based on scintillating fibres, was developed for triton burnup measurements in JT-60U deuterium discharges. The detector consists of an array of scintillating fibres embedded in an aluminum matrix, coupled to a magnetic-resistant phototube with a high current output base, enabling count rates up to 100 MHz. The detector characteristics have been investigated using the fusion neutron source DT neutron generator. From the time-dependent measurement of 14 MeV neutron emission in neutral beam heated plasma discharges, the fast triton diffusivity has been evaluated to be 0.05-0.15 m{sup 2} s{sup -1}. Loss of fast tritons caused by toroidal Alfven eigenmode instability was observed in the peripheral region of the plasma. (9 refs.)
Omelchenko, Y. A. and R. N. Sudan (1997). "A 3-D Darwin-EM hybrid PIC code for ion ring studies." Journal of Computational Physics 133(1): 146-59.
A new, 3-D electromagnetic (EM), hybrid, particle-in-cell (PIC) code, FLAME has been constructed to study low-frequency, large orbit plasmas in realistic cylindrical configurations. The stability and equilibrium of strong ion rings in magnetized plasmas are the first issues suitable for its application. In FLAME the EM-field is governed by Maxwell's equations in the quasi-neutral Darwin approximation (with displacement current neglected), the ion components are represented by discrete macro-particles, and the plasma electrons are modeled as a massless cold fluid. All physical quantities are expanded into finite Fourier series in the azimuthal ( theta ) direction. The discretization in the poloidal (r, z) plane is done by a finite-difference staggered grid method. The electron fluid equations include a finite scaler resistivity and macro-particles experience slowing-down collisions. A substantial reduction of computation time is achieved by enabling separate time advances of background and beam particle species in the time-averaged fields. FLAME has been optimized to run on parallel, MIMD systems, and has an object-oriented (C++) structure, The results of normal mode tests intended to verify the code ability to correctly model plasma phenomena are presented. We also investigate in 3-D the injection of a powerful annular ion beam into a plasma immersed in a magnetic cusp followed by an axially ramped applied magnetic field. A nonaxisymmetric perturbation is applied to the magnetic field and its effect on ion ring formation is analysed. (28 refs.)
Park, H. K., S. A. Sabbagh, et al. (1997). "The role of the neutral beam fueling profile in the performance of the Tokamak Fusion Test Reactor and other tokamak plasmas." Physics of Plasmas
38th Annual Meeting of the Division-of-Plasma-Physics of the American-Physical-Society 4(5): 1699-1706.
Scalings for the stored energy and neutron yield, determined from experimental data, are applied to both deuterium-only and deuterium-tritium plasmas in different neutral-beam-heated operational domains in Tokamak Fusion Test Reactor [Nucl. Fusion 25, 1167 (1985)]. The domain of the data considered includes the Supershot, high poloidal beta, low-mode, and limiter high-mode operational regimes, as well as discharges with a reversed magnetic shear configuration. The new important parameter in the present scaling is the peakedness of the heating beam fueling profile shape. Ion energy confinement and neutron production are relatively insensitive to other plasma parameters compared to the beam fueling peakedness parameter and the heating beam power when considering plasmas that are stable to magnetohydrodynamic modes. However, the stored energy of the electrons is independent of the beam fueling peakedness. The implication of the scalings based on this parameter is related to theoretical transport models such as radial electric field shear and ion temperature gradient marginality models. Similar physics interpretation is provided for beam heated discharges on other major tokamaks. (32 refs.)
Park, H. K. and S. A. Sabbagh (1997). "Effect of the neutral beam fuelling profile on fusion power, confinement and stability in TFTR." Nuclear Fusion 37(5): 629-42.
The stored energy and neutron yields of TFTR plasmas have been modelled, with small dispersion, by considering a parameter associated with the peakedness of the beam fuelling profile as an independent variable. Scalings with fixed multiplicative constants and exponential powers determined from the experimental data are applied to deuterium discharges at different major radii in TFTR, and also to deuterium-tritium (DT) plasmas. The domain of the data considered includes the supershot, high poloidal beta, L mode and limiter H mode operational regimes, as well as discharges with a reversed shear (RS) magnetic configuration. The role of plasma parameters, such as plasma current I{sub p}, edge safety factor q{sub a} and toroidal field B{sub t}, are considered in this study. Improved energy confinement and neutron production are relatively insensitive to these parameters compared with the beam fuelling peakedness parameter when considering plasmas that are stable to magnetohydrodynamic modes. In contrast, plasma stability is dependent on these quantities and also on parameters describing the peakedness of the equilibrium profiles. On the basis of externally controllable plasma parameters, the limitation and optimization of fusion power production of the present TFTR is investigated and a path towards a discharge condition with fusion power gain Q>1 is suggested. (40 refs.)
Perkins, L. J. (1997). "Shape-enhanced fusion: increasing the reactivity for some advanced fusion fuels." Physics Letters A 236(4): 345-50.
We show that an order of magnitude enhancement of the fusion cross section is possible for some advanced fusion fuels involving deformed nuclei providing the reacting nuclei fuse in a directed orientation. This process is a manifestation of Coulomb barrier reduction and we identify some candidate advanced fusion fuels which could exploit this phenomenon including p-{sup 11}B, p-{sup 7}Li and p-{sup 10}B. (22 refs.)
Peterson, B. J. and S. Sudo (1997). "Plans for two-dimensional steady state bolometric measurements of the Large Helical Device." Fusion Engineering and Design
7th International Toki Conference on Plasma Physics and Controlled Nuclear Fusion Research (ITC-7) 34/35: 307-10.
Preliminary plans for two-dimensional steady state bolometric measurements of core plasma and divertor regions have been made for the Large Helical Device (LHD) under construction at the National Institute for Fusion Science. Using a semi-tangential plane through the plasma volume, a two-dimensional view of the core plasma is achieved from the outboard, upper and lower diagnostic ports. Divertor regions will also be viewed from the upper, lower and outboard ports. In the case of divertor and core plasma arrays, duplicate detector arrays will alternate in turn between viewing and active cooling phases, using fast shutters to enable continuous detection during long-pulse (30 min) operation of the LHD. Each detector array will consist of 32 metal film bolometers and 64 XUV photodiode arrays, resulting in a spatial resolution of about 2-4 cm. The temporal resolution should be about 1 ms for the film bolometers and 1 mu s for the photodiodes. Additional arrays will include a one-dimensional array that views from an outboard port, and an array of toroidally spaced detectors for investigating the toroidal variation of the radiated power. During the present stage of development, various detector samples are being acquired and tested. Schematics of the bolometer arrays, as well as plans for the complete bolometer diagnostic system will be shown. (7 refs.)
Redi, M. H., R. B. White, et al. (1997). "Calculations of alpha particle loss for reversed magnetic shear in the Tokamak Fusion Test Reactor." Physics of Plasmas 4(11): 4001-8.
Hamiltonian coordinate, guiding center code calculations of the toroidal field ripple loss of alpha particles from a reversed shear plasma in the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)] predict 40% total alpha losses and 20% ripple diffusion losses. This is about double the loss rate of a comparable non-reversed magnetic shear plasma. High central q is found to increase alpha ripple losses as well as first orbit losses of alphas in the reversed shear simulations. Alpha ripple transport on TFTR affects ions within r/a=0.5, not at the plasma edge. The entire plasma is above threshold for stochastic ripple loss of alpha particles at birth energy in the reversed shear case simulated, so that all trapped 3.5 MeV alphas are lost stochastically or through prompt losses. (24 refs.)
Sasao, M., G. A. Wurden, et al. (1997). "Spectroscopic measurement of the Doppler broadening region of He II line emission of DT plasmas using impurity pellets." Fusion Engineering and Design
7th International Toki Conference on Plasma Physics and Controlled Nuclear Fusion Research (ITC-7) 34/35: 333-6.
The feasibility of spectroscopic measurement of the velocity distribution of alpha particle using the ablation cloud of a lithium or boron pellet was tested in the DT experiment of TFTR. The measurement was performed using a 10-channel narrow-band filtered spectrometer in the wavelength covering the 3.5 MeV alpha particle Doppler broadening region of the He II 468.6 nm line and its vicinity. The spectra from a lithium pellet consists of the continuum bremsstrahlung background, lithium line emissions and possibly a 468.6 nm helium line. However, no clear evidence of alpha particles was observed, even when boron pellets were injected. (4 refs.)
Sharma, P. K., J. P. Singh, et al. (1997). "Experimental study of a microwave produced toroidal plasma." Plasma Physics and Controlled Fusion 39(10): 1669-80.
We report production of hydrogen plasma using a 2.45 GHz microwave source in a toroidal device BETA. The equilibrium density, floating potential and plasma temperature in the poloidal cross section have been measured. Contours of equal density and floating potential exhibit a slab nature in the vertical direction of the poloidal cross section. Plasma is formed with a peak density of 4*10{sup 10} cm{sup -3}. The electron temperature is about 4-7 eV. Our measurements reveal that the X-mode wave is mode converted to electrostatic modes near upper hybrid resonance (UHR), which propagates close to the electron cyclotron resonance (ECR) region where the entire energy of the waves is deposited. Thus we observe plasma profiles broadened towards the weaker magnetic field region up to the UHR region, with an inner boundary near ECR region. This is further supported by numerical calculation. The fluctuating component of the density and floating potential is also measured. Low-frequency (LF) electrostatic instabilities are observed to be radially localized. The presence of both flute and drift modes is reported. Our measurements reveal that the excitation of LF fluctuations depends strongly on the sign of the product Del n. Del B where Del n and Del B are the radial density and magnetic field gradients, respectively. LF coherent fluctuating modes are observed in the region where Del n. Del B>0 and a turbulent spectrum is observed where the condition Del n. Del B<0 is satisfied. Our measurements reveal that collisional drift modes could be excited when eta {sub e}<0 (where eta {sub c}= Del lnT{sub c}/ Del In n{sub c}), whereas Rayleigh-Taylor modes could be excited when eta {sub e}>0 and the density gradient is antiparallel to the effective gravity `g'. (17 refs.)
Sheehan, D. P., J. Bowles, et al. (1997). "Ion acceleration and anomalous transport in the near wake of a plasma limiter." Physics of Plasmas 4(9): 3177-86.
Ion acceleration and anomalous transport were studied experimentally in the near wake region of an electrically floating disk limiter immersed in two different types of collisionless, supersonically flowing, magnetized plasmas: the first initially quiescent, the second initially turbulent. Ion densities and velocity distributions were obtained using a nonperturbing laser induced fluorescence diagnostic. Large-amplitude, low-frequency turbulence was observed at the obstacle edge and in the wake. Rapid ion and electron configuration space transport and ion velocity space transport were observed. Configuration space and velocity space transport were similar for both quiescent and turbulent plasma-obstacle systems, suggesting that plasma-obstacle effects outweigh the effects of initial plasma turbulence levels. (63 refs.)
Skinner, C. H., J. Kamperschroer, et al. (1997). "Measurements of tritium recycling and isotope exchange in TFTR." Journal of Nuclear Materials
12th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices 241: 887-891.
Tritium Balmer-alpha (T-alpha) emission, along with H-alpha and D-alpha is observed in the current D-T experimental campaign in TFTR. The data are a measure of the fueling of the plasma by tritium accumulated in the TFTR limiter and the spectral profile maps neutral hydrogenic velocities. T-alpha is relatively slow to appear in tritium neutral beam heated discharges, (T-alpha/(H-alpha + D-alpha + T-alpha) = 11% after 8 tritium-only neutral beam discharges). In contrast, the T-alpha fraction in a sequence of six discharges fueled with tritium puffs, increased to 44%. Larger transient increases (up to 75% T-alpha) were observed during subsequent tritium gas puffs. Analysis of the Doppler broadened spectral profiles revealed overall agreement with the dissociation, charge exchange, sputtering and reflection velocities predicted by the neutral Monte-Carlo code DEGAS with some deficiency in the treatment of dissociation products in the 10-100 eV range.
Skinner, C. H., E. Amarescu, et al. (1997). "Plasma wall interaction and tritium retention in TFTR." Journal of Nuclear Materials
12th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices 241: 214-226.
The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced a number of record breaking results including core fusion power, similar to 2 MW/m(3), comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently the in-toms tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. Recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues are presented.
Strachan, J. D., S. Batha, et al. (1997). "TFTR DT experiments." Plasma Physics and Controlled Fusion
24th European-Physical-Society Conference on Controlled Fusion and Plasma Physics 39: B103-B114.
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.
Takahashi, T., Y. Tomita, et al. (1997). "Collisionless pitch angle scattering of plasma ions at the edge region of a field-reversed configuration." Physics of Plasmas 4(12): 4301-8.
The motion of a plasma ion gyrating around the separatrix of a field-reversed configuration is studied. Numerical studies showed that the action integral of a particle changes abruptly when a particle passes through the vicinity of a field null x point. This phenomena is understood as collisionless stochastic scattering of the pitch angle. In the case of a particle with positive canonical angular momentum P{sub theta } the resultant correlation coefficients of the action integral between before and after the scattering appear to be stochastic in some cases. As the action integral increases for a particle with negative P{sub theta } its motion tends to be adiabatic. If the negative P{sub theta } of a particle approaches zero, a stochastic motion is observed. (17 refs.)
Taylor, T. S. (1997). "Physics of advanced tokamaks." Plasma Physics and Controlled Fusion
24th European-Physical-Society Conference on Controlled Fusion and Plasma Physics 39: B47-B73.
Significant reductions in the size and cost of a fusion power plant core can be realized if simultaneous improvements in the energy replacement time, tau(E), and the plasma pressure or beta, beta(T) = 2 mu(0) (P)/B-2 can be achieved in steady-state conditions with high self driven, bootstrap current fraction. Significant recent progress has been made in experimentally achieving these high performance regimes and in developing a theoretical understanding of the underlying physics. Three operational scenarios have demonstrated potential for steady-state high performance, the radiative improved mode, the high internal inductance or high l(i) scenario, and the negative central magnetic shear, NCS (or reversed shear) scenario. In a large number of tokamaks, reduced ion thermal transport to near neoclassical values, and reduced particle transport have been observed in the region of negative or very low magnetic shear: the transport reduction is consistent with stabilization of microturbulence by sheared E x B Bow. There is strong temporal and spatial correlation between the increased sheared E x B Bow, the reduction in the measured turbulence, and the reduction in transport. The DIII-D tokamak, the JET tokamak and the JT-60U tokamak have all observed significant increases in plasma performance in the NCS operational regime. Strong plasma shaping and broad pressure profiles, provided by the H-mode edge, allow high beta operation, consistent with theoretical predictions; and normalized beta values up to beta(T)/(I/a beta) = beta(N) similar to 4.5% m T MA(-1) simultaneously with confinement enhancement over L-mode scaling, H = tau/tau(ITER-89P) similar to 4, have been achieved in the DIII-D tokamak. In the JT-60U tokamak, deuterium discharges with negative central magnetic shear have reached equivalent breakeven conditions, Q(DT) (equiv) = 1.
Toyoda, H., M. Watanabe, et al. (1997). "Laboratory experiments on hydrogen and impurity behaviors in lithium-deposited environment." Journal of Nuclear Materials
12th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices 241: 1031-1035.
Wall conditioning with deposition of thin lithium layers gives rise to low hydrogen recycling and reduction in oxygen and carbon impurities. To understand such effects, small-scale laboratory experiments are carried out mainly focusing on the chemical state of hydrogen involved in the lithium layer. First, measurements of the methane yield from graphite walls in a hydrogen glow discharge reveal that the lithium deposition leads to similar to 25% reduction of chemical sputtering. Second, thermal desorption of hydrogen molecules from the lithium layer suggests existence of different states of hydrogen: one is weakly bound hydrogen which is released from the lithium layer at T similar to 200 degrees C, and the other is attributed to lithium hydride which decomposes for T > 400 degrees C. Finally, a new wall conditioning based on a Lithium borohydride (LiBH4) discharge is demonstrated, which enables co-deposition of lithium and boron.
Wagner, F. (1997). "Topics in toroidal confinement." Plasma Physics and Controlled Fusion
International Conference on Plasma Physics 39(5A): 23-50.
In this paper we shall highlight some topics of specific concern for both concepts for magnetic confinement: 2D (tokamaks) and 3D (helical) systems. Section 2 will address the primary goals of fusion: density, confinement and temperature. We shall summarize the density limit, the global confinement time scaling with engineering parameters and some physical elements of improved confinement regimes with edge (H-mode) and core transport barriers (reversed shear cases); progress in relation to temperature originates from the evidence for alpha -particle heating from TFTR and the successes in drift optimization in helical systems with some confirmation from W7-AS. In the section 3 on the secondary goals of fusion, beta will be discussed in terms of equilibrium (including the necessary improvements in helical systems with examples from W7-AS) and in terms of stability (including fast particle-driven Alfven waves with toroidally induced eigenmodes in cases with high shear and global eigenmodes in cases with low shear). Finally, we shall summarize the status of exhaust physics, with reference to the poloidal field divertor in tokamaks and the island divertor in helical systems. The intention of this report is to emphasize the complementarity of the two concepts of toroidal confinement. (90 refs.)
Wong, K. L., R. Majeski, et al. (1997). "Evolution of toroidal Alfven eigenmode instability in Tokamak Fusion Test Reactor." Physics of Plasmas 4(2): 393-404.
The nonlinear behavior of the toroidal Alfven eigenmode (TAE) driven unstable by energetic ions in the Tokamak Fusion Test Reactor (TFTR) [Phys. Plasmas 1, 1560 (1994)] is studied. The evolution of instabilities can take on several scenarios: a single mode or several modes can be driven unstable at the same time, the spectrum can be steady or pulsating, and there can be negligible or anomalous loss associated with the instability. This paper presents a comparison between experimental results and recently developed nonlinear theory. Many features observed in experiment are compatible with the consequences of the nonlinear theory. Examples include the structure of the saturated pulse that emerges from the onset of instability of a single mode, and the decrease, but persistence, of TAE signals when the applied rf power is reduced or shut off. (25 refs.)
Wurden, G. A. (1997). "A radiation-hard, steady state, digital imaging bolometer system." Fusion Engineering and Design
7th International Toki Conference on Plasma Physics and Controlled Nuclear Fusion Research (ITC-7) 34/35: 301-5.
The concept and design of a new type of bolometer system which can function with excellent spatial resolution and good time resolution in the next generation of long-pulse (or steady state), harsh-neutron environment, fusion plasmas, is outlined. It uses a cooled pinhole-camera design, employing a robust, passive, segmented radiation absorber. In one scenario, IR emission from the absorber's front surface is relayed by metal mirror optics to a shielded, high resolution IR video camera with +or-0.01 degrees C temperature resolution. In a second scenario, the matrix is viewed from its back side, in order to eliminate any extraneous IR radiation from the vessel or plasma. The system can make thousands of simultaneous `pixel' measurements at up to 50-60 Hz, with no wire leads through the vacuum interface. (9 refs.)
Wurden, G. A., B. J. Peterson, et al. (1997). "Design of an imaging bolometer system for the large helical device." Review of Scientific Instruments
11th Topical Conference on High-Temperature Plasma Diagnostics 68(1): 766-769.
We describe a radical design for a bolometer system employing infrared (IR) imaging of a segmented-matrix absorber in a cooled-pinhole camera geometry, which we will prototype and demonstrate on the large helical device (LHD). LHD will be operational in early 1998, with an l=2 superconducting winding, a major radius of 3.9 m, a minor radius of 0.5-0.65 m, and input powers ranging from 3 MW (steady state) to 30 MW (pulsed). The bolometer design parameters are determined by modeling the temperature of the foils making up the detection matrix using a two-dimensional time-dependent solution of the heat conduction equation. This design will give a steady-state bolometry capability, with modest (60 Hz) time resolution, while simultaneously providing hundreds of channels of spatial information. No wiring harnesses will be required, as the temperature-rise data is measured via a 12-bit, +or-0.025 degrees C resolution, 3-5 mu m band, 256*256 pixel IR camera. The spatial data will be used to tomographically invert the profile of the highly shaped stellarator main plasma and divertor radiation, in conjunction with more conventional fanned arrays of traditional bolometers. (10 refs.)
Yagi, Y., V. Antoni, et al. (1997). "Measurement of superthermal electron flow and temperature in a reversed-field pinch experiment by an electrostatic electron energy analyser." Plasma Physics and Controlled Fusion 39(11): 1915-27.
An electrostatic electron energy analyser has been inserted for the first time in the outer region of the RFX reversed-field pinch experiment, in order to investigate and characterize the presence of a superthermal electron population. It has been found that these electrons carry most of the current density parallel to the magnetic field. The time evolution during a single discharge of the superthermal electrons current density and parallel temperature indicates that the distortion of the electron distribution function is stationary during the plasma current flat-top phase. The dependence of the superthermal temperature on the plasma parameters has been investigated by varying the plasma density, and a relationship with the ratio of the on-axis applied electric field to the critical electric field for runaway generation has been identified. (28 refs.)
Zakharov, A. P., A. E. Gorodetsky, et al. (1997). "Hydrogen retention in plasma-facing materials and its consequences on tokamak operation." Journal of Nuclear Materials
12th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices 241: 52-67.
The current status of research in the area of hydrogen retention and release for the prime candidate plasma-facing materials is briefly reviewed. Physical understanding of the basic problems of hydrogen behavior in the surface layers and material bulk of graphite, beryllium and tungsten is emphasized. The data base in the field obtained in laboratory set-ups permits more thorough consideration of the recent experimental results in large tokamaks. It is known that hydrogen isotopes retention and release has a great influence on tokamak operating conditions and its plasma parameters. The processes occurring in tokamaks with hydrogen participation, Such as recycling, fuelling, codeposition, isotopic exchange and conditioning, now can be better understood, explained and controlled. Some consequences of hydrogen isotopes behavior in plasma and plasma-facing materials are described for long pulse tokamak operation.
Zarnstorff, M. C., F. M. Levinton, et al. (1997). "The effect of E{sub r} on motional-Stark effect measurements of q, a new technique for measuring E{sub r}, and a test of the neoclassical E{sub r}." Physics of Plasmas 4(4): 1097-102.
Previous analysis of motional-Stark effect (MSE) data to measure the q-profile ignored contributions from the plasma electric field. The MSE measurements are shown to be sensitive to the electric field and require significant corrections for plasmas with large rotation velocities or pressure gradients. MSE measurements from rotating plasmas on the Tokamak Fusion Test Reactor (TFTR) [Phys. Plasmas 2, 2176 (1975)] confirm the significance of these corrections and verify their magnitude. Several attractive configurations are considered for future MSE-based diagnostics for measuring the plasma radial electric field. MSE data from TFTR are analyzed to determine the change in the radial electric field between two plasmas. The measured electric field quantitatively agrees with the predictions of neoclassical theory. These results confirm the utility of a MSE electric field measurement. (35 refs.)
Zweben, S. J., V. Arunasalam, et al. (1997). "Alpha-particle physics in the Tokamak Fusion Test Reactor DT experiment." Plasma Physics and Controlled Fusion
International Conference on Plasma Physics 39(5A): 275-83.
A summary is presented of recent alpha-particle experiments on the Tokamak Fusion Test Reactor. Alpha particles are generally well confined in MHD-quiescent discharges, and alpha heating of electrons has been observed. The theoretically predicted toroidicity-induced Alfven eigenmode has been seen in discharges of <or=1 MW of alpha power, but only in plasmas with weak magnetic shear. (32 refs.)
Barnes, C. W., H. S. Bosch, et al. (1998). "Triton burnup measurements and calculations on TFTR." Nuclear Fusion 38(4): 597-618.
Measurements of the burnup of fusion product tritons in TFTR are presented. Interpretation of triton burnup experiments requires three accurate components: the measurement of the 2.5 MeV neutron emission, the measurement of the 14 MeV neutron emission and a calculation of the expected burnup ratio from the measured plasma parameters. The absolute calibration for the 14 MeV neutron measurements is provided by an NE213 proton recoil spectrometer. Time dependent burnup measurements for three plasma conditions selected for optimum detector operation are shown. Measurements of the time integrated triton burnup from copper activation foils (cross-calibrated to the NE213 measurements) are presented. Descriptions are provided of the neutron detectors and the plasma diagnostics whose data are used as input to the calculation of the expected burnup. All these measurements find that the triton burnup on TFTR is 1/2+or-1/4 the classical expectations for a wide variety of discharges. The burnup decreases for relatively longer triton slowing down times, implying possible fast ion diffusion coefficients of ~0.1 m{sup 2}/s. Alternatively, burnup appears to decrease with increasing major radius of the triton source and edge safety factor q{sub cyl}, implying that ripple losses may be playing a role. Triton burnup is a very sensitive measure of anomalous fast ion transport; similar levels of diffusive transport in an ignited reactor would have minimal impact on the alpha particles. (76 refs.)
Chang, Z., E. D. Fredrickson, et al. (1998). "Neoclassical tearing modes in Tokamak Fusion Test Reactor experiments. I. Measurements of magnetic islands and Delta '." Physics of Plasmas 5(4): 1076-84.
Tearing-type modes are observed in most high confinement operation regimes in the Tokamak Fusion Test Reactor (TFTR) [Nucl. Fusion 35, 1429 (1995)]. Three different methods are used to measure the magnetic island widths: external magnetic coils, internal temperature fluctuation from electron cyclotron emission (ECE) diagnostics and an experiment where the plasma major radius is rapidly shifted ("Jog" experiments). A good agreement between the three methods is observed. Numerical and analytic calculations of Delta ' (the tearing instability index) are compared with an experimental measurement of Delta ' using the tearing mode eigenfunction mapped from the jog data. The obtained negative Delta ' indicates that the observed tearing modes cannot be explained by the classical current-gradient-driven tearing theory. (23 refs.)
Ernst, D. R., B. Coppi, et al. (1998). "Unifying role of radial electric field shear in the confinement trends of TFTR supershot plasmas." Physical Review Letters 81(12): 2454-7.
A model is presented to explain the favorable ion thermal confinement trends of supershot plasmas in the Tokamak Fusion Test Reactor (TFTR). Turbulence suppression by radial electric field shear is important to reproduce the measured temperatures. Supershot confinement scalings are reproduced in more than sixty discharges, including favorable core power scaling and variation with isotopic mass, density peakedness, edge recycling, and toroidal rotation. The results connect the transitionless supershot regime with improved confinement regimes which are attained through sharp confinement transitions in the core or edge. (26 refs.)
Ernst, D. R., M. G. Bell, et al. (1998). "Notched velocity profiles and the radial electric field in high ion temperature plasmas in the Tokamak Fusion Test Reactor." Physics of Plasmas 5(3): 665-81.
A large "notch," or non-monotonic feature, appears in measured toroidal velocity profiles of the carbon impurity in the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Fusion 26, 11 (1984)], centered near the radius of strongest ion temperature gradient. This is explained as a consequence of radial momentum transport dominated by anomalous diffusion together with parallel heat friction on the impurity ions arising from the hydrogenic neoclassical parallel heat flow. The toroidal velocity profile of the hydrogenic species is predicted to be monotonic, from measurements of the impurity toroidal velocity, consistent with the anomalous radial diffusion of toroidal momentum. This supports a neoclassical calculation of the radial electric field for near-balanced beam injection. In supershot plasmas [Phys. Rev. Lett. 58, 1004 (1987)], a well structure in the radial electric field profile is found in the enhanced confinement region. An associated shear layer separates the core, where the local confinement trends are favorable, from the degraded outer region. This provides a mechanism for the nonlinear coupling of the ion temperature gradient, ion thermal confinement, and the radial electric field, which may help explain the favorable core confinement trends of very high temperature supershot plasmas. (58 refs.)
Gormezano, C., Y. F. Baranov, et al. (1998). "Internal transport barriers in JET deuterium-tritium plasmas." Physical Review Letters 80(25): 5544-7.
The observation of internal transport barriers (ITBs) in which ion thermal diffusivity is reduced to a neoclassical level has been made for the first time in tokamak plasmas fueled with deuterium and tritium ions using a broad current density profile. The heating and current profiles required to obtain an ITB are similar in D-T and D-D plasmas. Central ion temperatures of 40 keV and plasma pressure gradients of 10{sup 6} Pa/m were observed in a D-T plasma, leading to a fusion triple product n{sub i}T{sub i} tau {sub E}=1.10{sup 21} m{sup -3} keV s and 8.2 MW of fusion power. Then is potential for further optimization as a step towards the development of efficient tokamak fusion reactors. (11 refs.)
Hawryluk, R. J., S. Batha, et al. (1998). "Fusion plasma experiments on TFTR: A 20 year retrospective." Physics of Plasmas
39th Annual Meeting of the Division-of-Plasma-Physics of the American-Physical-Society 5(5): 1577-89.
The Tokamak Fusion Test Reactor (TFTR) (R.J. Hawryluk, to be published in Rev. Mod. Phys.) experiments on high-temperature plasmas, that culminated in the study of deuterium-tritium D-T plasmas containing significant populations of energetic alpha particles, spanned over two decades from conception to completion. During the design of TFTR, the key physics issues were magnetohydrodynamic (MHD) equilibrium and stability, plasma energy transport, impurity effects, and plasma reactivity. Energetic particle physics was given less attention during this phase because, in part, of the necessity to address the issues that would create the conditions for the study of energetic particles and also the lack of diagnostics to study the energetic particles in detail. The worldwide tokamak program including the contributions from TFTR made substantial progress during the past two decades in addressing the fundamental issues affecting the performance of high-temperature plasmas and the behavior of energetic particles. The progress has been the result of the construction of new facilities, which enabled the production of high-temperature well-confined plasmas, development of sophisticated diagnostic techniques to study both the background plasma and the resulting energetic fusion products, and computational techniques to both interpret the experimental results and to predict the outcome of experiments. (79 refs.)
Hawryluk, R. J. (1998). "Results from deuterium-tritium tokamak confinement experiments." Reviews of Modern Physics 70(2): 537-87.
Recent scientific and technical progress in magnetic fusion experiments has resulted in the achievement of plasma parameters (density and temperature) which made possible the production of significant bursts of fusion power from deuterium-tritium fuels and the first studies of the physics of burning plasmas. The key scientific issues in studying the reacting plasma core are plasma confinement, magnetohydrodynamic (MHD) stability, and the confinement and loss of energetic fusion products from the reacting fuel ions. Progress in the development of regimes of operation that both have good confinement and are MHD stable has made possible a broad study of problems in burning-plasma physics. The technical and scientific results from deuterium-tritium experiments on the Joint European Torus (JET) and the Tokamak Fusion Test Reactor (TFTR) are reviewed, with particular emphasis on alpha-particle physics issues. (311 refs.)
Haynes, L. A., J. P. Kelly, et al. (1998). "Neutral atom modeling of the TFTR first wall, pump ducts, and neutral beams." Fusion Technology 33(1): 74-83.
The DEGAS neutral transport code is used in two separate cases to simulate the neutral beam box and vessel of the Tokamak Fusion Test Reactor (TFTR). For the neutral beam box simulation, known input parameters include the ion density at the source exit and the proportion of input gas that is converted to the high-energy atomic beam. The T{sup 0} current to the torus is (1.61+or-0.03)*10{sup 20} s{sup -1}, with the high-energy beam having a median energy above 95 keV Corresponding results are found for the D{sup 0} current. In addition, the amount of gas reaching the torus, the pressure, and the flux and energy distributions of the ions and neutrals to the walls are found. For the tritium case, it is calculated that 92.4+or-0.2% of the input tritium reaches the cryopanels, 6.64+or-0.05% reaches the torus, and 1.0+or-0.2% reaches the ion dump. In the second run, DEGAS was used to calculate the neutral atom flux and energy of particles incident on the walls of the vacuum vessel and the neutral pressure in the pump duct of TFTR during a typical supershot with a 50/50 mixture of deuterium-tritium. Output quantities are the current and energy to the bumper limiter and first wall. The total amount of tritium implanted in the vacuum vessel after 150 shots of 1-s duration is estimated to be 0.5+or-0.1 g in the bumper limiter and 0.042+or-0.023 g in the outer wall and pumping duct, which is well within the 5-g on-site inventory and the 2-g in-vessel inventory. The implications of these results are discussed. (24 refs.)
Hedin, G. (1998). "MHD behaviour in a resistive-shell reversed-field pinch." Plasma Physics and Controlled Fusion 40(8): 1529-39.
The magnetic field on the external surface of the resistive shell has been measured in the T2 reversed-field pinch. Wall-locked internally-resonant modes, which produce toroidally localized ( Delta phi ~60 degrees ), large quasi-stationary m=1 radial field perturbations are observed. At the position of the wall-locked mode, m=0 perturbations are generated in local dynamo events and produce a large stationary toroidal field perturbation. The m=0 modes rotate toroidally from the location where they are produced and diminish within one toroidal turn. Numerical, resistive magnetohydrodynamic simulations verify the correlation between a m=1 radial field and a toroidal field perturbation. (21 refs.)
Hoffman, A. L. (1998). "Flux buildup in field reversed configurations using rotating magnetic fields." Physics of Plasmas 5(4): 979-88.
Rotating magnetic field (RMF) current drive is a very attractive method for both increasing the flux and sustaining the current in field reversed configurations (FRC). It has been demonstrated in low temperature, low field rotamaks, and will now be applied to a new translation, confinement, and sustainment (TCS) experiment attached to the LSX/mod (Large S field-reversed configuration Experiment) facility [Hoffman et al. Fusion Technol. 23, 185 (1993)]. Previous RMF calculations have been concerned primarily with the plasma currents and particle orbits produced in one-dimensional cylinders with the rotating field strength of near equal magnitude to the confining axial field. Both fluid current and particle orbits are calculated here in the more interesting regime appropriate to TCS and reactors where the confinement field far exceeds the rotating field strength. New insight is gained into both the flux buildup requirements for two-dimensional equilibria and into the limits on ion rotation in this high confinement field regime. (19 refs.)
Humphries, S. and C. Ekdahl (1998). "Finite-element simulation code for high-power magnetohydrodynamics." Laser and Particle Beams 16(3): 405-30.
We describe the mathematical basis and organization of Crunch, a 1D shock-hydrodynamics code to analyze pulsed-power experiments at Los Alamos National Laboratory. The program uses finite-element methods that preserve stability during material collisions and shock convergence on axis. It handles coupled calculations of nonlinear magnetic diffusion to simulate imploding liners. These calculations may be driven by multiple current waveforms or a self-consistent current variation derived from a pulsed-power generator model. Crunch incorporates elastic material contributions and calculates element break and melt points. The primary goal in program development was effective use by experimentalists. Crunch is controlled by a streamlined script language and runs on standard personal computers. An interactive graphical postprocessor expedites analysis of results. To support the program we have assembled data resources in machine-independent format including Sesame equation-of-state tables, a material strength library and a library of temperature-dependent conductivities. (29 refs.)
Kimura, H., Y. Kusama, et al. (1998). "Alfven eigenmode and energetic particle research in JT-60U." Nuclear Fusion
5th IAEA Technical Committee Meeting on Alpha Particles in Fusion Research 38(9): 1303-14.
Recent results on investigations of Alfven eigenmodes, fast ion confinement and fast ion diagnostics in JT-60U are presented. It was found that toroidicity induced Alfven eigenmodes (TAEs) were stable in negative shear discharges with a large density gradient at the internal transport barrier (ITB). If the density gradient was small at the ITB, multiple TAEs appeared around the q=2 surface (pitch minimum) and showed a large frequency chirping ( Delta f approximately=80 kHz). In low-q positive shear discharges, the location of the TAEs changed from outside to inside the q=1 surface; owing to a temporal change of the q profile. (39 refs.)
Kugel, H. W., G. Ascione, et al. (1998). "Neutron activation cool-down of the Tokamak Fusion Test Reactor." Fusion Technology
13th Topical Meeting on the Technology of Fusion Energy 34(3): 985-990.
TFTR final operations and post-shutdown neutron activation measurements were made. Ionization chambers were used to follow TFTR activation during operations and after shutdown. Gamma-ray energy spectroscopy measurements were performed to characterize TFTR activation at accessible vessel-bays, and on sample hardware removed from structures at various distances from the vessel. The results demonstrate long-lived activations from common, commercially available materials used in the fabrication and field engineering of TFTR. The measurements allow characterization of residual TFTR neutron activation, the projection of residual activation decay, and benchmarking of low activation issues simulations. (4 refs.)
Kuramoto, H., K. Toi, et al. (1998). "Study of toroidal current penetration during current ramp in JIPP T-IIU with fast response Zeeman polarimeter." Nuclear Fusion 38(1): 59-73.
Toroidal current penetration is studied in current ramp experiments on the JIPP T-IIU tokamak. The poloidal magnetic field profile in the peripheral region of a plasma (0.5<or=r/a<or=1.0) has been measured directly with a newly developed fast response Zeeman polarimeter. The experimental results indicate that a skin effect of the toroidal current density is clearly observed during both the current ramp-up (CRU) and the current ramp-down (CRD) experiments. The experimentally obtained toroidal current density profiles are well described by the profiles calculated on the assumption of neoclassical electrical resistivity, although the difference between the neoclassical electrical resistivity and the Spitzer electrical resistivity is not clearly distinguished because of the relatively collisional peripheral plasma region. Quasi-linear Delta ' analysis of tearing modes for the measured current density profile is consistent with the time behaviour of the coherent magnetohydrodynamic (MHD) modes such as m=4/n=1 or m=3/n=1 (m is the poloidal mode number and n is the toroidal mode number) often observed during the CRU phase. An obvious enhancement of the current penetration by the MHD modes is not detected in this ramp-up experiment, where the relative amplitude of the poloidal field fluctuations is less than 0.8%. (33 refs.)
Lackner, K. (1998). "Compatibility of improved confinement regimes with reactor requirements." Plasma Physics and Controlled Fusion
6th IAEA Technical Committee Meeting on H-Mode Physics 40(5): 557-68.
The reactor physics requirements are outlined in terms of characteristic performance measures. Four reactor designs of different scope are taken as examples. ITER is based on the ELMy H-mode as an operating regime demonstrated over many energy confinement times. Designs for power producing reactors require higher power densities and bootstrap current fractions, and therefore higher values of normalized beta . The capability of different enhanced confinement regimes, combining internal transport barriers with the L- or H-mode edge to satisfy these demands, together with the requirements on helium pumping and divertor power load, is discussed. (51 refs.)
Maehara, T. and R. Sugaya (1998). "Electron heating at half the electron cyclotron frequency by an extraordinary wave." Physics Letters A 247(4/5): 309-12.
We derive a simple formula to estimate the power absorption due to the sub-harmonic resonance at half the electron cyclotron frequency in a tokamak, and show that electron heating due to the resonance can be applicable to a large tokamak device like ITER. (15 refs.)
Messiaen, A. M. (1998). "Heating, confinement and extrapolation to reactors." Fusion Technology
3rd Carolus Magnus Summer School on Plasma Physics 33(2T): 373-84.
We discuss confinement, scaling, plasma thermalization and heating in tokamaks. (24 refs.)
Mirnov, S., I. Semenov, et al. (1998). "Phenomenology of major and minor disruptions in high beta deuterium and tritium tokamak fusion test reactor plasma." Physics of Plasmas 5(11): 3950-60.
The main subject of this work is the experimental study of the low m and n magnetohydrodynamics (MHD) perturbations during disruptive instability. This work presents the magnetic probe data, electron cyclotron emission (ECE), alpha -particle losses, and neutron flux data measured during the disruptive instability in high beta tokamak fusion test reactor (TFTR) [D.J. Grove and D.M. Meade, Nucl. Fusion 25, 1167 (1985)] plasmas. The major disruptions in high beta regimes go through several phases. The first phase is the fast (150-250 mu s) minor disruption (predisruption), causing a drop of the central temperature (and possibly, density). In this phase a powerful central m=1/n=1 mode initiates the sequential development of m=4/n=1, 3/1, 2/1, 3/2 peripheral modes, which lead to a 3/1 locked mode. The second phase is the slow thermal quench (2 ms) in the presence of a locked mode. The third phase is a fast positive current spike generation (5%-10% increase in Ip in less than 0.5 ms) and finally, the current quench occurs with a loss of 2.5 MA in 5 ms. (25 refs.)
Ohkuma, Y., M. Urano, et al. (1998). "Production of a low density field reversed configuration plasma." Nuclear Fusion 38(10): 1501-9.
The plasma density of a field reversed configuration (FRC) needs to be decreased below the present experimental regime in order to heat the plasma and sustain the configuration by a high energy neutral beam in an FRC reactor. However, as the plasma is produced in a linear vacuum vessel, there exists a severe breakdown limit at a low fill pressure as compared with a toroidal system. A method to form FRCs beyond the breakdown limit is proposed here. The preionized plasma is compressed by a strong bias field to enhance the plasma flow from the confinement region to the outside region and is then diluted before the start of the confinement field on the NUCTE device. The use of a diluted plasma enables the critical density of the FRC to be lowered from 1.1*10{sup 21} to 5.6*10{sup 20} m{sup -3} and the sum of the electron and ion temperatures to be increased from 0.35 to 0.81 keV. (24 refs.)
Okabayashi, M., E. D. Fredrickson, et al. (1998). "Mode structure of disruption precursors in TFTR enhanced reversed shear discharges." Nuclear Fusion 38(8): 1149-1160.
The mode structure of the disruption precursors in the TFTR enhanced reversed shear (ERS) parameters has been studied by using T-e fluctuation profiles and q profiles obtained from electron cyclotron emission (ECE) and motional Stark effect (MSE) measurements. The observed profiles of the radial displacement associated with the MHD modes were consistent with the displacement profiles expected from the ideal MHD external kink mode. The observed mode frequencies differ from the plasma toroidal rotation frequency measured with C VI charge exchange recombination light. The independence of the mode frequency from the plasma rotation frequency supports the ideal MHD hypothesis. Possible causes of the frequency difference are discussed.
Rogister, A. L. (1998). "Anomalous transport models for fusion plasmas: Review and perspectives." Fusion Technology
3rd Carolus Magnus Summer School on Plasma Physics 33(2T): 170-80.
The basic physical concepts underlying the theories of anomalous transport in magnetic confinement devices are reviewed. Anomalous transport is a consequence of electric and/or magnetic fluctuations driven by various linear and/or nonlinear instability mechanisms. The latter saturate by inducing a relaxation of the profiles towards a marginally stable state or/and by nonlinear coupling of the various modes. Specific theoretical models are described, together with their successes and drawbacks in the light of observed characteristics of plasma confinement, a non exhaustive list of which is given. A rough estimate of the nuclear heating power required to balance the anomalous losses in the International Tokamak Experimental Reactor (ITER) is calculated on the basis of the electrostatic drift wave instability model. (66 refs.)
Russell, D. A., J. R. Myra, et al. (1998). "Role of ponderomotive density expulsion in ion Bernstein wave coupling to the core plasma." Physics of Plasmas 5(3): 743-51.
When the density at the ion Bernstein wave (IBW) antenna is relatively low, mode transformation of the electron plasma wave to the IBW is sensitive to the density gradient scale length, and hence to ponderomotive effects. A second-order nonlinear ordinary differential equation that describes mode transformation at the lower-hybrid layer, including self-consistent ponderomotive density profile modification, is solved for the rf electrostatic potential in front of the IBW antenna, for the particular case of heating just below the second harmonic of the deuterium cyclotron frequency. The complex antenna impedance and a local reflectivity are calculated, assuming vacuum within the antenna box. These calculations reveal diminished antenna coupling to the IBW with increasing ponderomotive density expulsion, as compared to the linear prediction. The ponderomotive force increases the density gradient in the edge plasma, thus enhancing reflection and lowering the loading resistance. The model also describes the direct launch of IBWs in high edge density regimes, lacking a lower-hybrid layer, where the impedance is found to be much smaller than in the low density regime. (35 refs.)
Schoenberg, K. F., R. A. Gerwin, et al. (1998). "Magnetohydrodynamic flow physics of magnetically nozzled plasma accelerators with applications to advanced manufacturing." Physics of Plasmas
39th Annual Meeting of the Division-of-Plasma-Physics of the American-Physical-Society 5(5): 2090-104.
The coaxial plasma accelerator is a simple, compact, and mechanically robust device that utilizes the Lorentz J*B force to accelerate plasma to high velocity. Originally developed in the 1950s for the purpose of providing energetic plasmas for fusion energy experiments, coaxial plasma accelerators are presently being investigated as an environmentally sound and economical means of materials processing and advanced manufacturing. While commercial applications of this technology are already on line, future commercial applications will require improving accelerator reproducibility and efficiency, better controlling the accelerated plasma flow velocity or energy, and better controlling the distribution of directed energy or power on target. In this paper, the magnetohydrodynamic flow physics of magnetically nozzled plasma accelerators is presented with a view to achieving the accelerator control necessary for future industrial applications. Included is a fundamental description of plasma production, acceleration, and flow in a magnetic nozzle. (42 refs.)
Sharma, P. K. and D. Bora (1998). "Modification of low-frequency instabilities with high-frequency pump waves." Plasma Physics and Controlled Fusion 40(1): 163-73.
The influence of a large amplitude electric field oscillating near the upper hybrid frequency ( omega {sub 0}~ omega {sub uh}) on low-frequency (LF) electrostatic flute type instabilities, where the pump frequency ( omega {sub 0}/2 pi )is 2.45 GHz, is investigated in a pure toroidal plasma. Argon plasma is produced using a hot cathode discharge and microwaves are launched in the ordinary mode. Measurements reveal that the amplitude of the dominant fluctuating component in the cross power spectrum of density and floating potential fluctuations is suppressed for omega {sub 0}< omega {sub uh}, whereas the amplitude of the fluctuations increases in the case omega {sub 0}> omega {sub uh}. (17 refs.)
Skinner, C. H., M. G. Bell, et al. (1998). "Transport of recycled deuterium to the plasma core in the Tokamak Fusion Test Reactor." Physics of Plasmas 5(4): 1062-7.
Fueling of the plasma core by recycling in the Tokamak Fusion Test Reactor [Phys. Plasmas 2, 2176 (1995)] has been studied. In plasmas fueled by deuterium recycled from the limiter and tritium-only neutral beam injection, the DT neutron rate provides a measure of the deuterium influx into the core plasma. A reduced influx with plasmas using lithium pellet conditioning and with plasmas of reduced major (and minor) radius is found. Modeling with the DEGAS [D.P. Stotler et al., Phys. Plasmas 3, 4084 (1996)] neutrals code shows that the dependence on radius can be related to the penetration of neutrals through the scrape-off layer. (24 refs.)
Stoneking, M. R., J. T. Chapman, et al. (1998). "Experimental scaling of fluctuations and confinement with Lundquist number in the reversed-field pinch." Physics of Plasmas 5(4): 1004-14.
The scaling of the magnetic and velocity fluctuations with Lundquist number (S) is examined experimentally over a range of values from 7*10{sup 4} to 10{sup 6} in a reversed-field pinch (RFP) plasma. Magnetic fluctuations do not scale uniquely with the Lundquist number. At high (relative) density, fluctuations scale as b varies as S{sup -0.18}, in agreement with recent numerical results. Fluctuations are almost independent of S at low (relative) density, b varies as S{sup -1/2}. The range of measured exponents is narrow and is in clear disagreement with theories predicting b varies as S{sup -1/2}. At high relative density, the scaling of the energy confinement time follows expectations for transport in a stochastic magnetic field. A confinement scaling law (n tau {sub E} varies as beta {sup 4/5}, T{sup -7/10}, a{sup -3/5}, I{sub phi }{sup 2}) is derived, assuming the persistent dominance of stochastic magnetic diffusion in the RFP and employing the measured scaling of magnetic fluctuations. The peak velocity fluctuations during a sawtooth cycle scale marginally stronger than magnetic fluctuations but weaker than a simple Ohm's law prediction. The sawtooth period is determined by a resistive-Alfvenic hybrid time (T{sub saw} varies as square root tau {sub R} tau {sub A}) rather than a purely resistive time. (54 refs.)
Stroth, U. (1998). "A comparative study of transport in stellarators and tokamaks." Plasma Physics and Controlled Fusion 40(1): 9-74.
Experimental results on transport in stellarators and tokamaks are reviewed in a comparative sense. The objective is to learn about the importance of plasma current and magnetic shear for anomalous transport in high-temperature plasmas. On the basis of scaling expressions, the absolute values and parameter dependences of the confinement time are similar if the plasma current is expressed in terms of magnetic field parameters. The degradation of confinement with heating power is of the same order in both devices and it is difficult to explain it in terms of a diffusivity which depends on temperature or temperature gradient. The density dependence of confinement observed in stellarators has similar features as in ohmically heated tokamak discharges. A linear and a saturated regime can be distinguished. The critical density, at which saturation sets in, has a similar value and it seems to decrease with increasing machine size. Power degradation, transient transport, profile consistency and non-local transport are treated as related problems, which are connected to the question of the temperature dependence of the thermal diffusivity. Results from the various experiments cannot yet be described with a consistent physical picture. However, the importance of non-local effects is established in stellarators and tokamaks, although observed in different types of perturbation experiments. In stellarators and tokamaks, fluctuation measurements in the scrape-off layer are consistent with drift-wave-like turbulence being responsible for anomalous transport. In the core, density fluctuation amplitudes increase together with the diffusivity when the density is increased but the two parameters show opposite trends with increasing heating power. It turns out that transport in the two classes of devices is more alike than it has previously appeared. This indicates that the strong toroidal plasma current, major rational values of the rotational transform inside the plasma or strong magnetic shear are not the central elements of a theoretical model for anomalous transport in fusion plasmas. (166 refs.)
Todd, T. N. and C. G. Windsor (1998). "Progress in magnetic confinement fusion research." Contemporary Physics 39(4): 255-282.
This article reviews the present status of research into controlled fusion using magnetic confinement of ionized gas or plasma. It concentrates on the world's leading magnetic configurations for this purpose, comprising the conventional tokamak, stellarator, reversed field pinch and spherical tokamak. These configurations are described in some detail, along with an introduction to the basics of this line of fusion research, the progress so far and the principal scientific and technical problems outstanding. The environmental implications are briefly addressed in the context of some power generation alternatives and every-day radiation exposures.
Wolf, G. H. (1998). "Overview of tokamak results." Fusion Technology
3rd Carolus Magnus Summer School on Plasma Physics 33(2T): 357-72.
We discuss MHD, resistivity, current profiles, sawteeth, plasma facing materials, heating, current drive and bootstrap, confinement and scaling. (97 refs.)
Wong, H. V. and H. L. Berk (1998). "Growth and saturation of toroidal Alfven eigenmode modes destabilized by ion cyclotron range of frequency produced tails." Physics of Plasmas 5(7): 2781-96.
The linear growth rates of TAE (toroidal Alfven eigenmode) modes destabilized by ICRF (ion cyclotron range of frequency) heating are calculated over a range of plasma parameters. Nonlinear saturation of a single unstable mode is investigated both analytically and numerically when wave-particle trapping is the dominant saturation mechanism. A numerical code has been developed based on a reduced resonance description of the wave-particle interaction (using a Hamiltonian formalism). A delta-f algorithm was incorporated to allow a low-noise description of mode evolution with particle sources and sinks present. The numerically observed saturation amplitudes correlate well with theoretical predictions to within 20%. Self-excited frequency sweeping resulting from the excitation of many simultaneous wave-particle resonances at different energies is demonstrated and explained as an extension of previous published theory [Berk et al., Phys. Lett. A 234, 213 (1997)]. (26 refs.)
Andrew, P., D. Brennan, et al. (1999). "Tritium recycling and retention in JET." Journal of Nuclear Materials
13th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices 269: 153-159.
JET's 1997 Deuterium Tritium Experiment (DTE1) allows a detailed study of hydrogenic isotope recycling and retention in a pumped divertor configuration relevant to ITER. There appear to be two distinct forms of retained tritium. (1) A dynamic inventory which controls the fueling behaviour of a single discharge, and in particular determines the isotopic composition. This is shown to be consistent with neutral particle implantation over the whole vessel surface area. (2) A continually growing inventory, which plays a small role in the particle balance of a single discharge, but ultimately dominates the hydrogenic inventory for an experimental campaign comprising thousands of pulses. This will be the dominant retention mechanism in long-pulse devices like ITER. The JET retention scaled-up to ITER proportions suggests that ITER may reach its tritium inventory limit in less than 100 pulses. (C) 1999 JET Joint Undertaking, published by Elsevier Science B.V. All rights reserved.
Bell, M. G., R. E. Bell, et al. (1999). "Core transport reduction in tokamak plasmas with modified magnetic shear." Plasma Physics and Controlled Fusion
25th EPS Conference on Controlled Fusion and Plasma Physics / International Congress on Plasma Physics 41: A719-31.
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>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. (41 refs.)
Burrell, K. H. (1999). "Tests of causality: Experimental evidence that sheared ExB flow alters turbulence and transport in tokamaks." Physics of Plasmas
American-Physical-Society Centennial Meeting 6(12): 4418-4435.
A prime goal in physics research is the development of theories which have the universality needed to explain a wide range of observations. Developed over the past decade, the model of turbulence decorrelation and stabilization by sheared ExB flow has the universality needed to explain the turbulence reduction and confinement improvement seen in the edge and core of a wide range of magnetic confinement devices. Because the ExB shear, turbulence, and transport are all intimately intertwined in multiple feedback loops, devising experiments to test whether ExB shear causes a change in turbulence and transport has been a major challenge for experimentalists. Over the past five years, there have been at least four clear demonstrations of causality performed in tokamak plasmas, both at the plasma edge on Doublet III-D (DIII-D) [Plasma Physics and Controlled Fusion Research 1985 (International Atomic Energy Agency, Vienna, 1986) Vol. I, p.159] and Tokamak Experiment for Technologically Oriented Research (TEXTOR) [Plasma Physics and Controlled Nuclear Fusion Research 1990 (International Atomic Energy Agency, Vienna, 1991) Vol. I, p. 473] and further into the plasma core on DIII-D and Tokamak Fusion Test Reactor [Phys. Plasmas 5, 1577 (1998)]. This paper discusses these tests in detail; the results agree with the expectations from the ExB shear model. This paper also discusses similarities between flow shear effects in plasmas and in neutral fluids and provides examples of flow shear reduction of turbulence in neutral fluids under the proper conditions. (C) 1999 American Institute of Physics. [S1070-664X(99)03312-1].
Campbell, D. J. (1999). "Physics and goals of RTO/RC-ITER." Plasma Physics and Controlled Fusion
26th European-Physical-Society Conference on Controlled Fusion and Plasma Physics 41: B381-94.
The ITER design activities are now focused on the development of a design for a tokamak burning plasma experiment which will achieve Q=10 in long-pulse, inductively driven plasmas and which aims to demonstrate steady-state operation at Q=5 using non-inductive current drive. The device is designed to allow operational flexibility in plasma equilibrium and operating scenarios, and it is intended that the possibility of ignited operation under favourable plasma scenarios should nor be excluded. A final design and parameter set has nor yet been selected, but considerable progress has been made in developing a series of candidate designs which allow the critical physics and engineering issues to be explored. Here the physics underlying the design options and performance projections, as well as the physics issues to be addressed in the proposed experiment, are analysed. (28 refs.)
Costley, A. E. and I. J. C. T. H. Teams (1999). "Key issues in ITER diagnostics: Problems and solutions (invited)." Review of Scientific Instruments
12th Topical Conference on High-Temperature Plasma Diagnostics 70(1): 391-2.
The key problems associated with designing diagnostic systems for the International Thermonuclear Experimental Reactor (ITER) are identified and representative solutions are described. The plans for dealing with some outstanding issues are briefly presented. The detailed work is specific to the ITER, but much of what has been learned will be applicable to any reactor grade tokamak plasma. (30 refs.)
Dendy, R. O. and J. G. Kirk (1999). "Energetic particles in plasma astrophysics." Plasma Physics and Controlled Fusion
25th EPS Conference on Controlled Fusion and Plasma Physics / International Congress on Plasma Physics 41: A427-35.
Cosmic ray and gamma-ray astrophysics provides an arena where concepts drawn from many branches of plasma physics can assist in answering questions of fundamental-indeed cosmological-significance. In recent years, striking advances have been made in the quality and resolution of the data available in this field. The present generation of cosmic ray detection arrays and space-based X-ray and gamma-ray observatories, coupled to measurements in other wavebands, are yielding information which invites detailed quantitative interpretation in terms of plasma processes. In parallel with these developments, a systematic picture of the behaviour of energetic particle populations in magnetically confined plasmas has emerged from the deuterium tritium experimental campaigns on JET and TFTR, and the acceleration of electrons to highly relativistic energies has been demonstrated in laser-plasma interaction experiments. This review is written primarily to assist the deployment in energetic particle astrophysics of methodologies used for studying high-temperature fusion plasmas: three topics are examined. What are the major plasma-oriented questions in energetic particle astrophysics? What solutions have been proposed? How can plasma physics assist in deepening, strengthening and extending current understanding?. (62 refs.)
Desideri, D., G. Serianni, et al. (1999). "A fast moving diagnostic system on the RFX experiment." Review of Scientific Instruments
12th Topical Conference on High-Temperature Plasma Diagnostics 70(1): 403-2.
A fast rotating graphite arm equipped with two Langmuir probes has been developed to investigate the temperature, density, and plasma potential profiles at the edge in the reversed field pinch experiment RFX. The system, driven by an electromagnetic torque, allows a 5 cm radial insertion in about 20 ms with a magnetic field of 0.3 T. The motion of the arm takes place on a plane having an angle of 45 degrees with respect to the direction of the poloidal magnetic field. The probe collecting area is constant during the movement. The initial and final positions of the graphite arm are monitored by electrical contacts. To reconstruct the motion of the equipment, an equivalent lumped-constant network has been developed. The equivalent resistance and inductance are measured before each discharge. By this system radially resolved measurements of edge plasma parameters in a single RFX plasma discharge have been obtained. (7 refs.)
Desideri, D. and M. Bagatin (1999). "A new linear amplifier for the RFX electron energy analyzer." Review of Scientific Instruments
12th Topical Conference on High-Temperature Plasma Diagnostics 70(1): 407-2.
A simple low cost novel linear amplifier has been developed for the electron energy analyzer employed on the reversed field pinch experiment RFX. The amplifier is designed to provide a 1 kV output voltage from a 10 V input. The current rating is +or-50 mA. It utilizes a configuration of four blocks in series, each of them with an output totem-pole configuration. The rise and fall times are of a few milliseconds. A fast protection against output short circuit has been implemented. The amplifier has been used in various experimental campaigns on RFX and multiple time resolved measurements of the edge electron energy distribution in a single plasma discharge have been obtained. (10 refs.)
Eriksson, L. G., M. J. Mantsinen, et al. (1999). "Theoretical analysis of ICRF heating in JET DT plasmas." Nuclear Fusion 39(3): 337-52.
A number of experiments with heating of DT plasmas using ICRF waves have been carried out at JET. The results of these experiments have been analysed by comparing experimentally measured quantities with the results of numerical simulations. In particular, four scenarios have been examined: (a) heating of minority (~5-20%) deuterons at the fundamental ion cyclotron frequency, omega = omega {sub cD}; (b) second harmonic heating of tritium, omega =2 omega {sub cT}; (c) fundamental minority heating of {sup 3}He with a few per cent of {sup 3}He; (d) second harmonic heating of deuterium, omega =2 omega {sub cD}. An important aim of the analysis was to assess whether the present understanding of the ICRF physics is adequate for predicting the performance of ICRF in DT plasmas. In general, good agreement between experimental results and simulations was found which increases the confidence in predictions of the impact of ICRF heating in future reactors. However, when a relatively high deuterium concentration was used in the omega = omega {sub cD} scenario, discrepancies were observed. In order to increase confidence in the simulations, the sensitivity of the simulation results to various plasma parameters has been studied. (27 refs.)
Fowler, T. K. (1999). "Nuclear power-fusion." Reviews of Modern Physics 71(2): S456-9.
In the 1990s, experiments in tokamak magnetic fusion devices have finally approached "breakeven"-power out equal to power in-at fusion power levels exceeding 10 MW, and great progress has also been made with inertial-confinement fusion laser experiments. Based on these results, the requirements to achieve ignition and high-energy gain are now fairly clear for both approaches. This article focuses on developments in modern plasma physics that led to these achievements and outlines the historical development of the field. Topics include the stability of magnetic fields, field reconnection and the magnetic dynamo, turbulent heat transport, and plasma absorption of intense beams of light. The article concludes with a brief discussion of future research directions. (11 refs.)
Furno, I., H. Weisen, et al. (1999). "Fast bolometric measurements on the TCV tokamak." Review of Scientific Instruments 70(12): 4552-4556.
The design and first results are presented from a bolometric diagnostic with high temporal resolution recently installed on the TCV tokamak. The system consists of two pinhole cameras viewing the plasma from above and below at the same toroidal location. Each camera is equipped with an AXUV-16ELO linear array of 16 p-n junction photodiodes, characterized by a flat spectral sensitivity from ultraviolet to x-ray energies, a high temporal response (< 0.5 mu s), and insensitivity to low-energy neutral particles emitted by the plasma. This high temporal resolution allows the study of transient phenomena such as fast magnetohydrodynamic (MHD) activity hitherto inaccessible with standard bolometry. In the case of purely electromagnetic radiation, good agreement has been found when comparing results from the new diagnostic with those from a standard metal foil bolometer system. This comparison has also revealed that the contribution of neutrals to the foil bolometer measurements can be extremely important under certain operating conditions, precluding the application of tomographic techniques for reconstruction of the radiation distribution. (C) 1999 American Institute of Physics. [S0034-6748(99)04512-8].
Garofalo, A. M., A. D. Turnbull, et al. (1999). "Stabilization of the external kink and control of the resistive wall mode in tokamaks." Physics of Plasmas
40th Annual Meeting of the Division of Plasma Physics of the American-Physical-Society 6(5): 1893-8.
One promising approach to maintaining stability of high beta tokamak plasmas is the use of a conducting wall near the plasma to stabilize low-n ideal magnetohydrodynamic instabilities. However, with a resistive wall, either plasma rotation or active feedback control is required to stabilize the more slowly growing resistive wall modes (RWMs). Previous experiments have demonstrated that plasmas with a nearby conducting wall can remain stable to the n=1 ideal external kink above the beta limit predicted with the wall at infinity. Recently, extension of the wall stabilized lifetime tau {sub L} to more than 30 times the resistive wall time constant tau {sub w} and detailed, reproducible observation of the n=1 RWM have been possible in DIII-D [Plasma Physics and Controlled Fusion Research (International Atomic Energy Agency, Vienna, 1986), p. 159] plasmas above the no-wall beta limit. The DIII-D measurements confirm characteristics common to several RWM theories. The mode is destabilized as the plasma rotation at the q=3 surface decreases below a critical frequency of 1-7 kHz (~1% of the toroidal Alfven frequency). The measured mode growth times of 2-8 ms agree with measurements and numerical calculations of the dominant DIII-D vessel eigenmode time constant tau {sub w}. From its onset, the RWM has little or no toroidal rotation ( omega {sub mode}<or= tau {sub w}{sup -1}<< omega {sub plasma}), and rapidly reduces the plasma rotation to zero. These slowly growing RWMs can in principle be destabilized using external coils controlled by a feedback loop. In this paper, the encouraging results from the first open loop experimental tests of active control of the RWM, conducted in DIII-D, are reported. (21 refs.)
Grazioso, R. F., A. S. Heger, et al. (1999). "Feasibility of using boron-loaded plastic fibers for neutron detection." Nuclear Instruments & Methods in Physics Research, Section A (Accelerators, Spectrometers, Detectors and Associated Equipment)
9th Symposium on Radiation Measurements and Applications 422(1/3): 59-63.
The results from simulations and laboratory experiments with boron-loaded plastic scintillating fibers as a nondestructive assay (NDA) tool are presented. Single and multi-clad fibers in three diameters of 0.25, 0.5, and 1 mm were examined for their application in neutron coincidence counting. For this application, the simulation results show that various configurations of boron-loaded plastic scintillating fibers have a die-away time ( tau ) of 12 mu s with an efficiency ( epsilon ) of 50%. For a comparable efficiency, {sup 3}He proportional tubes have a typical die-away time of 50 mu s. The shortened die-away time can reduce the relative error for measurement of similar samples by up to 50%. Plastic scintillating fibers (PSF) also offer flexible configurations with the potential to discriminate between signals from gamma-ray and neutron events. Quantitative calculations and experiments have been conducted to determine the light output, evaluate the noise, quantify light attenuation, and determine neutron detection efficiency. Current experimental data support the analytical results that boron-loaded plastic fibers can detect thermal neutrons. (24 refs.)
Hawryluk, R. J. and T. Grp (1999). "Results from D-T experiments on TFTR and implications for achieving an ignited plasma." Philosophical Transactions of the Royal Society London, Series A (Mathematical, Physical and Engineering Sciences)
Meeting of the Royal-Society on the Approach to Ignited Plasma 357(1752): 443-69.
Progress in the performance of tokamak devices has enabled not only the production of significant bursts of fusion energy from deuterium-tritium plasmas in the Tokamak Fusion Test Reactor (TFTR) and the Joint European Torus (JET) but, more importantly, the initial study of the physics of burning magnetically confined plasmas. As a result of the worldwide research on tokamaks, the scientific and technical issues for achieving an ignited plasma are better understood and the remaining questions more clearly defined. The principal research topics that have been studied on TFTR are transport, magnetohydrodynamic stability and energetic-particle confinement. The integration of separate solutions to problems in each of these research areas has also been of major interest. Although significant advances-such as the reduction of turbulent transport by means of internal transport barriers, identification of the theoretically predicted bootstrap current, and the study of the confinement of energetic fusion alpha -particles-have been made, interesting and important scientific and technical issues remain for achieving a magnetic fusion energy reactor. In this paper, the implications of the TFTR experiments for overcoming these remaining issues will be discussed. (88 refs.)
Hill, K. W., M. G. Bell, et al. (1999). "Highly radiative plasmas for local transport studies and power and particle handling in reactor regimes." Nuclear Fusion
17th IAEA Fusion Energy Conference 39(11Y): 1949-1954.
To study the applicability of artificially enhanced impurity radiation for mitigation of the plasma-limiter interaction in reactor regimes, krypton and xenon gases were injected into TFTR super-shots and high internal inductance plasmas. At NBI powers P-B greater than or equal to 30 MW, carbon influxes ('blooms') were suppressed, leading to improved energy confinement and neutron production in both D and DT plasmas, and the highest DT fusion energy production (7.6 MJ) in a TFTR pulse. Comparisons of the measured radiated power profiles with predictions of the MIST impurity transport code have guided studies of highly radiative plasmas in ITER. The response of the electron and ion temperatures to greatly increased radiative losses from the electrons was used to study thermal transport mechanisms. A change in the radial electric field E-r is associated with the improved confinement observed.
Hill, K. W., S. D. Scott, et al. (1999). "Tests of local transport theory and reduced wall impurity influx with highly radiative plasmas in the Tokamak Fusion Test Reactor." Physics of Plasmas 6(3): 877-84.
The electron temperature (T{sub e}) profile in neutral beam-heated supershot plasmas (T{sub e0}~6-7 keV ion temperature T{sub i0}~15-20 keV, beam power P{sub b}~16 MW) was remarkably invariant when radiative losses were increased significantly through gas puffing of krypton and xenon in the Tokamak Fusion Test Reactor [McGuire et al., Phys. Plasmas 2, 2176 (1995)]. Trace impurity concentrations (n{sub z}/n{sub e}~10{sup -3}) generated almost flat and centrally peaked radiation profiles, respectively, and increased the radiative losses to 45%-90% of the input power (from the normal ~25%). Energy confinement was not degraded at radiated power fractions up to 80%. A 20%-30% increase in T{sub i}, in spite of an increase in ion-electron power loss, implies a factor of ~3 drop in the local ion thermal diffusivity. These experiments form the basis for a nearly ideal test of transport theory, since the change in the beam heating power profile is modest, while the distribution of power flow between (1) radiation and (2) conduction plus convection changes radically and is locally measurable. The decrease in T{sub e} was significantly less than predicted by two transport models and may provide important tests of more complete transport models. At input power levels of 30 MW, the increased radiation eliminated the catastrophic carbon influx (carbon "bloom") and performance (energy confinement and neutron production) was improved significantly relative to that of matched shots without impurity gas puffing. (58 refs.)
Hillis, D. L., J. T. Hogan, et al. (1999). "Investigation of tritium pathways in the Joint European Torus (JET) tokamak." Physics of Plasmas
40th Annual Meeting of the Division of Plasma Physics of the American-Physical-Society 6(5): 1985-94.
The neutral tritium concentration in the subdivertor region of JET [P.-H. Rebut, R.J. Bickerton, and B.E. Keen, Nucl. Fusion 25, 1011 (1985)] is measured during deuterium-to-tritium changeover experiments with a novel species-selective Penning gauge coupled to a high-resolution spectrometer. The subdivertor measurements, when compared with edge and strike point values, are a sensitive characterization of the status of the wall saturation. The neutral transport code (EIRENE) [D. Reiter, Forschungszentrum, Juelich: Report Juel-2599 (1992)] and a wall hydrogen trapping and diffusion code (WDIFFUSE) [J. Hogan, R. Maingi, P. Mioduszewski et al., J. Nucl. Mater. 241, 612 (1997)] evaluate the wall tritium recycling coefficients (R{sub T}) and compare them with quantitative, testable models for the dynamic exchange between recycling surfaces and the edge/pedestal region. Using the dynamic inventory model, tritium transport following the injection of trace amounts of tritium has been investigated for JET high-confinement mode (H-mode) discharges with Edge Localized Modes (ELMs). Upon explicitly treating the ELMs as instantaneous magnetohydrodynamic events, residual (intra-ELM) radial diffusivities are reduced and can be compared with neoclassical levels. (30 refs.)
Hoffman, A. L., P. Gurevich, et al. (1999). "Inductive field-reversed configuration accelerator for tokamak fueling." Fusion Technology 36(2): 109-25.
Compact toroids can be used for fueling other fusion devices by accelerating them to high enough velocities to penetrate strong magnetic fields. In the simplest analysis, the kinetic energy density of a flux-excluding object 1/2 rho v{sup 2} must exceed the magnetic field energy density B{sup 2}/2 mu {sub 0} of the field to be pushed aside. Field reversed configurations (FRCs) are a type of compact toroid that are particularly efficient for this application due to their high density and thus lower required energy per unit mass. FRCs are also formed and accelerated inductively, thus minimizing possible impurity contamination. The Tokamak Refueling by Accelerated Plasmoids (TRAP) experiment was built to develop the inductive acceleration method and test the ability of high-velocity FRCs to penetrate transverse magnetic fields. Simple models have been developed for both the acceleration and penetration processes to determine fueler parameters required for a given tokamak field. Experimental results are given for the acceleration process. Half-milligram FRCs with number densities of 10{sup 22} m{sup -3} were accelerated to velocities of 200 km/s, sufficient to fuel tokamaks with Tesla magnetic fields. The technology is easily extendable to much higher FRC densities and velocities, sufficient to fuel the largest highest-field tokamaks. (17 refs.)
Idomura, Y., S. Tokuda, et al. (1999). "Gyrokinetic theory of slab ion temperature gradient mode in negative shear tokamaks." Physics of Plasmas 6(12): 4658-71.
In a slab configuration modeling the negative shear tokamak, a linear analysis of the ion temperature gradient (ITG) mode is performed based on a gyrokinetic integral eigenvalue equation with retaining the full finite Larmor radius (FLR) effect. Numerical results show that characteristics of the slab ITG mode are greatly changed in the negative shear configuration. When a single mode-rational surface exists at the q{sub min}-surface, the steep ion temperature gradient and the associated FLR effect produce an asymmetric mode structure with respect to the mode-rational surface. In the weak magnetic shear region near the q{sub min}-surface, an unstable region of the ITG mode is divided into two separate regions, which are located in both sides of the mode-rational surface. Since the ion Larmor radius varies significantly between these regions, low-k{sub y} (high-k{sub y}) modes appear in the high-T{sub i} (low-T{sub i}) side due to the asymmetric FLR effect, where k{sub y} corresponds to the wavenumber in the poloidal direction. When double mode-rational surfaces exist near the q{sub min}-surface, it is found that the ITG mode becomes unstable in the interior region between the two mode-rational surfaces. Since the shear stabilization disappears in this region, the unstable k{sub y} region spreads up to an extremely short wavelength region or k{sub y} rho {sub ti}<or=10, where rho {sub ti} is the ion Larmor radius. (31 refs.)
Innocente, P., B. Boscolo, et al. (1999). "Three-dimensional time-resolved H pellet trajectory reconstruction in RFX by position sensitive detector H{sub alpha } diagnostic." Review of Scientific Instruments
12th Topical Conference on High-Temperature Plasma Diagnostics 70(1): 943-2.
Hydrogen pellets injected in reversed field pinches (RFPs) experience large poloidal and toroidal accelerations along magnetic field lines due to the rocket effect caused by asymmetric ablation processes. The 3D pellet trajectory and the global ablation rate is diagnosed in RFX by an absolutely calibrated multiple position sensitive detector (PSD) system looking at the H{sub alpha } radiation coming from the pellet ablation cloud. The design of the diagnostic is presented. Then the performance is highlighted by cross-comparison between different PSD measurements and with time-integrated images from a CCD camera. Overall accuracy of the order of a few millimeters is achieved, which makes the instrument well-suited to assist transport studies (particle deposition), to estimate the radial profile of fast electron tails (derived from the modulus of pellet acceleration) and to obtain information on internal magnetic field pitch (from the direction of pellet acceleration). (6 refs.)
Jacquinot, J., S. Putvinski, et al. (1999). "Chapter 5: Physics of energetic ions." Nuclear Fusion 39(12): 2471-2494.
Physics knowledge (theory and experiment) in energetic particles relevant to design of a reactor scale tokamak is reviewed, and projections for ITER are provided in this Chapter of the ITER Physics Basis. The review includes single particle effects such as classical alpha particle heating and toroidal field ripple loss, as well as collective instabilities that might be generated in ITER plasmas by energetic alpha particles. The overall conclusion is that fusion alpha particles are expected to provide an efficient plasma heating for ignition and sustained burn in the next step device. The major concern is localized heat loads on the plasma facing components produced by alpha particle loss, which might affect their lifetime in a tokamak reactor.
Jarboe, T. R. (1999). "Steady inductive helicity injection and its application to a high-beta spheromak." Fusion Technology 36(1): 85-91.
A steady inductive helicity injection (SIHI) method is described that has the following properties: (a) helicity is injected at a nearly constant rate; (b) neither magnetic energy nor helicity flow out of plasma at any time; (c) no open field lines penetrate the walls; (d) the equilibrium is produced in a close-fitting flux conserver; (e) a rotating magnetic structure is produced directly; and (f) in the frame of the rotating field, the current profile is nearly time independent and nearly optimum for the application discussed. SIHI can be applied to any toroidal plasma. Application of SIHI to a high-beta spheromak is described. (35 refs.)
Jassby, D. L., J. D. Strachan, et al. (1999). "TFTR experiments with 2% tritium beam fuelling." Nuclear Fusion 39(2): 189-208.
Tritium has been injected into TFTR supershots using a gas feed containing 2% tritium and 98% deuterium in one or more neutral beam sources. A total of 38 beam-injected tritium shots required about 400 Ci of tritium, of which less than 5% actually entered the plasma. The ratio of D-T to D-D neutron rates varied from about 0.5 to 4.5. A peak neutron rate of 7.7*101{sup 6} D-T n/s together with 1.7*10{sup 16} D-D n/s was attained with eight 2% tritium beams delivering 18.7 MW. The measured D-T neutron rates are extrapolated to fusion powers in 50:50 D-T operation with similar plasma parameters. The experiments have also provided data for the D-T calibration of neutron and lost alpha particle detectors and for optimizing tritium beam injection. With the aid of a multichannel neutron collimator, these experiments have enabled the comparison of radial profiles of D-T neutron production and global neutron output with the predictions of simulation codes. (38 refs.)
Kato, S., M. Watanabe, et al. (1999). "Laboratory experiment on lithium chemistry and its application to effective wall conditioning." Journal of Nuclear Materials
13th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices 269: 406-411.
Laboratory studies on the physical chemistry of lithium wall conditioning effects observed in fusion reactors are reported, focusing on a role of lithium in dramatic decrease in carbon impurities in TFTR. Auger depth analysis of lithium-deposited graphite shows deep intercalation of lithium into graphite at room temperature. Helium ion bombardment (150-200 eV) on the lithium-saturated graphite leads to preferential sputtering of lithium and reduction of the net sputtering yield of carbon by an order of magnitude. Finally, general precautions to achieve reproducible lithium effects in fusion machines are described. (C) 1999 Elsevier Science B.V. All rights reserved.
Kaye, S. M., M. Ono, et al. (1999). "Physics design of the National Spherical Torus Experiment." Fusion Technology 36(1): 16-37.
The mission of the National Spherical Torus Experiment (NSTX) is to prove the principles of spherical torus physics by producing high- beta plasmas that are noninductively sustained and whose current profiles are in steady state. The NSTX will be one of the first ultralow-aspect-ratio tori (R/a<or=1.3) to operate at high power (P{sub input} up to 11 MW) to produce high- beta {sub t} (25 to 40%), low-collisionality, high-bootstrap-fraction (<or=70%) discharges. Both radio-frequency and neutral beam heating and current drive will be employed. Built into the NSTX is sufficient configurational flexibility to study a range of operating space and the resulting dependences of the confinement, micro- and magnetohydrodynamic stability, and particle- and power-handling properties. NSTX research will be carried out by a nationally based science team. (42 refs.)
Keilhacker, M., M. L. Watkins, et al. (1999). "D-T experiments in the JET Tokamak." Journal of Nuclear Materials
13th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices 269: 1-13.
During the second half of 1997, JET carried out a broad-based series of D-T experiments (DTE1) producing a total of 675MJ of fusion energy. A large scale tritium supply and processing plant, the first of its kind, allowed the repeated use of the 20 g tritium on site, supplying a total of 99.3 g of tritium to the machine. After DTE1, the tritium inventory in the torus remained a factor of about three higher than expected from the Preliminary Tritium Experiment in 1991, and this is thought to be related to tritium-saturated carbon films on surfaces which are shadowed from erosion by the plasma. During DTE1 records were set for peak (16.1 MW) and quasi steady-state (4 MW for 4 s) fusion power and for the ratio of fusion power to input power (0.62; if a similar plasma could be obtained in steady-state, the Q would be 0.94 +/- 0.17). Alpha particle heating was clearly demonstrated and shown to be consistent with classical expectations. In the optimised shear mode of operation internal transport barriers were established for the first time in D-T, with a threshold power roughly equal to that in D-D. ELMy H-mode studies in D-T have considerably strengthened the basis for predicting the heating requirements and performance of ITER. Candidate ICRF heating schemes for ITER were successfully tested and the relevant simulation codes validated. With regard to isotope effects in ELMy H-modes, the ITER scaling for the I-I-mode threshold power had to be modified to include an inverse mass dependence (approximate to A(-1)), while energy transport showed little dependence on isotope and seems to involve different physics in the edge and the core of the plasma. JET confinement data obtained under conditions which were identical to ITER in most dimensionless parameters scale close to gyro-Bohm and predict ignition for ITER provided the required densities can be reached. (C) 1999 JET Joint Undertaking, published by Elsevier Science B.V. All rights reserved.
Keinigs, R. K., W. L. Atchison, et al. (1999). "One- and two-dimensional simulations of imploding metal shells." Journal of Applied Physics 85(11): 7626-34.
We report results of one- and two-dimensional (2D) magnetohydrodynamic simulations of imploding, cylindrical metal shells. One-dimensional simulations are used to calculate implosion velocities of heavy liners driven by 30 MA currents. Accelerated by the j*B force, 45 g aluminum/tungsten composite liners achieve velocities on the order of 13 km/s. Used to impact a tungsten target, the liner produces shock pressures of approximately 14 Mbar. The first 2D simulations of these liners are also described. These simulations have focused on two problems: (1) the interaction of the liner with the electrically conducting glide planes, and (2) the effect of realistic surface perturbations on the dynamics of the implosion. The former interaction is confined primarily to the region of the contact point between the liner and glide plane, and does not seriously affect the inner liner surface. However a 0.2 mu m surface perturbation has a significant effect on the implosion dynamics. (13 refs.)
Kilkenny, J. D., E. M. Campbell, et al. (1999). "The role of the National Ignition Facility in energy production from inertial fusion." Philosophical Transactions of the Royal Society London, Series A (Mathematical, Physical and Engineering Sciences)
Meeting of the Royal-Society on the Approach to Ignited Plasma 357(1752): 533-54.
The 1993 declassification of virtually all inertial confinement fusion (ICF) target information relevant to fusion energy development, and demonstrable successes in the physics and technology related to ICF, have laid the ground work for a development plan for an Inertial Fusion Energy (IFE) programme. The ICF programme, funded by the Defense Program in the USA. Has clearly demonstrated there is sufficient confidence in ignition and gain to proceed with construction of the National Ignition Facility, which will test the detailed physics of targets suitable for IFE. In September 1998, the facility was about 40% complete and on schedule and within budget for completion in 2003. X-ray-drive ignition is planned for 2007, followed by direct-drive ignition experiments. The other major elements of an IFE development programme, namely, driver, target factory, and target chamber developments can be investigated separately in affordable programmes. Although much work remains, there are concepts for adequately high driver efficiency and target gain, target cost and target chamber survivability to make an exploratory programme in IFE attractive. (27 refs.)
Litaudon, X., T. Aniel, et al. (1999). "Electron and ion internal transport barriers in Tore Supra and JET." Plasma Physics and Controlled Fusion
25th EPS Conference on Controlled Fusion and Plasma Physics / International Congress on Plasma Physics 41: A733-46.
Formation of core regions in Tore Supra and JET tokamaks with reduced transport coefficients is reported. Characteristics of the enhanced confinement regions and the physics process involved in their formation and maintenance should be considered separately when the electron or ion components are predominantly heated. In Tore Supra and JET, central electron temperature transitions are observed by injecting lower hybrid waves at modest power levels during the current ramp-up phase of the discharges. Transport analyses stress the importance of the low magnetic shear in the core to explain the anomalous electron transport reduction. With high-power dominant ion heating schemes in JET (neutral beam injection and ion cyclotron resonance heating), internal transport barriers have been obtained in plasmas fuelled with a mixture of deuterium-tritium (D-T) ions leading to a successful production of fusion power (8.2 MW) in this regime. Similar additional power levels to those applied in pure deuterium (D-D) plasmas are required to establish internal transport barriers in D-T plasmas. In D-D and D-T plasmas, ion thermal diffusivities are reduced close to their neoclassical levels in the plasma core and electron thermal diffusivities decrease by one order of magnitude at midplasma radius. The combined role of magnetic shear and E*B velocity shear can explain the formation and evolution of plasma core regions with low energy transport coefficients. (40 refs.)
Manheimer, W. (1999). "Back to the future: the historical, scientific, naval, and environmental case for fission fusion." Fusion Technology 36(1): 1-15.
A return to fission fusion, and especially the development of the thorium cycle, is proposed as a means to revitalize magnetic fusion research. Recent history is analyzed, causes are sought for the current state of fusion research, and possibilities for how its prospects can be improved are examined. Recent tokamak results are also analyzed and the conclusion is reached that a research tokamak reactor could now be built that could generate significant amounts of nuclear fuel. Finally, possible naval involvement and environmental issues are discussed. (68 refs.)
Maqueda, R. J., G. A. Wurden, et al. (1999). "The new infrared imaging system on Alcator C-Mod." Review of Scientific Instruments
12th Topical Conference on High-Temperature Plasma Diagnostics 70(1): 734-2.
A new infrared imaging system has been installed on Alcator C-Mod. This system uses an Amber Radiance 1 IR video camera (filtered to the 4.2-4.4 mu m band) to view a 30 cm*30 cm region of the lower divertor from above by means of a re-entrant 5-m long ZnSe based periscope. Capture of the standard 30 Hz video frames (8-bit) and camera control are performed remotely over fiber optic links by a Windows 95 PC, using a MuTech MV-1000 video grabber board. Plans are under way to directly capture the 60 Hz, 12-bit, 256*256 pixel images using a digital video camera interface with a fiber optic link from EDT (Beaverton, Oregon). Preliminary results show that during nondisruptive discharges no substantial surface temperature increase is observed on the upper sections of the divertor, with the exception of "hot spots," although occasionally, increased heating in toroidal bands is seen. Bands can also be observed after disruptions that result in a downward movement of the plasma. (3 refs.)
Maqueda, R. J. and G. A. Wurden (1999). "Images of plasma disruption effects in the Tokamak Fusion Test Reactor." IEEE Transactions on Plasma Science 27(1): 112-13.
Fast-framing imaging of visible radiation from magnetically confined plasmas has lately become a useful tool for both machine operation and physics studies. Using an intensified, commercial Kodak Ektapro imaging system, the effects of a plasma disruption were observed in the Tokamak Fusion Test Reactor (TFTR). The high-energy "runaway" electrons created soon after the disruption collide with the plasma facing components damaging this surface and producing a shower of debris that traverses the toroidal vessel and "falls" over the inner bumper limiter. (4 refs.)
Matsuda, T., T. Tsugita, et al. (1999). "Systems for remote participation in JT-60 experiments." Review of Scientific Instruments
12th Topical Conference on High-Temperature Plasma Diagnostics 70(1): 502-2.
Since 1996, the remote participation in JT-60 experiments has been successfully conducted in collaboration with JAERI, Los Alamos National Laboratory and Princeton Plasma Physics Laboratory. The remote analysis system is a computer complex consisting of an analysis server and a data server for access to JT-60 data. Collaborators of the remote participation can analyze JT-60 data by using this system through the Internet. For the remote diagnostic system, an X Window control program has been customized to communicate via the overseas line using a UNIX workstation at the remote site and one in the JT-60 control room. It has a feature of real-time remote control of diagnostic equipment and remote access to CAMAC data. Both systems are used together with the ISDN-based video-conferencing system for real-time communication between the remote laboratory and the JT-60 control room. (6 refs.)
Matsuda, T., T. Aoyagi, et al. (1999). "Recent developments in the JT-60 data processing system." Fusion Engineering and Design
Workshop on Data Acquisition and Management for Fusion Research 43(3/4): 285-91.
The JT-60 data processing system was originally a large computer complex system including a lot of micro-computers, several mini-computers, and a mainframe computer. Recently, several improvements have been made to the original system to modernize the system. Many sub-systems composed of aged mini-computers have been replaced with workstations by utilizing recent progress in computer and network technologies. The system can handle ~300 MB of raw data per discharge, which is three times larger than the amount in the original system. These improvements have been applied to develop element technologies necessary to the remote participation in JT-60 experiments. A remote diagnostics control and monitoring system and a computer system for access to JT-60 data from the Internet are used together with video conferencing systems for a real-time communication. In 1996, the remote participation based on them was successfully demonstrated in collaboration with Japan Atomic Energy Research Institute, Los Alamos National Laboratory, and Princeton Plasma Physics Laboratory. (11 refs.)
Mirnov, S., J. Wesley, et al. (1999). "Chapter 8: Plasma operation and control." Nuclear Fusion 39(12): 2577-625.
Wall conditioning of fusion devices involves removal of desorbable hydrogen isotopes and impurities from interior device surfaces to permit reliable plasma operation. Techniques used in present devices include baking, metal film gettering: deposition of thin films of low-Z material, pulse discharge cleaning, glow discharge cleaning, radio frequency discharge cleaning, and in situ limiter and divertor pumping. Although wall conditioning techniques have become increasingly sophisticated, a reactor scale facility will involve significant new challenges, including the development of techniques applicable in the presence of a magnetic field and of methods for efficient removal of tritium incorporated into co-deposited layers on plasma facing components and their support structures. The current status of various approaches is reviewed, and the implications for reactor scale devices are summarized. Creation and magnetic control of shaped and vertically unstable elongated plasmas have been mastered in many present tokamaks. The physics of equilibrium control for reactor scale plasmas will rely on the same principles, but will face additional challenges; exemplified by the ITER/FDR design. (142 refs.)
Mirnov, S., J. Wesley, et al. (1999). "Chapter 3: MHD stability, operational limits and disruptions." Nuclear Fusion 39(12): 2251-389.
The present physics understandings of magnetohydrodynamic (MHD) stability of tokamak plasmas, the threshold conditions for onset of MHD instability, and the resulting operational limits on attainable plasma pressure (beta limit) and density (density limit), and the consequences of plasma disruption and disruption related effects are reviewed and assessed in the context of their application to a future DT burning reactor prototype tokamak experiment such as ITER. (357 refs.)
Najmabadi, F., J. Drake, et al. (1999). "The report of the subpanel to FESAC concerning alternative concepts." Journal of Fusion Energy 18(3): 161-87.
This is the July 1996 report of a subpanel of the US Department of Energy Fusion Energy Sciences Advisory Committee (FESAC), charged with reviewing the present status of fusion alternative concept development and the prospects for alternative concepts not only as fusion power systems but also the scientific contributions of alternative concept research to the fusion energy sciences program and to plasma science in general. (65 refs.)
Pamela, J. (1999). "Ten years of operation and developments on Tore Supra." Fusion Engineering and Design
20th Symposium on Fusion Technology (SOFT 20) 46(2/4): 313-22.
Tore Supra started operation 10 years ago, on April 1, 1988. This tokamak, today the third in size world wide, has several features representative of the needs of future fusion power production devices: it is equipped with superconducting toroidal field coils (NbTi cooled with He II at 1.8 K); actively cooled plasma facing components; non-inductive current drive systems and real time feedback systems for plasma control. These features provide unique capabilities for achieving high-performance long-pulse plasma discharges and studying several key physics issues in truly steady-state conditions. Several key experimental achievements could thereby be obtained, such as: (i) steady-state high bootstrap fraction discharges (I{sub BS}/I{sub p}50%, >or=5 s); (ii) long discharges sustained by non-inductive current drive, with enhanced confinement, in particular, 2 min discharges, with up to 280 MJ of energy coupled to the plasma, and fully non-inductive steady-state discharges held for duration of up to 70 s; (iii) highly radiating discharges with P{sub rad}/P{sub tot}60+or-10% and confinement as good as the ELMy H-Mode selected for ITER. These results could be achieved thanks to a tight integration between physics and technology. In line with these achievements, Tore Supra is now preparing its evolution, with the CIEL project, which aims at renewing the inner components of the machine. This should bring the overall power handling capability up to about 25 MW, with plasma facing components able to sustain power densities in the 10 MW m{sup -2} range. The pulse length capability of the tokamak should reach 1000 s. (39 refs.)
Perkins, F. W., D. E. Post, et al. (1999). "Chapter 1: Overview and summary." Nuclear Fusion 39(12): 2137-2174.
The ITER Physics Basis presents and evaluates the physics rules and methodologies for plasma performance projections, which provide the basis for the design of a tokamak burning plasma device whose goal is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. This Chapter summarizes the physics basis for burning plasma projections, which is developed in detail by the ITER Physics Expert Groups in subsequent chapters. To set context, the design guidelines and requirements established in the report of ITER Special Working Group 1 are presented, as are the specifics of the tokamak design developed in the Final Design Report of the ITER Engineering Design Activities, which exemplifies burning tokamak plasma experiments. The behaviour of a tokamak plasma is determined by the interaction of many diverse physics processes, all of which bear on projections for both a burning plasma experiment and an eventual tokamak reactor. Key processes summarized here are energy and particle confinement and the H-mode power threshold; MHD stability, including pressure and density limits, neoclassical islands, error fields, disruptions, sawteeth, and ELMs; power and particle exhaust, involving divertor power dispersal, helium exhaust, fuelling and density control, H-mode edge transition region, erosion of plasma facing components, tritium retention; energetic particle physics; auxiliary power physics; and the physics of plasma diagnostics. Summaries of projection methodologies, together with estimates of their attendant uncertainties, are presented in each of these areas. Since each physics element has its own scaling properties, an integrated experimental demonstration of the balance between the combined processes which obtains in a reactor plasma is inaccessible to contemporary experimental facilities: it requires a reactor scale device. It is argued, moreover, that a burning plasma experiment can be sufficiently flexible to permit operation in a steady state mode, with non-inductive plasma current drive, as well as in a pulsed mode where current is inductively driven. Overall, the ITER Physics Basis can support a range of candidate designs for a tokamak burning plasma facility. For each design, there mill remain a significant uncertainty in the projected performance, but the projection methodologies outlined here do suffice to specify the major parameters of such a facility and form the basis for assuring that its phased operation will return sufficient information to design a prototype commercial fusion power reactor, thus fulfilling the goal of the ITER project.
Perkins, F. W., A. Bondeson, et al. (1999). "Neoclassical islands, beta-limits, error fields and ELMs in reactor scale tokamaks." Nuclear Fusion
17th IAEA Fusion Energy Conference 39(11Y): 2051-2054.
An assessment is presented of the impact of recent magnetohydrodynamic research results on performance projections for reactor scale tokamaks as exemplified by the ITER Final Design Report (ITER/FDR) facility. For nominal ELMy H mode operation, the presence and amplitude of neoclassical tearing modes governs the achievable beta value. Recent work finds that the scaling of beta at which such modes onset agrees well with a polarization drift model, with the consequence that, with reasonable assumptions regarding seed island width, the mode onset beta will be lower in reactor scale tokamaks than in contemporary devices. Confinement degradation by such modes, on the other hand, depends on relative saturated island size which is governed principally by beta and secondarily by upsilon* effects on bootstrap current density. Relative saturated island size should be comparable in present and reactor devices. DT ITER demonstration discharges in JET exhibited no confinement degradation at the planned ITER operating value of beta(N) = 2.2. Theory indicates that electron cyclotron current drive can either stabilize these modes or appreciably reduce saturated island size. Turning to operation in candidate steady state, reverse shear, high bootstrap fraction configurations, wall stabilization of external kink modes is effective while the plasma is rotating but (so far) rotation has not been maintained. Recent error field observations in JET imply an error field size scaling that leads to a projection that the ITER/FDR facility will be somewhat more tolerant to error fields than thought previously. ICRF experiments on JET and Alcator C-Mod indicate that plasmas heated by central energetic particles have benign ELMs compared with the usual type 1 ELM of NBI heated discharges.
Petrov, M. P., R. Bell, et al. (1999). "Effective temperatures, sawtooth mixing, and stochastic diffusion ripple loss of fast H{sup +} minority ions driven by ion cyclotron heating in the Tokamak Fusion Test Reactor." Physics of Plasmas 6(6): 2430-6.
This paper presents studies of the H{sup +} minority ions driven by Ion Cyclotron Radio Frequency (ICRF) heating in the Tokamak Fusion Test Reactor (TFTR) [R.J. Hawryluk et al., Phys. Plasmas 5, 1577 (1998)] deuterium plasmas using primarily passive H degrees flux detection in the energy range of 0.2-1.0 MeV. The measured passive H{sup +} energy spectra are compared with active (Li pellet charge exchange) results. It is shown that in the passive mode the main donors for the neutralization of H{sup +} ions in this energy range are C{sup 5+} ions. The measured effective H{sup +} tail temperatures range from 0.15 MeV at an ICRF power of 2 MW to 0.35 MeV at 6 MW. Radial redistribution of ICRF-driven H{sup +} ions was detected when giant sawtooth crashes occurred during the ICRF heating. The redistribution affected ions with energy below 0.7-0.8 MeV. The sawtooth crashes displaces H{sup +} ions outward along the plasma major radius into the stochastic ripple diffusion domain where those ions are lost in ~10 msec. These observations are consistent with the model of the redistribution of energetic particles developed previously to explain the results of deuterium-tritium (DT) alpha-particle redistribution due to sawtooth oscillations observed in TFTR. The experimental data are also consistent with numerical simulations of H{sup +} stochastic ripple diffusion losses. (25 refs.)
Prager, S. C. (1999). "Dynamo and anomalous transport in the reversed field pinch." Plasma Physics and Controlled Fusion
25th EPS Conference on Controlled Fusion and Plasma Physics / International Congress on Plasma Physics 41: A129-42.
The reversed field pinch (RFP) is an effective tool in the study of the macroscopic consequences of magnetic fluctuations, such as the dynamo effect and anomalous transport. Several explanations exist for the dynamo (the self-generation of plasma current)-the magnetohydrodynamic dynamo, the kinetic dynamo and the diamagnetic dynamo. There is some experimental evidence for each, particularly from measurements of ion velocity and electron pressure fluctuations. Magnetic fluctuations are known to produce energy and particle flux in the RFP core. Current profile control is able to decrease fluctuation-induced transport by a factor of five. Improved confinement regimes are also obtained at deep reversal and, possibly, with flow shear. (53 refs.)
Putvinski, S., W. Heidbrink, et al. (1999). "Alpha-particle physics in tokamaks." Philosophical Transactions of the Royal Society London, Series A (Mathematical, Physical and Engineering Sciences)
Meeting of the Royal-Society on the Approach to Ignited Plasma 357(1752): 493-513.
Efficient plasma heating by energetic fusion alpha -particles is a key element of achieving ignition or high fusion gain regimes (Q>>1) in a tokamak reactor. The paper summarizes issues and reviews the latest theoretical and experimental results in the area of energetic particle physics in tokamaks. The discussion includes the classical physics of alpha -particle heating, effects of perturbations on single-particle confinement, collective instabilities driven by energetic alpha -particles, description of the potential alpha -particle loss channels and their impact on a design of the plasma-facing components in a tokamak reactor. Most of the extrapolations to tokamak reactors are done with the use of ITER parameters and operational scenarios as an example. (103 refs.)
Redi, M. H., S. H. Batha, et al. (1999). "Neoclassical simulations of fusion alpha particles in pellet charge exchange experiments on the Tokamak Fusion Test Reactor." Physics of Plasmas 6(7): 2826-33.
Neoclassical simulations of alpha particle density profiles in high fusion power plasmas on the Tokamak Fusion Test Reactor [Phys. Plasmas 5, 1577 (1998)] are found to be in good agreement with measurements of the alpha distribution function made with a sensitive active neutral particle diagnostic. The calculations are carried out in Hamiltonian magnetic coordinates with a fast, particle-following Monte Carlo code which includes the neoclassical transport processes, a recent first-principles model for stochastic ripple loss and collisional effects. New calculations show that monotonic shear alpha particles are virtually unaffected by toroidal field ripple. The calculations show that in reversed shear the confinement domain is not empty for trapped alphas at birth and allow an estimate of the actual alpha particle densities measured with the pellet charge exchange diagnostic. (22 refs.)
Ruskov, E., M. Bell, et al. (1999). "Fusion performance analysis of plasmas with reversed magnetic shear in the Tokamak Fusion Test Reactor." Physics of Plasmas 6(8): 3247-62.
A case for substantial loss of fast ions degrading the performance of tokamak fusion test reactor plasmas [Phys. Plasmas 2, 2176 (1995)] with reversed magnetic shear (RS) is presented. The principal evidence is obtained from an experiment with short (40-70 ms) tritium beam pulses injected into deuterium beam heated RS plasmas [Phys. Rev. Lett. 82, 924 (1999)]. Modeling of this experiment indicates that up to 40% beam power is lost on a time scale much shorter than the beam-ion slowing down time. Critical parameters which connect modeling and experiment are: The total 14 MeV neutron emission, its radial profile, and the transverse stored energy. The fusion performance of some plasmas with internal transport barriers is further deteriorated by impurity accumulation in the plasma core. (51 refs.)
Sharapov, S. E., D. Borba, et al. (1999). "Stability of alpha particle driven Alfven eigenmodes in high performance JET DT plasmas." Nuclear Fusion 39(3): 373-88.
The stability of alpha particle driven Alfven eigenmodes (AEs) is analysed in high fusion power DT discharges on JET. Both hot ion H mode and shear optimized discharges are considered. Unstable AEs are not observed in hot ion H mode DT discharges even at the highest fusion power with alpha particle beta beta {sub 0}(0) approximately=0.7%. Theoretical analysis shows that the AE stabilization is caused by the large plasma pressure, which prevents the existence of core localized AEs at peak fusion performance. Kinetic toroidal AEs (KTAEs), which persist at high plasma pressure, are found to be radially extended and subject to strong damping. The stability analysis based on the CASTOR-K code confirms that AEs cannot be driven unstable by alpha particles in high performance hot ion H mode discharges performed at JET. Alfven eigenmodes in shear optimized regimes are more unstable than those in the hot ion H mode mainly due to the elevated central safety factor q, which increases the efficiency of AE interaction with energetic ions. As a consequence, AEs are observed in shear optimized DT discharges when ion cyclotron heating as low as 1 MW is applied. (41 refs.)
Shimomura, Y., R. Aymar, et al. (1999). "ITER overview." Nuclear Fusion
17th IAEA Fusion Energy Conference 39(9Y): 1295-1308.
The article summarizes six years of technical work carried out by the ITER Joint Central Team and Home Teams under the terms of the ITER Engineering Design Activities (EDA) Agreement. The major products of this effort are as follows: a comprehensive and detailed engineering design with supporting assessments, cost estimates and a schedule based on industrial studies; a comprehensive safety and environmental assessment; and technology R&D to validate and qualify; the design, including proof of technologies and industrial manufacture and testing of full size or scalable models of key components. The ITER design is at an advanced stage of maturity and contains sufficient technical information to serve as a basis for a decision on construction. The operation of ITER will demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes.
Skinner, C. H., C. A. Gentile, et al. (1999). "Flaking of co-deposited hydrogenated carbon layers on the TFTR limiter." Nuclear Fusion 39(9): 1081-5.
Flaking of co-deposited layers on the inner limiter tiles was recently observed in TFTR. This phenomenon was unexpected and has occurred since the termination of plasma operations on 4 April 1997. Flaking affects approximately 15% of the observable tiles and appears on isotropic graphite but not on carbon fibre composite tiles. Photographic images of the flakes and precise measurements of the limiter geometry are reported. The mobilizability of tritium retained in co-deposited layers is an important factor in safety analyses of future DT reactors. A programme to analyse the flakes and tiles is underway. (29 refs.)
Stallard, B. W., C. M. Greenfield, et al. (1999). "Electron heat transport in improved confinement discharges in DIII-D." Physics of Plasmas
40th Annual Meeting of the Division of Plasma Physics of the American-Physical-Society 6(5): 1978-84.
In DIII-D [J.L. Luxon and L.G. Davis, Fusion Technol. 8, 441 (1985)] tokamak plasmas with an internal transport barrier (ITB), the comparison of gyrokinetic linear stability (GKS) predictions with experiments in both low and strong negative magnetic shear plasmas provide improved understanding for electron thermal transport within the plasma. Within a limited region just inside the ITB, the electron temperature gradient (ETG) modes appear to control the electron temperature gradient and, consequently, the electron thermal transport. The increase in the electron temperature gradient with more strongly negative magnetic shear is consistent with the increase in the ETG mode marginal gradient. Closer to the magnetic axis the T{sub e} profile flattens and the ETG modes are predicted to be stable. With additional core electron heating, FIR scattering measurements near the axis show the presence of high k fluctuations (12 cm{sup -1}), rotating in the electron diamagnetic drift direction. This turbulence could impact electron transport and possibly also ion transport. Thermal diffusivities for electrons, and to a lesser degree ions, increase. The ETG mode can exist at this wave number, but it is computed to be robustly stable near the axis. Consequently, in the plasmas we have examined, calculations of drift wave linear stability do not explain the observed transport near the axis in plasmas with or without additional electron heating, and there are probably other processes controling transport in this region. (30 refs.)
Strachan, J. D., M. G. Bell, et al. (1999). "Rollover: confinement degradation during TFTR discharges." Nuclear Fusion 39(7): 919-31.
The time evolution is described of TFTR neutral beam heated plasmas that often featured a decreasing energy content called `rollover'. Rollover affects the gross plasma parameters such as energy content and fusion power produced, as well as the central plasma parameters and the plasma transport coefficients. Rollover was studied using a database of all the TFTR plasmas produced over a two year period. The plasma behaviour at the time of peak energy content was related to the behaviour at a later quasi-equilibrium time. The main conclusion is that limiter power handling issues were the underlying cause of TFTR rollover. This study re-emphasizes the fact that power handling is an important issue for plasma facing components in tokamaks. (26 refs.)
Stratton, B. C., R. V. Budny, et al. (1999). "Energetic particle transport and alpha driven instabilities in advanced confinement DT plasmas on TFTR." Nuclear Fusion
17th IAEA Fusion Energy Conference 39(9Y): 1309-1319.
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 q(min). In 10-20% of DT discharges, normal low n MHD activity causes alpha loss at levels above the first orbit loss rate.
Stutman, D., M. Finkenthal, et al. (1999). "Ultrasoft X-ray imaging system for the National Spherical Torus Experiment." Review of Scientific Instruments
12th Topical Conference on High-Temperature Plasma Diagnostics 70(1): 572-2.
A spectrally resolved ultrasoft X-ray imaging system, consisting of arrays of high resolution (<2 AA) and throughput (>or=tens of kHz) miniature monochromators, and based on multilayer mirrors and absolute photodiodes, is being designed for the National Spherical Torus Experiment. Initially, three poloidal arrays of diodes filtered for C 1s-np emission will be implemented for fast tomographic imaging of the colder start-up plasmas. Later on, mirrors tuned to the C Ly{sub alpha } emission will be added in order to enable the arrays to "see" the periphery through the hot core and to study magnetohydrodynamic activity and impurity transport in this region. We also discuss possible core diagnostics, based on tomographic imaging of the Ly{sub alpha } emission from the plume of recombined, low Z impurity ions left by neutral beams or fueling pellets. The arrays can also be used for radiated power measurements and to map the distribution of high Z impurities injected for transport studies. The performance of the proposed system is illustrated with results from test channels on the CDX-U spherical torus at Princeton Plasma Physics Laboratory. (13 refs.)
Takeno, H., Y. Yasaka, et al. (1999). "Poloidal phasing effects on particle transport during ion Bernstein heating." Plasma Physics and Controlled Fusion 41(5): 613-23.
Ion Bernstein wave (IBW) heating experiments are performed with a poloidally phased antenna. Between two kinds of poloidal phasing. No difference is found with the heating effect which is expected from the mechanism of heating presented previously. The significant differences due to the difference of poloidal phasing are found in observations of electron density variation, emission intensities of impurity lines and scrape-off layer plasma. All the results show that the poloidal phasing brings about variations of particle transport, and a ponderomotive effect of the antenna near-field might be the most probable mechanism. The phasing effect should be considered on an intense radiofrequency (RF) injection by poloidally aligned antenna elements. (20 refs.)
Taylor, E. D., C. Cates, et al. (1999). "Nonstationary signal analysis of magnetic islands in plasmas." Review of Scientific Instruments 70(12): 4545-51.
Rotating magnetic islands produce fluctuations on a variety of diagnostics in magnetic fusion energy plasmas. The analysis of these fluctuations requires the calculation of the amplitude, phase, and frequency of the oscillations. These three spectral quantities generally evolve in time, necessitating nonstationary signal analysis techniques. The Hilbert transform offers an efficient and accurate method of calculating these three quantities from one diagnostic signal. This feature allows the Hilbert transform to determine the success of the active rotation control of magnetic islands, and to calculate the profile of the diagnostic measurements in a frame of reference co-rotating with the magnetic island. Comparisons to quadrature and spectrogram techniques demonstrate the accuracy of the Hilbert transform method. (25 refs.)
Throumoulopoulos, G. N. and D. Pfirsch (1999). "A potential mechanism for the creation of reversed-magnetic-shear transport barriers in tokamaks." Physics of Plasmas 6(8): 3226-32.
The impact of reversed magnetic shear (s<0) along with a radial electric field E{sub r} on negative-energy perturbations (NEP's) is investigated for the case of cylindrical tokamak-like equilibria. NEP's can imply instabilities leading to anomalous transport [G.N. Throumoulopoulos and D. Pfirsch, Phys. Rev. E 56, 5979 (1997) and references therein]. For equilibria with E{sub r} corresponding to |e{sub nu } phi |<< beta T{sub nu } and eta {sub nu }<4/3 ( phi is the electrostatic potential, beta identical to 8 pi P/B{sup 2}, eta {sub nu } identical to delta lnT{sub nu }/ delta lnN{sub nu } and nu denotes the particle species) s<0 results in a reduction of the fraction of particles associated with NEP's (active particles) as compared with equilibria with s>0, by making the pressure profile steeper. The reduction is stronger the closer the minimum of the safety factor (q{sub min}) to the plasma center and the lower the negative value of s. For equilibria with |e{sub nu } phi |<< beta T{sub nu } and eta {sub nu }>4/3 the fraction of active particles decreases as the magnetic tension B{sub theta }{sup 2}/r becomes larger. For E{sub r}<0 corresponding to |e{sub nu } phi | approximately= beta T{sub nu } the electric field leads to a reduction of the fraction of active ions for eta {sub i}<4/3 and of the fraction of active electrons for eta {sub e}>4/3 in addition to that caused by s<0 and B{sub theta }{sup 2}/r, respectively. For T{sub nu }>or=|e{sub nu } phi |>> beta T{sub nu }, which corresponds to experimental observations, the reduction of active particles caused by E{sub r}<0 is insensitive to the sign and the value of s, the position of q{sub min}, and the value of B{sub theta }{sup 2}/r. In addition, it is shown that the experimentally evidenced facts that (a) s<0 is associated with a steepness of the pressure profile and (b) the position of q{sub min} is displaced towards the plasma edge as the toroidal current is increased are consistent with equilibrium considerations. It is emphasized that these results cannot yet give a complete picture, since the present paper still neglects toroidal effects. (30 refs.)
Wakatani, M., V. S. Mukhovatov, et al. (1999). "Chapter 2: Plasma confinement and transport." Nuclear Fusion 39(12): 2175-2249.
Physics knowledge in plasma confinement and transport relevant to design of a reactor-scale tokamak is reviewed and methodologies for projecting confinement properties to ITER are provided. Theoretical approaches to describing a turbulent plasma transport in a tokamak are outlined and phenomenology of major energy confinement regimes observed in tokamaks, including those with edge and internal transport barriers, is described. The chapter is focused on the energy confinement in the high confinement regime (H-mode) with the edge localized MHD modes, the basic operational regime of ITER. Three approaches are being pursued: (i) derivation of empirical global scaling laws; (ii) non-dimensionally similar studies; and (iii) one dimensional transport modelling codes, with the first approach recommended as the most robust at the present time. Special attention is paid to analysis of uncertainties in confinement predictions. Empirical scaling relations for projecting the L-mode to H-mode power threshold based on re,regression analysis of an extensive database are discussed. Particle and toroidal momentum confinement and their relation to energy confinement are reviewed.
Wolle, B. (1999). "Tokamak plasma diagnostics based on measured neutron signals." Physics Reports 312(1/2): 1-86.
Neutron diagnostics are of increasing importance for future fusion devices. Consequently, efforts are being made to improve the accuracy of underlying experimental and computational methods. The present article reviews the modelling and the analysis of measured neutron signals relevant for plasma diagnostics on tokamaks. The underlying numerical simulation of neutron signals involves various aspects. Firstly, a realistic characterization of the plasma as a neutron source is needed. Secondly, detailed knowledge about changes in energy spectra and total number of the initially emitted neutrons due to scattering and absorption in the volume between the neutron source and the detector system is required. Finally, the detection properties of the measuring systems have to be taken into account. Presently, a sophisticated numerical procedure which directly relates detector signals to physics properties of the emitted neutrons from the plasma is not available and progress is found to be incremental rather than revolutionary. This is mainly attributable to problems with modelling the plasma neutron source based on measured plasma data and modelling difficulties for the neutron transport. However, more recent results of plasma parameters derived from neutron measurements provide evidence for the improvements in the measurement, simulation and analysis procedures over the past two decades. (445 refs.)
Wong, K. L. (1999). "A review of Alfven eigenmode observations in toroidal plasmas." Plasma Physics and Controlled Fusion 41(1): R1-R56.
In toroidal magnetically confined plasmas, eigenmodes of Alfven waves can be destablized by energetic ions with velocities comparable to the Alfven velocity. With the advent of tokamak experiments in which populations of energetic ions can be introduced by neutral beam injection, radio frequency wave heating or by fusion reactions, major advances have been made in Alfven eigenmode research in the past 10 years. After introducing the basic concepts on the Alfven eigenmode instability, data on this subject from various toroidal devices are described, emphasizing the interplay between experiment and theory. Experimental results on mode identification, instability drive. mode damping and saturation, and energetic ion redistribution are compared with theory.
Wurden, G. A., A. J. Wurden, et al. (1999). "Plasma tails: Comets Hale-Bopp and Hyakutake." IEEE Transactions on Plasma Science 27(1): 142-3.
Comet Hale-Bopp was one of the largest comets ever recorded, and it exhibited both a massive dust tail and a plasma tail, which developed as it approached the Sun over the course of six months in 1996-1997. Because the dust responds to gravity and light pressure, but plasmas also respond to the local solar wind (Coulomb collisions and magnetic fields), there is typically an angular separation between the two tails. (5 refs.)
Wurden, G. A. and B. J. Peterson (1999). "Development of imaging bolometers for long-pulse MFE experiments (invited)." Review of Scientific Instruments
12th Topical Conference on High-Temperature Plasma Diagnostics 70(1): 255-2.
We have developed the concept of an imaging bolometer, capable of operation with 100's of individual channels, while relying on optical (infrared) readout of the temperature rise in a thin foil. A thin gold foil (0.5-5 mu m thick) is sandwiched between pieces of copper. The copper mask (a large thermal mass) has a hole pattern drilled into it to form many "individual pixels," effectively forming many separate sensors. This segmented foil/mask combination is exposed on its front side to plasma radiation through a cooled pinhole camera geometry. Simultaneously, a high-resolution infrared camera monitors any temperature change on the backside of the thin foil. A sensitive infrared (IR) camera views the foil through an IR telescope/periscope system, and is shielded from the magnetic and nuclear radiation fields, either by distance and/or material shielding. A simple time-dependent design algorithm, using 1D heat transport to a cold boundary, has been written in MathCad, which allows us to select optimal material and geometries to match the expected plasma conditions. We have built a compact prototype with 149 channels, and tested it successfully both in a vacuum test stand in the laboratory, and on a plasma in the CHS at the National Institute for Fusion Science, subjecting it to electron cyclotron heated and neutral beam injection heated conditions. A water-cooled version has been built for the new LHD. Since the IR imaging bolometer uses only metal parts near the plasma, and has no need for wiring or wiring feedthrus, it is intrinsically radiation hard, and has direct application to ignition device to test engineering concepts (ITER), or ITER-class experiments. (8 refs.)
Young, K. M., A. E. Costley, et al. (1999). "Chapter 7: Measurement of plasma parameters." Nuclear Fusion 39(12): 2541-75.
The physics issues of the measurements of the plasma properties necessary to provide both the control and science data for achieving the goals of the ITER device are discussed. The assessment of the requirements for these measurements is first discussed, together with priorities that relate to the experimental program. Subsequently, some of the proposed measurement techniques, the plasma diagnostics, are described with particular emphasis on their implementation on ITER and their capability to meet the requirements. A judgement on the present status of the diagnostic program on ITER is provided with some indication of the research and development program necessary to demonstrate viability of techniques or their implementation. (138 refs.)
Akhmetov, T. D., V. I. Davydenko, et al. (2000). "Measurements of the electron distribution function in the AMBAL-M startup plasma by an electrostatic analyzer." Plasma Physics Reports 26(2): 152-156.
The electron distribution function over longitudinal energies in the startup plasma of the end cell of the AMBAL-M device is measured with a small-size movable electrostatic analyzer, IE is found that, in the region where a substantial longitudinal current hows, the electron distribution function over longitudinal energies has a plateau in the 150-350-eV energy range. (C) 2000 MAIK "Nauka/Interperiodica".
Boedo, J. A., J. Ongena, et al. (2000). "Turbulent transport and turbulence in radiative I mode plasmas in TEXTOR-94." Nuclear Fusion 40(2): 209-21.
First measurements of turbulence levels and turbulence induced transport in the outer edge of the plasma of TEXTOR-94 during radiative improved mode discharges show a reduction by a factor of 4-7 of the radial particle turbulent transport. The quenching is most evident on the normalized potential fluctuations and is strongest above 100 kHz. Non-linear gyrokinetic particle-in-cell simulations of these discharges show impurity induced suppression of the electrostatic fluctuations associated with the ion temperature gradient driven mode over most of the cross-section, including the edge. Such a mechanism is proposed as the explanation for the improved confinement and turbulence reduction. The reduction in the edge turbulent transport levels is consistent with increased particle confinement time and the reduction of the SOL thickness. Particle and energy fluxes to the limiter are reduced by an order of magnitude. A concomitant increase of the measured energy and particle confinement times tau {sub E} and tau {sub p} versus radiated fraction suggests a common underlying suppression mechanism. (50 refs.)
Boucher, D., J. W. Connor, et al. (2000). "The international multi-tokamak profile database." Nuclear Fusion 40(12): 1955-81.
An international multi-tokamak profile database has been assembled, constituting a representative set of reference tokamak discharges for the purpose of testing local transport models against well documented data. In particular, it will allow one to measure the accuracy with which the models can reproduce experiments and draw confidence intervals for the predictions of the models outside the range covered in the database. This database is now available to the fusion community and may be accessed by anonymous ftp to iterphys.naka.go.jp; the purpose of this article is to describe the structure of the database and the discharges contributing to it so that all can take full advantage of this resource. Thus, after an introductory general discussion of the database, there is a more detailed description of its structure, with listings of variables emphasized and how to access the database. There is then a brief description of each contributing tokamak and information on the type of discharges available from that tokamak. This is followed by a more quantitative description of the data, giving the ranges of dimensional and dimensionless variables available. Some typical modelling results to illustrate the use of the database are given in the conclusion. (91 refs.)
Cates, C., M. Shilov, et al. (2000). "Suppression of resistive wall instabilities with distributed, independently controlled, active feedback coils." Physics of Plasmas 7(8): 3133-6.
External kink instabilities are suppressed in a tokamak experiment by either (1) energizing a distributed array of independently controlled active feedback coils mounted outside a segmented resistive wall or (2) inserting a second segmented wall having much higher electrical conductivity. When the active feedback coils are off and the highly conducting wall is withdrawn, kink instabilities excited by plasma current gradients grow at a rate comparable to the magnetic diffusion rate of the resistive wall. (39 refs.)
Clark, D. S. and N. J. Fisch (2000). "The possibility of high amplitude driven contained modes during ion Bernstein wave experiments in the tokamak fusion test reactor." Physics of Plasmas 7(7): 2923-32.
Extremely high quasilinear diffusion rates for energetic beam ions can be deduced from mode conversion experiments in the Tokamak Fusion Test Reactor (TFTR) [K.M. McGuire, H. Adler, P. Alling et al., Phys. Plasmas 2(6), 2176 (1995)]. A comparison of the experimental loss rates with the theoretical prediction for the interaction of energetic ions with mode converted ion Bernstein waves showed the theory to underpredict the diffusion coefficient by a factor of 30-70. An anomalously high diffusion coefficient might enhance the advantageous channeling of energetic alpha particle energy in a tokamak reactor. Resolving this discrepancy is thus of importance from the standpoint of practical interest in an improved tokamak reactor as well as from the standpoint of academic interest in basic wave-particle theory. A mechanism is proposed for this accelerated diffusion involving the excitation of a contained mode, possibly similar to that used in explaining the ICE (ion cyclotron emission) phenomenon, near the edge of a tokamak. (22 refs.)
Davis, H. A., E. O. Ballard, et al. (2000). "The Atlas power-flow system-a status report." IEEE Transactions on Plasma Science 28(5): 1405-13.
Atlas is a high-energy, pulsed-power facility under development to study materials properties and hydrodynamics under extreme conditions. Atlas will implode heavy liner loads (m ~ 45 gm) with a peak current of 27-32 MA delivered in 4 mu s, and it is energized by 96, 240-kV Marx generators storing a total of 23 MJ. The power-flow system transports current from the Marx generators to the load. It includes a load protection switch to protect the load in case of a prefire; 24 oil-insulated, vertically oriented, radially converging, triplate transmission lines; a transition region that couples the transmission lines to the power-flow channel; and a radially converging, dielectric-insulated, horizontal and conical, power-flow channel. Proposed experiments, the design of the power-flow system, test results, and status are presented. (12 refs.)
Ernst, D. R., R. E. Bell, et al. (2000). "Transitionless enhanced confinement and the role of radial electric field shear." Physics of Plasmas
Meeting of the American-Institute-of-Physics Division of Plasma Physics 7(2): 615-25.
Evidence is presented for the role of radial electric field shear in enhanced confinement regimes attained without sharp bifurcations or transitions. Temperature scans at constant density, created in the reheat phase following deuterium pellet injection into supershot plasmas in the Tokamak Fusion Test Reactor [J.D. Strachan, et al., Phys. Rev. Lett. 58, 1004 (1987)] are simulated using a physics-based transport model. The slow reheat of the ion temperature profile, during which the temperature nearly doubles, is not explained by relatively comprehensive models of transport due to Ion Temperature Gradient Driven Turbulence (ITGDT), which depends primarily on the (unchanging) electron density gradient. An extended model, including the suppression of toroidal ITGDT by self-consistent radial electric field shear, does reproduce the reheat phase. The extended reheat at constant density is observed in supershot but not L-Mode plasmas. (44 refs.)
Fasoli, A., D. Borba, et al. (2000). "Fast particles-wave interaction in the Alfven frequency range on the Joint European Torus tokamak." Physics of Plasmas
41st Annual Meeting of the Division of Plasma Physics of the American-Physical-Society 7(5): 1816-24.
Wave-particle interaction phenomena in the Alfven Eigenmode (AE) frequency range are investigated at the Joint European Torus [P.H. Rebut and B.E. Keen, Fusion Technol. 11, 13 (1987)] using active and passive diagnostic methods. Fast particles are generated by neutral beam injection, ion cyclotron resonance heating, and fusion reactions. External antennas are used to excite stable AEs and measure fast particle drive and damping separately. Comparisons with numerical calculations lead to an identification of the different damping mechanisms. The use of the active AE diagnostic system to generate control signals based on the proximity to marginal stability limits for AE and low-frequency magnetohydrodynamic (MHD) modes is explored. Signatures of the different nonlinear regimes of fast particle driven AE instabilities predicted by theory are found in the measured spectra. The diagnostic use of AE measurements to get information both on the plasma bulk and the fast particle distribution is assessed. (38 refs.)
Fitzpatrick, R. and F. L. Waelbroeck (2000). "Nonlinear dynamics of feedback modulated magnetic islands in toroidal plasmas." Physics of Plasmas 7(12): 4983-95.
An analysis is presented of the dynamics of a helical magnetic island chain embedded in a toroidal plasma, in the presence of an externally imposed, rotating, magnetic perturbation of the same helicity. Calculations are carried out in the large aspect-ratio, zero- beta , resistive magnetohydrodynamical limit, and incorporate a realistic treatment of plasma viscosity. There are three regimes of operation, depending on the modulation frequency (i.e., the difference in rotation frequency between the island chain and the external perturbation). For slowly modulated islands, the perturbed velocity profile extends across the whole plasma. For strongly modulated islands, the perturbed velocity profile is localized around the island chain, but remains much wider than the chain. Finally, for very strongly modulated islands, the perturbed velocity profile collapses to a boundary layer on the island separatrix, plus a residual profile which extends a few island widths beyond the separatrix. Analytic expressions are obtained for the perturbed velocity profile, the island equation of motion, and the island width evolution equation in each of these three regimes. The ion polarization correction to the island width evolution equation, which has previously been reported to be stabilizing, is found to be destabilizing in all three regimes. (39 refs.)
Fitzpatrick, R. and E. P. Yu (2000). "Nonlinear dynamo mode dynamics in reversed field pinches." Physics of Plasmas
41st Annual Meeting of the American-Physical-Society-Division-of-Plasma-Physics 7(9): 3610-24.
The nonlinear dynamics of a typical dynamo mode in a reversed field pinch, under the action of the braking torque due to eddy currents excited in a resistive vacuum vessel and the locking torque due to a resonant error-field, is investigated. A simple set of phase evolution equations for the mode is derived: these equations represent an important extension of the well-known equations of Zohm et al. [Europhys. Lett. 11, 745 (1990)] which incorporate a self-consistent calculation of the radial extent of the region of the plasma which corotates with the mode; the width of this region being determined by plasma viscosity. Using these newly developed equations, a comprehensive theory of the influence of a resistive vacuum vessel on error-field locking and unlocking thresholds is developed. Under certain circumstances, a resistive vacuum vessel is found to strongly catalyze locked mode formation. Hopefully, the results obtained in this paper will allow experimentalists to achieve a full understanding of why the so-called "slinky mode" locks in some reversed field pinch devices, but not in others. The locking of the slinky mode is currently an issue of outstanding importance in reversed field pinch research. (47 refs.)
Garzotti, L., B. Pegourie, et al. (2000). "Investigation of electron-distribution function and dynamo mechanisms in a reversed-field pinch by analysis of hydrogen-pellet deflection." Physical Review Letters 84(24): 5532-5.
In reversed-field pinches, two different mechanisms have been proposed to explain the dynamo process which drives the poloidal current needed to sustain the magnetic configuration: the kinetic dynamo theory and the magnetohydrodynamic (MHD) dynamo theory. Experimentally, they can be distinguished by the radial behavior of the electron distribution function. In this letter the trajectory deflection of frozen hydrogen pellets has been used as a diagnostic of suprathermal electrons in the plasma. The classical Spitzer-Harm distortion of the electron distribution function consistent with the MHD dynamo electric field is found to give a better modeling of the pellet trajectory. (25 refs.)
Heidbrink, W. W., E. M. Carolipio, et al. (2000). "The weak effect of static, externally imposed, helical fields on fusion product confinement in the DIII-D tokamak." Nuclear Fusion 40(5): 935-40.
Stationary helical fields with toroidal mode numbers n=1 and n=3 are applied to beam heated DIII-D plasmas. Measurements of the 14 MeV neutron emission monitor the confinement of the 1 MeV tritium fusion product. To within ~15% uncertainty, static magnetic fields with vacuum amplitudes of delta B/B ~ O(10{sup -3}) have no impact on fusion product confinement. (30 refs.)
Idomura, Y., M. Wakatani, et al. (2000). "Gyrokinetic theory of slab electron temperature gradient mode in negative shear tokamaks." Physics of Plasmas 7(6): 2456-68.
With a gyrokinetic integral eigenvalue code, it is shown that both the slab ion temperature gradient (ITG) mode and the slab electron temperature gradient (ETG) mode have three types of branches in the negative shear configuration: a single mode-rational surface mode, a double mode-rational surface mode, and a nonresonant mode. For typical fusion plasma parameters satisfying lambda {sub De}{sup 2}>> rho {sub te}{sup 2}, a Weber-type differential eigenmode equation of the ETG mode becomes essentially different from that of the ITG mode, because of the Debye shielding effect, where lambda {sub De} is the Debye length and rho {sub te} is the electron Larmor radius. A scale length of the ETG modes is characterized by lambda {sub De}, and different types of analytic solutions are obtained for the ETG modes. From a comparison of the transport coefficient based on the mixing length theory, it is shown that in the negative shear configuration, the slab ETG mode gives an order of magnitude larger transport coefficient compared with an estimate for the conventional normal-sheared slab ETG mode. (26 refs.)
Idomura, Y., M. Wakatani, et al. (2000). "Stability of E x B zonal flow in electron temperature gradient driven turbulence." Physics of Plasmas 7(9): 3551-3566.
The electron temperature gradient driven turbulence in a slab configuration modeling the negative shear tokamak is studied using a gyrokinetic finite element particle-in-cell code. It is found that quasisteady E(r)xB zonal flows are generated in finite magnetic shear regions in both sides of the q(min)-surface, where the electron thermal transport is reduced substantially compared with the q(min)-surface region. Stability analyses of the electrostatic Kelvin-Helmholtz (KH) mode show that the quasisteady E(r)xB zonal flow pattern is closely related to the q profile or the magnetic shear, which has a stabilizing effect on the KH mode. By changing the q profile to reduce the magnetic shear, the KH mode becomes unstable for the quasisteady E(r)xB zonal flow, and the E(r)xB zonal flows disappear in the weak magnetic shear region. Numerical results show a possibility of controlling E(r)xB zonal flows with the magnetic shear, which depends on the stability of the KH mode. (C) 2000 American Institute of Physics. [S1070-664X(00)02709-9].
Jain, K. K. and A. S. Sree (2000). "Formation of a reversed magnetic shear configuration with electrodes and fluctuation behaviour in a toroidal plasma." Plasma Physics and Controlled Fusion 42(3): 229-36.
A different method for the generation of a hollow toroidal current profile, and thus a reversed magnetic shear configuration, is demonstrated in a toroidal plasma. A set of an identical pair of rings were used to drive the hollow toroidal plasma current. The measurements of a poloidal magnetic field show the formation of a reversed magnetic shear in the core region. Low levels of density fluctuation are observed away from the large reversed magnetic shear region, and fluctuation minima is found in the vicinity of a potential well. (11 refs.)
Karasik, M., A. L. Roquemore, et al. (2000). "Experiment and modeling of an atmospheric pressure arc in an applied oscillating magnetic field." Physics of Plasmas 7(6): 2715-27.
A set of experiments are carried out to measure and understand the response of a free-burning atmospheric pressure carbon arc to applied transverse dc and ac magnetic fields. The arc is found to deflect parabolically for the dc field and assumes a growing sinusoidal structure for the ac field. A simple analytic two-parameter fluid model of the arc dynamics is derived, in which the arc response is governed by the arc jet originating at the cathode, with the applied J*B force balanced by inertia. Time variation of the applied field allows evaluation of the parameters individually. A fit of the model to the experimental data gives a value for the average jet speed an order of magnitude below Maecker's estimate of the maximum jet speed [H. Maecker, Z. Phys. 141, 198 (1955)]. An example industrial application of the model is considered. (34 refs.)
Lanier, N. E., D. Craig, et al. (2000). "Control of density fluctuations and electron transport in the reversed-field pinch." Physical Review Letters 85(10): 2120-3.
A recent study conducted on the Madison Symmetric Torus reversed-field pinch has shown that control of density fluctuations can be achieved through modification of the current density profile. Most of the power in the density fluctuations is directly associated with core-resonant resistive tearing modes. We report that, during auxiliary current drive experiments, these density fluctuations are reduced about an order of magnitude over the entire plasma cross section and the resulting electron confinement is increased eightfold. (24 refs.)
Nadle, D. L., C. Cates, et al. (2000). "The feedback phase instability in the HBT-EP tokamak." Nuclear Fusion 40(10): 1791-4.
Observations of a performance limiting feedback phase instability in the HBT-EP tokamak are reported. The phase instability consists of a rapid growth of the phase difference between an m/n=2/1 tearing mode and an external resonant magnetic perturbation. Observations of mode angular dynamics during phase instability test discharges show good agreement with theoretical estimates of the phase instability timescale. The phase instability limits feedback performance in HBT-EP by decreasing the feedback loop's phase accuracy as gain increases. (9 refs.)
Ohno, N. and T. Kato (2000). "Magnetic fluctuations near the relaxed state of RFP." Journal of the Physical Society of Japan 69(3): 716-21.
We investigate fluctuations near the relaxed state of RFP using MHD equations in terms of the cylindrical coordinates. We consider the fluctuations as perturbations near the relaxed state of RFP which is specified by Taylor's Bessel function model. The wave length of the fluctuations are regarded as small comparing to the scale length of the magnetic field of unperturbed relaxed state. The fluctuations of magnetic flux density and mass density are expressed by the shear Alfven mode, and fast and slow magnetosonic mode, respectively. The comparison of the calculated frequencies and the reported experimental results shows us some discrepancy. (15 refs.)
Ongena, J. and A. M. Messiaen (2000). "Heating, confinement and extrapolation to reactors." Fusion Technology
4th Carolus Magnus Summer School on Plasma Physics 37(2T): 455-66.
With regard to ITER and tokamaks we discuss confinement and extrapolation to the reactor, plasma thermalization, heating and confinement results in present machines. (25 refs.)
Peterson, B. J. (2000). "Infrared imaging video bolometer." Review of Scientific Instruments 71(10): 3696-701.
A new concept for an infrared imaging bolometer is proposed which provides full video (two-dimensional) imaging of the radiated power from the plasma. This concept preserves all the advantages (compared to conventional metal foil resistive bolometers) of the previously proposed and tested segmented mask infrared imaging bolometer (SIB). It avoids the problems associated with the copper mask of the SIB, while giving a full frame video image of the plasma radiation with improved experimental flexibility regarding the pixel size. Analysis of the noise equivalent power shows that compared to a SIB with the same pixel area, the infrared imaging video bolometer is 2-5 times more sensitive with improved spatial resolution. These benefits are gained at the expense of the mechanical support, which the mask provides for the foil in the SIB. A numerical algorithm is used to solve the two-dimensional heat diffusion equation for the foil and determine the time-dependent spatial distribution of incident power on the foil from the infrared (IR) video camera measurements of the foil temperature. Testing the algorithm using a Gaussian model of the incident power shows that it can accurately reproduce the Gaussian model to within 6%. A simple scheme to absolutely calibrate the entire foil is described and several design points are detailed pointing out the range and experimental flexibility of the diagnostic using currently available IR camera technology. (16 refs.)
Post, R. F. (2000). "Resource Letter IMCF-1: Inertially and magnetically confined fusion." American Journal of Physics 68(2): 105-14.
This Resource Letter is intended to introduce the reader to the history of fusion research, and to provide references covering the research since its first beginnings and up to the present for those who wish to dig deeper. (144 refs.)
Ray, N. R. and K. Bhattacharya (2000). "Analysis of spiky net toroidal current in the magnetized toroidal plasma from the point of view of helicity conservation." Plasma Physics and Controlled Fusion 42(12): 1321-9.
It has been observed that `runaway oscillations' in the toroidal current of the magnetized resistive toroidal plasma may cause continuous tearing activity, resulting in `spiky' net toroidal current and modulation of toroidal and poloidal magnetic fields of the plasma with a definite phase relationship, depending upon the magnitude of the vertical magnetic field. The present experimental results on current decay and recovery are explained from the point of view of helicity conservation. (18 refs.)
Redi, M. H., A. Diallo, et al. (2000). "Robustness and flexibility in compact quasiaxial stellarators: Global ideal magnetohydrodynamic stability and energetic particle transport." Physics of Plasmas 7(6): 2508-16.
Concerns about the flexibility and robustness of a compact quasiaxial stellarator design are addressed by studying the effects of varied pressure and rotational transform profiles on expected performance. For thirty, related, fully three-dimensional configurations the global, ideal magnetohydrodynamic (MHD) stability and energetic particle transport are evaluated. It is found that tokamak intuition is relevant to understanding the magnetohydrodynamic stability, with pressure gradient driving terms and shear stabilization controlling both the periodicity preserving, N=0, and the nonperiodicity preserving, N=1, unstable kink modes. Global kink modes are generated by steeply peaked pressure profiles near the half radius and edge localized kink modes are found for plasmas with steep pressure profiles at the edge as well as with edge rotational transform above 0.5. Energetic particle transport is not strongly dependent on these changes of pressure and current (or rotational transform) profiles, although a weak inverse dependence on pressure peaking through the corresponding Shafranov shift is found. While good transport and MHD stability are not anticorrelated in these equilibria, stability only results from a delicate balance of the pressure and shear stabilization forces. A range of interesting MHD behaviors is found for this large set of equilibria, exhibiting similar particle transport properties. (26 refs.)
Rogister, A. L. (2000). "Phenomenology and theory of transport in fusion plasmas." Fusion Technology
4th Carolus Magnus Summer School on Plasma Physics 37(2T): 271-86.
The phenomenology of transport in magnetically confined plasmas is briefly described and the basic physical concepts underlying the theories of both anomalous and neoclassical transport are reviewed. Anomalous transport is a consequence of supra-thermal electric and magnetic fluctuations driven unstable by various mechanisms. The excited modes saturate by inducing a relaxation of the profiles towards the marginally stable state and via nonlinear coupling of the various modes. Specific theoretical models are described, together with their successes and drawbacks in the light of observed characteristics of plasma confinement. An estimate of the nuclear heating power required to balance the anomalous losses in the International Tokamak Experimental Reactor (ITER) is obtained on the basis of the electrostatic drift wave instability model. Large-scale gyrokinetic turbulence simulations and various "theoretical" transport models are discussed. Recent improvements of neoclassical theory, required in the vicinity of transport barriers, are described. (96 refs.)
Rosa, P. R. D. and G. Giruzzi (2000). "RF current drive by electron cyclotron waves in the presence of magnetic islands." Plasma Physics and Controlled Fusion 42(7): 755-769.
The influence of the presence of magnetic islands, and the consequent modification of the tokamak magnetic surface topology, on electron cyclotron current drive is analysed. To this end, a new three-dimensional Fokker-Planck code has been developed, taking into account the modifications of the magnetic equilibrium topology owing to the presence of the islands. Significant differences between the electron cyclotron current drive efficiency with and without islands inside the plasma are found, particularly in the case of interaction with locked modes.
Sen, A. K. (2000). "Control and diagnostic uses of feedback." Physics of Plasmas
41st Annual Meeting of the Division of Plasma Physics of the American-Physical-Society 7(5): 1759-66.
Recent results on multimode feedback control of magnetohydrodynamic (MHD) modes and a variety of diagnostic uses of feedback are summarized. First, is the report on reduction and scaling of transport under feedback. By controlling the fluctuation amplitudes and consequently the transport via feedback, it is found that the scaling of the diffusion coefficient is linear with root-mean-square rms fluctuation level. The scaling appears not to agree with any generic theory. A variety of other diagnostic uses of feedback have been developed. The primary goal is an experimental methodology for the determination of dynamic models of plasma turbulence, both for better transport understanding and more credible feedback controller designs. A specific motivation is to search for a low-order dynamic model, suitable for the convenient study of both transport and feedback. First, the time series analysis method is used for the determination of chaotic attractor dimension of plasma fluctuations. For E*B rotational flute modes it is found to be close to three, indicating that a low-order dynamic model may be adequate for transport prediction and feedback controller design. Second, a new method for direct experimental determination of nonlinear dynamical models of plasma turbulence using feedback has been developed. Specifically, the process begins with a standard three-wave coupling model and introduces a variable feedback gain. The power spectrum, delayed power spectrum, and bispectrum of fluctuations are then experimentally obtained. By varying the feedback gain continuously, an arbitrary number of numerical equations for a fixed number of unknowns can be generated. Their numerical solution yields the linear dispersion, as well as nonlinear coupling coefficients. This method has been successfully applied for E*B rotationally driven flute modes. (39 refs.)
Sharapov, S. E., B. Alper, et al. (2000). "Energetic particle physics in JET." Nuclear Fusion
6th IAEA Technical Committee Meeting on Energetic Particles in Magnetic Confinement Systems 40(7): 1363-81.
Results achieved on JET during the 1997-1999 experimental campaigns in the physics of energetic ions and runaway electrons are reviewed. Heating of deuterium-tritium (DT) plasmas by fusion born alpha particles is found to be similar to that achieved by comparable ICRF heating of deuterium plasmas. The stability of alpha particle driven Alfven eigenmodes (AEs) in the highest fusion power ELM-free H mode discharges is shown to be consistent with the existing theoretical analysis for AEs. Direct measurements of the trapped alpha particle and knock-on deuteron distribution functions by a neutral particle analyser (NPA) are described. New ICRF heating scenarios tested in JET DT plasmas are presented in view of the possible use of the ICRF heating of ITER-like DT plasmas on route to ignition. The energetic ion pinch in the presence of toroidally asymmetric ICRF waves is studied experimentally on JET. Recent experimentally observed effects of ICRF accelerated ions on sawteeth in JET are reviewed. Detailed time and space resolved X ray images of the electron runaway beam in flight spontaneously generated by disruptions on JET are described. (57 refs.)
Sudo, S., H. Nakanishi, et al. (2000). "Overview of LHD diagnostics and data acquisition system." Fusion Engineering and Design
2nd IAEA Technical Committee Meeting on Control Data Acquisition and Remote Participation on Fusion Research 48(1/2): 179-85.
The overview on the recent status and future plan of plasma diagnostics and data acquisition for LHD is briefly described. According to the successful starting of LHD experiments on March 31, 1998, the diagnostics have also launched in the adequate pace with providing the necessary basic data for plasma performance analysis. For this, the LHD data acquisition system, handling large amounts of data, is playing an important role in the experiment and data analysis. (7 refs.)
Watts, C., R. F. Gandy, et al. (2000). "Quatos: A university-scale test of the quasi-toroidal stellarator concept." Fusion Technology 37(3): 211-24.
A university-scale concept exploration experiment is proposed to investigate the improved confinement properties of the quasi-toroidal stellarator. The experiment would investigate three issues germane to a larger proof-of-principle experiment: first, the improved neoclassical confinement through measurement of the ion temperature dependence on axisymmetry; second, the impact of internal currents on the stability of this configuration; and third, the effect of the reduced viscosity on plasma rotation and the formation of transport barriers. (54 refs.)
Zweben, S. J., R. V. Budny, et al. (2000). "Alpha particle physics experiments in the Tokamak Fusion Test Reactor." Nuclear Fusion 40(1): 91-149.
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. (191 refs.)
Antar, G. Y., S. I. Krasheninnikov, et al. (2001). "Experimental evidence of intermittent convection in the edge of magnetic confinement devices." Physical Review Letters 87(6): 065001-4.
Probe measurements in the PISCES linear device indicate the presence of plasma radially far from where it is produced. We show that this is mainly caused by large-scale structures of plasma with high radial velocity. Data from the Tore Supra tokamak show striking similarities in the shape of these intermittent events as well as the fluctuation density probability distribution and frequency spectrum. The fact that intermittent, large-scale events are so similar in linear devices and tokamaks indicates the universality of convective transport in magnetically confined plasmas. (21 refs.)
Bialek, J., A. H. Boozer, et al. (2001). "Modeling of active control of external magnetohydrodynamic instabilities." Physics of Plasmas
42nd Annual Meeting of the Division of Plasma Physics Of the American-Physical-Society/10th International Congress on Plasma Physics 8(5): 2170-80.
A general circuit formulation of resistive wall mode (RWM) feedback stabilization developed by Boozer [Phys. Plasmas 5, 3350 (1998)] has been used as the basis for the VALEN computer code that calculates the performance of an active control system in arbitrary geometry. The code uses a finite element representation of a thin shell structure in an integral formulation to model arbitrary conducting walls. This is combined with a circuit representation of stable and unstable plasma modes. Benchmark comparisons of VALEN results with large aspect ratio analytic model of the current driven kink mode are in very good agreement. VALEN also models arbitrary sensors, control coils, and the feedback logic connecting these sensors and control coils to provide a complete simulation capability for feedback control of plasma instabilities. VALEN modeling is in good agreement with experimental results on DIII-D [Garofalo et al., Nucl. Fusion 40, 1491 (2000)] and HBT-EP [Cates et al., Phys. Plasmas 7, 3133 (2000)]. VALEN feedback simulations have also been used to evaluate and optimize the sensor/coil configurations for present and planned RWM experiments on DIII-D. These studies have shown a clear advantage for the use of local poloidal field sensors driving a "mode control" feedback logic control loop and configurations which minimize the control coil coupling to the stabilizing resistive wall. (32 refs.)
Boedo, J. A., D. Rudakov, et al. (2001). "Transport by intermittent convection in the boundary of the DIII-D tokamak." Physics of Plasmas 8(11): 4826-33.
Intermittent plasma objects (IPOs) featuring higher pressure than the surrounding plasma, and responsible for ~50% of the E*B{sub T} radial transport, are observed in the scrape off layer (SOL) and edge of the DIII-D tokamak [J. Watkins et al., Rev. Sci. Instrum. 63, 4728 (1992)]. Conditional averaging reveals that the IPOs, produced at a rate of ~3*10{sup 3} s{sup -1}, are positively charged and also polarized, featuring poloidal electric fields of up to 4000 V/m. The IPOs move poloidally at speeds of up to 5000 m/s and radially with E*B{sub T}/B{sup 2} velocities of ~2600 m/s near the last closed flux surface (LCFS), and ~330 m/s near the wall. The IPOs slow down as they shrink in radial size from 4 cm at the LCFS to 0.5 cm near the wall. The IPOs appear in the SOL of both L and H mode discharges and are responsible for nearly 50% of the SOL radial E*B transport at all radii; however, they are highly reduced in absolute amplitude in H-mode conditions. (40 refs.)
Buchner, P., H. Schubert, et al. (2001). "Diagnostics of an RF plasma flash evaporation process using the monochromatic imaging technique." Plasma Chemistry and Plasma Processing 21(1): 1-21.
The high densities and high gas temperature of RF plasmas at pressures near 1 atm are favorable for the development of plasma sources capable of evaporating solid precursors in the plasma zone. In the cooler region downstream of the plasma, the evaporated material condenses to nanoparticles and/or coatings. The complete evaporation of precursors injected into a thermal plasma depends on plasma and precursor parameters and is studied in this paper. Since many parameters contribute to the evaporation, fast experimental techniques are necessary to carry out a systematic study of the evaporation process. The monochromatic imaging technique applied in this work uses an intensified CCD camera with optical filters for the detection of characteristic plasma emission lines. The high spatial and temporal resolution of this technique results in a detailed picture of plasma emission and particle evaporation for different process parameters. These results are compared to model calculations for particle evaporation. (15 refs.)
Campbell, D. J. (2001). "The physics of the International Thermonuclear Experimental Reactor FEAT." Physics of Plasmas
42nd Annual Meeting of the Division of Plasma Physics Of the American-Physical-Society/10th International Congress on Plasma Physics 8(5): 2041-9.
The International Thermonuclear Experimental Reactor FEAT design for a long-pulse tokamak burning plasma experiment (R=6.2 m, a=2 m, B=5.3 T, I=15 MA) is intended to achieve extended burn in inductively driven deuterium-tritium plasmas with the ratio of fusion power to auxiliary heating power, Q, of at least 10 and a nominal fusion power output of ~500 MW. It also aims to demonstrate steady-state plasma operation using noninductive current drive with a Q of at least 5. Particular features of the design are: a significant operating window for Q=10 inductive operation; long inductive pulses (several hundred seconds burn); a capability for studying steady-state scenarios, specifically in cases where alpha -particles make a significant contribution to the plasma pressure; disruption physics processes which are comparable to those expected at the reactor scale; and an alpha -particle density and heating power which permit the key issues of alpha -particle confinement and alpha -particle driven magnetohydrodynamic instabilities to be investigated under conditions appropriate to a reactor. (40 refs.)
Chitarin, G., A. Masiello, et al. (2001). "The electromagnetic system for the active control of the non-axisymmetric helical deformation in RFX." Plasma Physics and Controlled Fusion 43(4): 543-58.
Experiments on resistive tearing mode control in a reversed field pinch by applying a rotating magnetic field are presented. After a brief survey of tearing mode characteristics, the actions taken to prevent their detrimental effects are described. These actions consist in the generation of a rotating magnetic field able to modify the dynamic of the m=0, n=1 tearing mode. The rotating field is generated by superimposing an alternating current component to the current circulating in the toroidal field winding. To achieve this, the toroidal field circuit was appropriately modified. Under some conditions, the rotating field exerts a dragging torque on the helical deformation of the plasma column, produced by phase-locked m=1 modes, so that the deformation can be either localized in a selected toroidal position or put into a continuous toroidal rotation. (33 refs.)
Clark, D. A., B. G. Anderson, et al. (2001). Liner velocity, current, and symmetry measurements on the 32 megamp flux compression generator experiment alt-1. 28th IEEE International Conference on Plasma Science/13th IEEE International Pulsed Power Conference, LAS VEGAS, NEVADA, I E E E.
A flux compression generator pulse power system, designed, built, and fielded by a Russian team at the All Russian Scientific Research Institute of Experimental Physics (VNIIEF), was used to successfully drive an aluminum liner to velocities greater than 12 km/sec. The experiment objective was to demonstrate performance of a precision liner implosion at an Atlas current of 30 MA or greater. Diagnostics to measure liner performance were an essential part of the experiment. An experimental team from Los Alamos National Laboratory (LANL) provided a suite of diagnostics to measure liner performance. Three diagnostics were fielded: 1. A velocity interferometer (VISAR) to continuously measure the liner inner-surface velocity throughout the entire range of travel, 2. Two Faraday rotation devices to measure liner current during the implosion, and, 3. Sixteen fiber optic impact pins to record liner impact time and provide axial and azimuthal symmetry information. All diagnostics performed very well. Major results are maximum current: 323 MA, velocity at impact: greater than 12 km/sec, symmetry: the impact pins indicated that the liner was smooth, solid, and axially symmetric upon arrival at the diagnostic package. The LANL team fabricated, installed, and recorded the three diagnostics presented here. All necessary equipment was brought to the site in Russia. The VNIIEF team fielded other diagnostics to measure machine performance. Results of machine diagnostics are reported in other presentations.
Davis, H. A., R. K. Keinigs, et al. (2001). "The Atlas high-energy density physics project." Japanese Journal of Applied Physics, Part 1 (Regular Papers, Short Notes & Review Papers) 40(2B): 930-4.
Atlas is a pulsed-power facility under development at Los Alamos National Laboratory to drive high-energy density experiments. Atlas will be operational in the summer of 2000 and is optimized for the study of dynamic material properties, hydrodynamics, and dense plasmas under extreme conditions. Atlas is designed to implode heavy-liner loads in a z-pinch configuration. The peak current of 30 MA is delivered in 4 mu s. A typical Atlas liner is a 47-gram-aluminum cylinder with ~4-cm radius and 3-cm length. Three to five MJ of kinetic energy will be delivered to the load. Using composite layers and a variety of interior target designs, a wide variety of experiments in ~cm{sup 3} volumes will be performed. Atlas applications, machine design, and the status of the project are reviewed. (8 refs.)
Degnan, J. H., J. M. Taccetti, et al. (2001). "Implosion of solid liner for compression of field reversed configuration." IEEE Transactions on Plasma Science 29(1): 93-8.
The design and first successful demonstration of an imploding solid liner with height to diameter ratio, radial convergence, and uniformity suitable for compressing a field reversed configuration is discussed. Radiographs indicated a very symmetric implosion with no instability growth, with ~13x radial compression of the inner liner surface prior to impacting a central measurement unit. The implosion kinetic energy was 1.5 megajoules, 34% of the capacitor stored energy of 4.4 megajoules. (25 refs.)
Federici, G., C. H. Skinner, et al. (2001). "Plasma-material interactions in current tokamaks and their implications for next step fusion reactors." Nuclear Fusion 41(12R): 1967-2137.
The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and thus is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project. (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust, and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.
Fitzpatrick, R., E. Rossi, et al. (2001). "Improved evolution equations for magnetic island chains in toroidal pinch plasmas subject to externally applied resonant magnetic perturbations." Physics of Plasmas 8(10): 4489-500.
An improved set of island evolution equations is derived that incorporates the latest advances in MHD (magnetohydrodynamical) theory. These equations describe the resistive/viscous-MHD dynamics of a nonlinear magnetic island chain, embedded in a toroidal pinch plasma, in the presence of a programmable, externally applied, resonant magnetic perturbation. A number of interesting example calculations are performed using the new equations. In particular, an investigation is made of a recently discovered class of multiharmonic resonant magnetic perturbations that have the novel property that they can lock resonant island chains in a stabilizing phase. (37 refs.)
Franz, P., L. Marrelli, et al. (2001). "Soft X ray tomographic imaging in the RFX reversed field pinch." Nuclear Fusion 41(6): 695-709.
SXR cross-sectional distributions measured in the European reversed field pinch (RFP) experiment RFX by means of a tomographic diagnostic are described. Various reconstruction techniques have been compared. A correction technique to minimize the systematic effect due to the non-identical thicknesses of the material filters has been developed. A number of experimental situations have been explored, and a study of the correlation between the MHD and SXR properties of the plasma has been made. The analysis concentrates both on standard discharges and on the recently discovered quasi-single-helicity states, whose signature is the growth of a sizeable m=1 structure in the emissivity distributions. Detailed imaging of the plasma in these conditions is presented. (57 refs.)
Idomura, Y., S. Tokuda, et al. (2001). "Gyrokinetic theory of drift waves in negative shear tokamaks." Nuclear Fusion 41(4): 437-45.
Linear and non-linear properties of slab drift waves in the negative sheared slab configuration modelling of the q{sub min} surface region in negative shear tokamaks are studied, where q{sub min} is the minimum value of the safety factor q. Linear calculations show that both the slab ion temperature gradient (ITG) driven mode and the slab electron temperature gradient (ETG) driven mode become strongly unstable around the q{sub min} surface. Non-linear simulations are performed for ETG turbulence, which evolves on a much faster timescale than ITG turbulence. It is found that quasi-steady E{sub r}*B zonal flows are generated by an inverse wave energy cascade process. Linear stability analyses of the electrostatic Kelvin-Helmholtz (KH) mode show that the quasi-steady E{sub r}*B zonal flow profile is closely related to the q profile or to the magnetic shear, which has a stabilizing effect on the KH mode. It is shown that the microscopic quasi-steady E{sub r}*B zonal flows arising from ETG turbulence have a strong stabilizing effect on the slab ITG mode. (11 refs.)
LeBlanc, B. P., R. E. Bell, et al. (2001). "High-Harmonic Fast-Wave Heating in NSTX." Aip Conference Proceedings
14th Topical Conference on Radio Frequency Power in Plasmas 595: 51-58.
High-Harmonic Fast-Wave (HHFW), a radio-frequency technique scenario applicable to high-beta plasmas, has been selected as one of the main auxiliary heating systems on NSTX. The HHFW antenna assembly comprises 12 toroidally adjacent current elements, extending poloidally and centered on the equatorial plane. This paper reviews experimental results obtained with a symmetrical (vacuum) launching spectrum with k(H) = 14 m(-1) at a frequency of 30 MHz. We describe results obtained when HHFW power is applied to helium and deuterium plasmas, during the plasma-current flattop period of the discharge. Application of 1.8-MW HHFW pulse to MHD quiescent plasmas resulted in strong electron heating, during which the central electron temperature, T-eo more than doubled from approximate to0.5 keV to 1.15 keV. In deuterium plasmas, HHFW heating was found less efficient, with a Teo increase of the order of 40% during a 1.8-MW HHFW pulse, from approximate to 400 eV to approximate to 550 eV. (At HHFW power of 2.4 MW, T-eo increased by 60%, reaching 0.625 keV.) HHFW heating in presence of MHD activity is also discussed. A short neutral beam pulse was applied to permit charge-exchange recombination spectroscopy (CHERS) measurement of the impurity ion temperature T-i. Preliminary CHERS analysis show that T-i approximate to T-e during HHFW heating. Of special interest are deuterium discharges, where the application of HHFW power was done during the current ramp-up. We observe the creation of large density gradients in the edge region. In the latter case, the density rose spontaneously to n(eo) less than or equal to 8 x 10(13) cm(-3).
Lindemuth, I. R., W. L. Atchison, et al. (2001). The magnetically driven imploding liner parameter space of the Atlas capacitor bank. 28th IEEE International Conference on Plasma Science/13th IEEE International Pulsed Power Conference, LAS VEGAS, NEVADA, I E E E.
The Atlas capacitor bank (23 MJ, 30 MA, 240 kV) is now operational at Los Alamos. Atlas was designed to magnetically drive imploding liners for use as impactors in shock and hydrodynamic experiments. We have conducted a computational "mapping" of the high-performance imploding liner parameter space accessible with Atlas. The effect of charge voltage, transmission inductance, liner thickness, liner initial radius, and liner length, as well as other parameters, has been investigated. Our study shows that Atlas is ideally suited to be a liner driver for liner-on-plasma experiments in a Magnetized Target Fusion (MTF) context.
Maki, S., S. Masamune, et al. (2001). "Formation mechanisms of the inversed current in the edge region of RFP plasma." Journal of the Physical Society of Japan 70(2): 415-20.
The edge toroidal reversed current outside the field reversal surface manifestly plays some important roles in reversal field pinches (RFPs). The crucial roles of the edge reversed toroidal current has included the implication of the deep field reversal in the recently attained improved confinement modes in some RFPs. We have concentrated our attention on the formation mechanism of the reversed edge toroidal current in the STP-3M reversed field pinch. Measurements of fast electron current and floating potential have revealed minutely that the reversed edge toroidal current is driven by fluctuational transport of superthermal electrons produced in the core region. Time evolution of the floating potential, measurement of the space potential profile, has shown the radially outerward electric field, an indication of the preferential loss of electrons from the core region. The dominance of the core resonant modes in the toroidal mode power spectrum of the edge magnetic fluctuations indicates the stochastic magnetic field line in the core region. This is evidence that the origin of the edge fast electrons is the stochastic diffusion of fast electrons produced in the core region. (10 refs.)
Mansfield, D. K., D. W. Johnson, et al. (2001). "Observations concerning the injection of a lithium aerosol into the edge of TFTR discharges." Nuclear Fusion 41(12): 1823-1834.
A new method of actively modifying the plasma-wall interaction was tested on the Tokamak Fusion Test Reactor. A laser was used to introduce a directed lithium aerosol into the discharge scrape-off layer. The lithium introduced in this fashion ablated and migrated preferentially to the limiter contact points. This allowed the plasma wall interaction to be influenced in situ and in real time by external means. Significant improvement in energy confinement and fusion neutron production rate as well as a reduction in the plasma Z(eff) have been documented in a neutral beam heated plasma. The introduction of a metallic aerosol into the plasma edge increased the internal inductance of the plasma column and also resulted in prompt heating of core electrons in ohmic plasmas. Preliminary evidence also suggests that the introduction of an aerosol leads to both edge poloidal velocity shear and edge electric field shear.
Mantsinen, M. J., M. L. Mayoral, et al. (2001). "ICRF heating scenarios in JET with emphasis on He-4 plasmas for the non-activated phase of ITER." Aip Conference Proceedings
14th Topical Conference on Radio Frequency Power in Plasmas 595: 59-66.
In the initial phase of ITER operation, He-4 plasmas could be used in order to avoid activating the machine. The main ICRH scenarios foreseen for ITER He plasmas are (He-3)He-4 and (H)He-4. ICRH experiments have been carried out on JET using He-4 plasmas to validate these scenarios. At the same time, conditions for access to H-mode in plasmas of various isotope compositions from dominantly He-4 to dominantly D have been studied. Experiments have also been carried out for the first time in He plasmas with the ICRF power added to He-4 neutral beam injection at the third harmonic of He-4 in order to produce a He-4 tail for alpha particle studies.
Mantsinen, M. J., O. N. Jarvis, et al. (2001). "First observation of pT fusion in JET tritium plasmas with ICRF heating of protons." Nuclear Fusion 41(12): 1815-1822.
High power ICRF heating of a hydrogen minority ion species in JET tritium plasmas has generated a total neutron rate that is about 40% larger than the 14 MeV neutron rate originating from fusion reactions between bulk tritium ions and deuterium minority ions. The T(p,n)(3) He fusion reaction, caused by ICRF accelerated protons, is identified as a source for producing the excess neutron emission. This reaction is endothermic and has a proton energy threshold of about I MeV and a peak cross-section at about 3.0 MeV. The presence of protons with such high energies is detected in gamma ray and high energy neutral particle analyser measurements and is also confirmed by ICRF modelling with the PION code. The fast proton energy content and the pT fusion reactivity as simulated by the PION code are compared with the experimental measurements when classical slowing down and confinement of ICRF accelerated protons are assumed in the simulations.
Maqueda, R. J., G. A. Wurden, et al. (2001). "Digital-image capture system for the IR camera used in Alcator C-Mod." Review of Scientific Instruments
13th Topical Conference on High-Temperature Plasma diagnostics 72(1): 927-2.
An infrared imaging system, based on an Amber Radiance 1 infrared camera, is used at Alcator C-Mod to measure the surface temperatures in the lower divertor region. Due to the supra-linear dependence of the thermal radiation with temperature it is important to make use of the 12-bit digitization of the focal plane array of the Amber camera and not be limited by the 8 bits inherent to the video signal. It is also necessary for the image capture device (i.e., fast computer) to be removed from the high magnetic field environment surrounding the experiment. Finally, the coupling between the digital camera output and the capture device should be nonconductive for isolation purposes (i.e., optical coupling). A digital video remote camera interface (RCI) coupled to a PCI bus fiber optic interface board is used to accomplish this task. Using this PCI-RCI system, the 60 Hz images from the Amber Radiance 1 camera, each composed of 256*256 pixels and 12 bits/pixel, are captured by a Windows NT computer. An electrical trigger signal is given directly to the RCI module to synchronize the image stream with the experiment. The RCI can be programmed from the host computer to work with a variety of digital cameras, including the Amber Radiance 1 camera. (5 refs.)
Maqueda, R. J., G. A. Wurden, et al. (2001). "Edge turbulence measurements in NSTX by gas puff imaging." Review of Scientific Instruments
13th Topical Conference on High-Temperature Plasma diagnostics 72(1): 931-2.
Turbulent filaments in visible light emission corresponding mainly to density fluctuations at the edge have been observed in large aspect ratio tokamaks: TFTR, ASDEX, Alcator C-Mod, and DIII-D. This article reports on similar turbulent structures observed in the National Spherical Torus Experiment (NSTX) using a fast-framing, intensified, digital visible camera. These filaments were previously detected mainly in high recycling regions, such as at limiters or antennas, where the line emission from neutral atoms was modulated by the fluctuations in local plasma density. However, by introducing controlled edge gas puffs, i.e., gas puff imaging, we have increased the brightness and contrast in the fluctuation images and allowed the turbulent structure to be measured independently of the recycling. A set discrete fiber-optically coupled sight-lines also measured the frequency spectra of these light fluctuations with a 200 kHz bandwidth. Initial results in NSTX show that the turbulent filaments are well aligned with the magnetic field which can be up to 45 degrees from the horizontal at the outer midplane of NSTX. The dominant wavelength perpendicular to the magnetic field is ~7-11 cm, corresponding to a k{sub perpendicular to } rho {sub s} of ~0.3 at an assumed T{sub e}=25 eV, and the frequency spectra has a typical broad shape characteristic of edge turbulence extending to about 100 kHz. By imaging a He gas puff along a magnetic field line the characteristic radial scalelength appears to be in the 3-5 cm range. (12 refs.)
Menard, J. E., B. P. LeBlanc, et al. (2001). "Ohmic flux consumption during initial operation of the NSTX spherical torus." Nuclear Fusion 41(9): 1197-206.
Spherical tokamaks, because of their slender central columns, have very limited volt second capability relative to that of standard aspect ratio tokamaks of similar plasma cross-section. Recent experiments on NSTX have begun to quantify and optimize the ohmic current drive efficiency in an MA class ST device. Sustainable ramp rates in excess of 5 MA/s during the current rise phase have been achieved on NSTX, while faster ramps generate significant MHD activity. Discharges with I{sub p} exceeding 1 MA have been achieved in NSTX with nominal parameters: aspect ratio A=1.3-1.4, elongation k=2-2.2, triangularity delta =0.4, internal inductance l{sub i}=0.6, Ejima coefficient C{sub E}=0.35. Flux consumption efficiency results, performance improvements associated with first boronization and comparisons with neoclassical resistivity are described. (35 refs.)
Moir, R. W., R. H. Bulmer, et al. (2001). "Thick liquid-walled, field-reversed configuration-magnetic fusion power plant." Fusion Technology
14th Topical Meeting on the Technology of Fusion Energy 39(2): 758-767.
A thick flowing layer of liquid (e.g., flibe-a molten salt, Sn{sub 80}Li{sub 20} or Li-liquid metals) protects the structural walls of the field-reversed configuration (FRC) so that they can last the life of the plant even with intense 14 MeV neutron bombardment from the D-T fusion reaction. The surface temperature of the liquid rises as it passes from the inlet nozzles to the exit nozzles due to absorption of line and bremsstrahlung radiation, and neutrons. The surface temperature can be reduced by enhancement of convection near the surface to transport hot surface liquid into the cooler interior. The resulting temperature for evaporation estimates called, T{sub eff}, is 660, 714 and 460 degrees C for flibe, SnLi and Li, where thermal conductivity was assumed enhanced by a factor often for flibe. The corresponding evaporative flux from the wall must result in an acceptable impurity level in the core plasma. The shielding of the core by the edge plasma is modeled with a 2D transport code for the resulting impurity ions; these ions are either swept out to the distant end tanks, or diffuse to the hot plasma core. The calculations show core impurity levels adequately low for Li and Sn{sub 80}Li{sub 20} but is about ten times too large for flibe. An auxiliary plasma between the edge plasma and the liquid wall can further attenuate evaporating flux of atoms and molecules by ionization. The current in this auxiliary plasma might serve as the antenna for the current drive method, which produces a rotating magnetic field. (12 refs.)
Moses, R. W., R. A. Gerwin, et al. (2001). "Transport implications of current drive by magnetic helicity injection." Physics of Plasmas 8(11): 4839-48.
It is shown that in fusion plasma configurations sustained by electrode helicity injection, the core electron temperature (in electron volts) can, at most, be 25% to 40% of the electrode voltage (in volts). This result is obtained by assessing magnetic helicity injection as a driver of macroscopic steady-state plasma currents in magnetic confinement devices. Coaxial helicity injection using electrodes (CHI) and oscillating-field current drive (OFCD) are compared to inductive current drive. Magnetic helicity, K, is uniquely defined as the time-dependent volume integral of A.B when the surface components of A are purely solenoidal. Using an Ohm's law including Hall terms, magnetic helicity transport modeling shows that no closed magnetic surfaces with time and volume averaged parallel currents can exist continuously within a plasma sustained only by CHI or OFCD. The 25% to 40% limitations are obtained by considering long and short electron mean-free-path models of parallel energy transport. (50 refs.)
Narihara, K., K. Y. Watanabe, et al. (2001). "Observation of the "self-healing" of an error field island in the Large Helical Device." Physical Review Letters 87(13): 135002-4.
It was observed that the vacuum magnetic island produced by an external error magnetic field in the Large Helical Device shrank in the presence of plasma. This was evidenced by the disappearance of flat regions in the electron temperature profile obtained by Thomson scattering. This island behavior depended on the magnetic configuration in which the plasmas were produced. (17 refs.)
Nelson-Melby, E., A. Mazurenko, et al. (2001). "Phase contrast imaging of ion Bernstein and fast waves in Alcator C-Mod." Aip Conference Proceedings
14th Topical Conference on Radio Frequency Power in Plasmas 595: 90-97.
Direct observation of ICRF (similar to 80 MHz) waves (k(perpendicular to) similar or equal to 0.5 cm(-1) - 10 cm(-1)) is now possible in Alcator C-Mod using an optical heterodyne technique on the Phase Contrast Imaging (PCI) system, which uses a CO2 laser to observe electron density fluctuations. The PCI observations are vertical chord averages, so the full-wave ICRF code TORIC[1] has been used to simulate the wave fields in these plasmas to aid in interpretation. Mode-converted ion Bernstein waves (IBW) in plasmas composed of H, D and He-3 at 6 T have been observed at both high (1 MA) and low (400 kA) current. The fast magnetosonic wave launched from the low-field side has been observed in high density (similar to 5 x 10(20) m(-3)) D(H) plasmas with of axis ICRH at 4.5 T. Comparison between PCI measurements and code results are presented. The measured wave numbers are in good agreement with the local dispersion relations for both types of waves.
Okabayashi, M., J. Bialek, et al. (2001). "Active feedback stabilization of the resistive wall mode on the DIII-D device." Physics of Plasmas
42nd Annual Meeting of the Division of Plasma Physics Of the American-Physical-Society/10th International Congress on Plasma Physics 8(5): 2071-82.
A proof of principle magnetic feedback stabilization experiment has been carried out to suppress the resistive wall mode (RWM), a branch of the ideal magnetohydrodynamic (MHD) kink mode under the influence of a stabilizing resistive wall, on the DIII-D tokamak device [Plasma Phys. Controlled Fusion Research (International Atomic Energy Agency, Vienna, 1986), p. 159; Phys. Plasmas 1, 1415 (1994)]. The RWM was successfully suppressed and the high beta duration above the no-wall limit was extended to more than 50 times the resistive wall flux diffusion time. It was observed that the mode structure was well preserved during the time of the feedback application. Several lumped parameter formulations were used to study the feedback process. The observed feedback characteristics are in good qualitative agreement with the analysis. These results provide encouragement to future efforts towards optimizing the RWM feedback methodology in parallel to what has been successfully developed for the n=0 vertical positional control. Newly developed MHD codes have been extremely useful in guiding the experiments and in providing possible paths for the next step. (31 refs.)
Omelchenko, Y. A., M. J. Schaffer, et al. (2001). "Nonlinear stability of field-reversed configurations with self-generated toroidal field." Physics of Plasmas 8(10): 4463-9.
The field-reversed configuration (FRC) is a high-beta compact toroidal plasma confinement scheme in which the external poloidal field is reversed on the geometric axis by azimuthal (toroidal) plasma current. A quasineutral, hybrid, particle-in-cell (PIC) approach [Y. A. Omelchenko and R. N. Sudan, Phys. Plasmas 2, 2773 (1995)] is applied to study long-term nonlinear stability of computational FRC equilibria to a number of toroidal modes, including the most disruptive tilt mode. In particular, a self-generated toroidal magnetic field is found to be an important factor in mitigating the instability and preventing the confinement disruption. This is shown to be a unique FRC property resulting from the Hall effect in the regions of vanishing poloidal magnetic field. The instability-driven toroidal field stabilizes kink formation by increasing the magnetic field energy without destabilizing curvature-driven plasma motion. Finally, the tilt instability saturates due to nonlinear, finite Larmor radius (FLR) effects and plasma relaxation to a quasisteady kinetic state. During this transition the FRC is shown to dissipate a substantial amount of initially trapped flux and plasma energy. These effects are demonstrated for kinetic and fluid-like, spherical and prolate FRCs. (25 refs.)
Petersen, P. I. and D.-D. Team (2001). "Recent findings in DIII-D relative to advanced tokamak modes and their implications for fusion energy." Fusion Technology
14th Topical Meeting on the Technology of Fusion Energy 39(2): 305-314.
An advanced tokamak is characterized by increased confinement, stability and steady state operation. The increased confinement and stability are obtained through modifications to the shape and profiles of the plasma and through stability feedback control. These modifications have to be self-consistent. The increased confinement makes it possible to make smaller and thereby lower cost reactors for the same power output as compared to conventional tokamaks. Four potential modes for advanced tokamaks are currently being studied on DIII-D: radiative improved mode, high internal inductance l{sub i} mode, negative central shear (NCS) mode, and quiescent double barrier (QDB) mode. High-density plasma are important for reactors and recent experiments in DIII-D have shown that it is possible to operate substantially above the Greenwald limit. Control of the internal transport barriers that are responsible for the increased confinement have been improved in counter injected neutral beam plasmas. One of the limiting instabilities for the performance of high bootstrap fraction negative central shear plasmas is the resistive wall mode. These modes have to a certain degree been suppressed in DIII-D by using the six-section correction coil. With a newly installed upper inner divertor in DIII-D it has been possible to obtain improved density and impurity control. An upgrade of the electron cyclotron system is being done on DIII-D. Three 1 MW gyrotrons are being added. This system has been used to completely suppress the neoclassical tearing mode by applying electron cyclotron current drive at definite positions and in very localized areas. Finally, the implication of the recent findings for fusion reactors will be discussed. (24 refs.)
Peterson, B. J., M. Osakabe, et al. (2001). "Infrared imaging video bolometer for the large helical device." Review of Scientific Instruments
13th Topical Conference on High-Temperature Plasma diagnostics 72(1): 923-2.
A new type of infrared imaging bolometer, known as the infrared imaging video bolometer (IRVB) has been under development at the National Institute of Fusion Science (NIFS) over the last year. A prototype of this diagnostic has been designed and constructed at NIFS and installed on the large helical device (LHD) and successfully operated. This diagnostic utilizes a 66*90*0.001 mm gold foil mounted in a frame to detect the radiation and neutrals from the plasma which are incident on the foil through a 1 cm diameter pinhole. The resulting temperature distribution on the foil is measured using an AGEMA THV 900 LW infrared camera having 136*272 pixels with a frame rate of 15 Hz and a nominal sensitivity of 80 mK. An image of the foil and surrounding frame consisting of 120*160 pixels is resampled down to 12*16 pixels using a linear interpolation scheme. Using a numerical technique the spatial and temporal derivatives of the temperature distribution on the foil are calculated and the incident power density on the foil is determined using a calibration of the foil obtained by means of a HeNe laser. The resulting 10*14 pixel view of the plasma radiation at a 15 Hz frame rate has a noise equivalent power density of 0.5 mW/cm{sup 2}. The IRVB was mounted on a LHD tangential port for the 1999 campaign. An image of the plasma radiation during a discharge using the inboard vacuum vessel wall as a limiter shows radiation localized near the limiting surface. (12 refs.)
Raman, R., T. R. Jarboe, et al. (2001). "Non-inductive current generation in NSTX using coaxial helicity injection." Nuclear Fusion 41(8): 1081-6.
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. (15 refs.)
Raman, R., T. R. Jarboe, et al. (2001). "Initial results from coaxial helicity injection experiments in NSTX." Plasma Physics and Controlled Fusion 43(3): 305-12.
Coaxial helicity injection has been investigated on the National Spherical Torus Experiment (NSTX). Initial experiments produced 130 kA of toroidal current without the use of the central solenoid. The corresponding injector current was 20 kA. Discharges with pulse lengths up to 130 ms have been produced. (18 refs.)
Robertson, S., J. Espejo, et al. (2001). "Neoclassical effects in the annular Penning trap." Physics of Plasmas
42nd Annual Meeting of the Division of Plasma Physics Of the American-Physical-Society/10th International Congress on Plasma Physics 8(5): 1863-9.
Neoclassical transport has been investigated with a modified Malmberg-Penning trap which has conductors along the axis to create an azimuthal magnetic field. The axial bounce motion of the electrons is accompanied by a radial drift which changes sign at the ends of the device causing drift orbits of finite radial extent. Analysis and numerical simulations show that the transport is neoclassical with mobility and diffusion coefficients depending upon the axial magnetic field alone rather than the absolute value of the magnetic field. Experiments with added helium gas to create electron-neutral collisions show that the electron mobility from an applied radial electric field and the Ware drift from an azimuthal electric field both have neoclassical values over a wide range of magnetic fields and collision frequencies. (17 refs.)
Sabbagh, S. A., S. M. Kaye, et al. (2001). "Equilibrium properties of spherical torus plasmas in NSTX." Nuclear Fusion 41(11): 1601-11.
Research in NSTX has been conducted to establish spherical torus plasmas to be used for high beta auxiliary heated experiments. This device has a major radius R{sub 0}=0.86 m and a midplane halfwidth of 0.7 m. It has been operated with toroidal magnetic field B{sub 0}<or=0.3 T and I{sub p}<or=1.0 MA. The evolution of the plasma equilibrium is analysed between discharges with an automated version of the EFIT code. Limiter, double null and lower single null diverted configurations have been sustained for several energy confinement times. The plasma stored energy reached 92 kJ ( beta {sub t}=17.8%) with neutral beam heating. A plasma elongation in the range 1.6<or= kappa <or=2.0 and a triangularity in the range 0.25<or= delta <or=0.45 have been sustained, with values of kappa =2.6 and delta =0.6 being reached transiently. The reconstructed magnetic signals are fitted to the corresponding measured values with low errors. Aspects of the plasma boundary, pressure and safety factor profiles are supported by measurements from non-magnetic diagnostics. Plasma densities have reached 0.8 and 1.2 times the Greenwald limit in deuterium and helium plasmas, respectively, with no clear limit encountered. Instabilities including sawteeth and reconnection events, characterized by Mirnov oscillations, and a perturbation of the I{sub p}, kappa and l{sub i} evolutions, have been observed. A low q limit was observed and is imposed by a low toroidal mode number kink instability. (17 refs.)
Schilham, A. M. R., G. M. D. Hogeweij, et al. (2001). "Electron thermal transport barriers in RTP: experiment and modelling." Plasma Physics and Controlled Fusion 43(12): 1699-721.
Experiments in which very localized electron cyclotron heating (ECH) is scanned through the RTP plasma show sharp transitions, in which the electron temperature profile abruptly changes shape. The phenomenology-the profiles shapes, the sharp transitions-can be reproduced with a transport model which features electron transport barriers near all (half-) integer q values. This model is tested against ECH deposition scans in seven series of typically 60 discharges. Between the series q{sub a} and <n{sub e}>- were varied 4 < q{sub a} < 10 and 1.5 < <n{sub e}> * 10{sup 19} m{sup -3} < 4.5. For a single scan, the model parameters are fixed. The model is capable of reproducing all data sets. A sensitivity analysis shows that the determination of the strength of the barriers at integer q values is very stable. These barriers show a trend chi {sub e}(barrier) ~ <n{sub e}>{sup -1}, while the dependence on q{sub a} is weak. The model reproduces the <n{sub e}> and q{sub a} components of L-mode scaling of global confinement. (35 refs.)
Smolyakov, A. I., E. Lazzaro, et al. (2001). "Stabilization of magnetic islands due to the sheared plasma flow and viscosity." Plasma Physics and Controlled Fusion 43(12): 1661-9.
It is shown that the asymmetric deformation of the magnetic island caused by the finite plasma viscosity gives a stabilizing effect on the magnetic islands in the Rutherford regime. Such stabilization typically overcomes the destabilizing effect of the ion polarization current. (41 refs.)
Snyder, P. B. and G. W. Hammett (2001). "Electromagnetic effects on plasma microturbulence and transport." Physics of Plasmas 8(3): 744-9.
Results are presented from three-dimensional kinetic-fluid simulations of pressure gradient driven microturbulence in toroidal long mean-free-path plasmas. A numerically efficient model which includes self-consistent magnetic fluctuations and nonadiabatic electron dynamics is employed. A transition from electrostatic ion-drift turbulence to Alfvenic turbulence is seen at modest values of the plasma pressure. Significant electromagnetic effects on heat conductivity are observed, including an increase as the ideal ballooning threshold is approached, particularly when electron Landau damping is included. Turbulent spectra show a number of similarities to experimental fluctuation measurements. (42 refs.)
Sudo, S., Y. Nagayama, et al. (2001). "Overview of large helical device diagnostics (invited)." Review of Scientific Instruments
13th Topical Conference on High-Temperature Plasma diagnostics 72(1): 483-2.
The Large Helical Device (LHD) is the largest helical machine with superconducting coils. Key diagnostics issues for LHD are: (a) capability for multidimensional measurements because of the nonaxisymmetric toroidal plasma; (b) measurements of the electric field; (c) cross check of fundamental parameters using different methods; (d) advanced measurements appropriate for steady-state operation; and (e) a satisfactory data acquisition system. The design and research and development of plasma diagnostics were carried out taking these issues into consideration. As a result, the present status of diagnostics is described: diagnostics for LHD operation, fundamental diagnostics for plasma performance, diagnostics for physics subjects, innovative diagnostics and diagnostics for long-pulse operation. The LHD experiment started in March, 1998. Since then, the development of diagnostics has kept pace with the experimental campaigns. (19 refs.)
Swain, D. W., J. R. Wilson, et al. (2001). "Results from HHFW system operation on NSTX." Aip Conference Proceedings
14th Topical Conference on Radio Frequency Power in Plasmas 595: 154-157.
The HHFW system on NSTX has operated with the full 12-antenna, 6-transmitter configuration, delivering over 2 MW reliably for pulse lengths over 100 ins with various phasings of the antenna system. A circuit model of the full 12-antenna coupled system has been developed that gives good agreement with vacuum measurements. When it is used to extract the effects of the plasma on the rf circuit, pronounced asymmetries in antenna loading are observed, even when antenna phasing is symmetrical (e.g., 0 pi0 pi0 pi.....). The loading of the plasma on the antenna has been calculated with the RANT3D code using measured edge density profiles in front of the antenna; these agree with measured loading values.
Sykes, A. (2001). "Overview of recent spherical tokamak results." Plasma Physics and Controlled Fusion
28th EPS Conference on Controlled Fusion and Plasma Physics 43: A127-A139.
Several new spherical tokamak (ST) experiments have become operational in the last two years. Results are now becoming available which increase the understanding of tokamak physics, extend confinement and threshold databases for next step devices such as ITER, and show the possibilities offered by the ST as a possible future burning plasma device or fusion power plant.
Wang, Z. H. and C. W. Barnes (2001). "Exact solutions to magnetized plasma flow." Physics of Plasmas 8(3): 957-963.
Exact analytic solutions for steady-state magnetized plasma flow (MPF) using ideal magnetohydrodynamics formalism are presented. Several cases are considered. When plasma flow is included, a finite plasma pressure gradient delp can be maintained in a force-free state JxB = 0 by the velocity gradient. Both incompressible and compressible MPF examples are discussed for a Taylor-state spheromak B field. A new magnetized nozzle solution is given for compressible plasma when U parallel toB. Transition from a magnetized nozzle to a magnetic nozzle is possible when the B field is strong enough. No physical nozzle would be needed in the magnetic nozzle case. Diverging-, drum- and nozzle-shaped MPF solutions when U perpendicular toB are also given. The electric field is needed to balance the UxB term in Ohm's law. The electric field can be generated in the laboratory with the proposed conducting electrodes. If such electric fields also exist in stars and galaxies, such as through a dynamo process, then these solutions can be candidates to explain single and double jets.
Wurden, G. A., T. P. Intrator, et al. (2001). "Diagnostics for a magnetized target fusion experiment." Review of Scientific Instruments
13th Topical Conference on High-Temperature Plasma diagnostics 72(1): 552-2.
We are planning experiments using a field reversed configuration plasma injected into a metal cylinder, which is subsequently electrically imploded to achieve a fusing plasma. Diagnosing this plasma is quite challenging due to the short timescales, high energy densities, high magnetic fields, and difficult access. We outline our diagnostic sets in both a phase I study (where the plasma will be formed and translated), and phase II study (where the plasma will be imploded). The precompression plasma (diameter of only 8-10 cm, length of 30-40 cm) is expected to have n~10{sup 17} cm{sup -3}, T~100-300 eV, B~5 T, and a lifetime of 10-20 mu s. We will use visible laser interferometry across the plasma, along with a series of fiber-optically coupled visible light monitors to determine the plasma density and position. Excluded flux loops will be placed outside the quartz tube of the formation region, but inside of the diameter of the theta -pinch formation coils. Impurity emission in the visible and extreme ultraviolet range will be monitored spectroscopically, and fast bolometers will measure the total radiated power. A 20 J Thomson scattering laser beam will be introduced in the axial direction, and scattered light (from multiple spatial points) will be collected from the sides. Neutron diagnostics (activation and time-resolved scintillation detectors) will be fielded during both phases of the DD experiments. (9 refs.)
Baldwin, M. J., R. P. Doerner, et al. (2002). "Deuterium retention in liquid lithium." Nuclear Fusion 42(11): 1318-23.
Measurements of deuterium retention in samples of lithium exposed in the liquid state to deuterium plasma are reported. Retention was measured as a function of plasma ion dose in the range 6 * 10{sup 19}-4 * 10{sup 22} D atoms and exposure temperature between 523 and 673 K using thermal desorption spectrometry. The results are consistent with the full uptake of all deuterium ions incident on the liquid metal surface and are found to be independent of the temperature of the liquid lithium over the range explored. Full uptake, consistent with very low recycling, continues until the sample is volumetrically converted to lithium deuteride. This occurs for exposure temperatures where the gas pressure during exposure was both below and slightly above the corresponding decomposition pressure for LiD in Li. (39 refs.)
Belchenko, Y. I., Y. Oka, et al. (2002). "Negative ion production in multicusp sources." Review of Scientific Instruments 73(4): 1746-51.
The study of negative ion production in the multicusp negative ion sources (MS) was done by the directed deposition of well-defined amount of cesium into MS, by the cesium recovery from the polluted layers in the MS and by the plasma grid masking. The data obtained evidences the surface-plasma mechanism of negative ion production in the MS. It is shown that a dynamic cesium-tungsten coverage on a plasma grid surface provides the enhanced H{sup -} production in the MS. The Cs+W coverage is produced on the hot plasma grid surface due to coadsorption of cesium and of tungsten, evaporated from filaments. The permanent flux of cesium to plasma grid coverage is produced by the thick Cs+W reservoir with a high (75%) percentage of cesium on the cold anode surface. A relatively high cesium seed with a rate of about 20 mg/1 h/30 shots operation is necessary to support the increased H{sup -} production in the regular MS. Electrode processing by an additional discharge recovers the cesium from the aged coverage and enhances the H{sup -} production without an additional cesium seed. Recovery processing permits the use of deposited cesium more efficiently and to minimize cesium addition during the MS long-term operation. An essential decrease of cesium escape to accelerator and an improvement of injector high voltage operation is produced by introducing the shutter mask at plasma grid surface. The shutter mask experiment displayed, that >60% of H{sup -} beam current was born on the hot shutter mask surface. The source operation with a high negative shutter mask biasing and an increased hydrogen pressure resulted in a 35% higher H{sup -} yield, than of regular MS. (15 refs.)
Bush, C. E., R. Maingi, et al. (2002). "Evolution and termination of H-modes in NSTX." Plasma Physics and Controlled Fusion
8th IAEA Technical Committee Meeting on H-Mode Physics and Transport Barriers 44: A323-32.
The dynamic evolution of the first National Spherical Torus Experiment (NSTX) H-modes will be discussed. The H-modes were obtained in lower-single null divertor discharges with various forms of plasma heating. The exact timing of divertor formation and also the NBI power level affects both whether or not the discharge exhibits ELMs and the duration of the ELM-free phase. ELM-free discharges had energy confinement as high as 120 ms, whereas the few discharges with ELMs had confinement times ~50-70 ms. Buildup of a steep edge density gradient and formation of 'ears' on the density profile were observed within a few ms of the L-H transition, yielding broader density and pressure profiles. The L-H transition was marked by a decrease in edge visible light and simultaneous increase in electron Bernstein wave emission, reflecting a steepening of the edge density gradient. Gas puff imaging of He-I light during the H-mode phase showed rapid formation of a narrow emission layer ~2 cm wide in the H-mode phase, which returned within 20 microseconds at termination of the H-mode phase to a broader turbulent emission layer. All of the H-modes were terminated by an MHD reconnection event. The first power threshold (P{sub th}) study showed the neutral beam injection power (P{sub b}) component of P{sub th} to be <or=0.84 MW, higher than that predicted by the ITER database scaling. (15 refs.)
Carraro, L., M. E. Puiatti, et al. (2002). "Impurity transport during pulsed poloidal current drive experiment in the reversed field pinch experiment RFX." Plasma Physics and Controlled Fusion 44(10): 2135-48.
The characteristics of the impurity transport in the RFX reversed field pinch experiment during enhanced confinement regimes, such as those obtained performing pulsed poloidal current drive and oscillating poloidal current drive, have been investigated. It is found that during the enhanced confinement period, the ratio between the Ly alpha line of C VI and the resonance line of C V increases systematically well beyond the expected enhancement due to the temperature improvement. Experimental data have been reproduced by means of a one-dimensional, time dependent, collisional-radiative impurity transport code, which simulates several C and O lines, continuum emission and radiated power. In standard RFX plasmas the typical diffusion coefficient is ~20 m{sup 2} s{sup -1} at the plasma centre decreasing to ~1 m{sup 2} s{sup -1} at the edge. Best simulations of the spectroscopic data during the enhancement confinement phases are obtained by expanding the low diffusion region at the edge from r/a = 0.9 to r/a = 0.8 +or- 0.05 and shrinking the neutral density profile at the edge. (42 refs.)
Causey, R. A. (2002). "Hydrogen isotope retention and recycling in fusion reactor plasma-facing components." Journal of Nuclear Materials 300(2/3): 91-117.
The proper design of a fusion reactor is not possible unless there is an understanding of the hydrogen isotope retention and recycling for the plasma-facing components. From the tritium inventory point of view, it is absolutely necessary to understand the short-term and long-term hydrogen isotopes retention characteristics of the individual materials used for the first wall and divertor. From the plasma density and fueling point of view, it is necessary to understand the recycling characteristics of these materials. This report is an overview of the available data on hydrogen isotope retention and recycling for beryllium, tungsten, carbon, and selected liquid metals. For each material discussed, recommendations are made as to the most appropriate values to use for parameters such,as diffusivity, solubility, recombination rate coefficient, and trapping. (160 refs.)
Counsell, G. F., A. Kirk, et al. (2002). "Boundary plasma and divertor phenomena in MAST." Plasma Physics and Controlled Fusion
IAEA Technical Committee Meeting on Divertor Concepts 44(6): 827-43.
The boundary and divertor plasma of a large spherical tokamak (ST), with significant auxiliary heating, is explored in detail for the first time. The extreme geometry of the ST is found to play a key role, giving rise to strong asymmetries in the transport of heat and particle fluxes (including those during ELMs) towards the outer divertor targets. Preliminary modelling reveals the very significant contribution of magnetic flux expansion to the particle flux balance of the ST scrape-off layer. Divertor detachment at the inboard targets is observed in L-mode at moderate core densities for plasmas with up to 1 MW of auxiliary heating. (11 refs.)
Donne, A. J. H. (2002). "Diagnostics for current density and radial electric field measurements: overview and recent trends." Plasma Physics and Controlled Fusion
29th EPS Conference on Plasma Physics and Controlled Fusion 44: B137-B158.
The measurements of the current density and the radial electric field profiles in tokamaks have considerably gained importance since advanced operational scenarios have come into play. A detailed knowledge of these profiles is desirable for operational scenarios involving reversed magnetic shear and/or the active control of internal and edge transport barriers. Unfortunately, the current density as well as the radial electric field are among the plasma parameters that are most difficult to diagnose. Routine methods to measure the current density profile are based on the Faraday effect (i.e. polarimetry) and the motional Stark effect (MSE). The most advanced diagnostics for the measurement of the radial electric field in the plasma core are the heavy ion beam probe and again the MSE. This paper will briefly explain the need for detailed measurements of the current density and radial electric field profile. Subsequently, the various diagnostics to measure these parameters will be reviewed. The emphasis will be especially put on recent trends, rather than on an exhaustive overview.
Ejiri, A. (2002). "Four-beam correlation method for local fluctuation measurements in plasmas." Review of Scientific Instruments 73(4): 1766-74.
In order to measure fine scale fluctuations of visible emission in plasmas, the four-beam correlation method is proposed and its feasibility is studied. This is an extension of the conventional cross beam techniques and uses four beams (i.e., four lines of sight), which lie in two parallel planes. While two beams in the same plane provide the fluctuation levels at the intersection, two beams in different planes provide correlation as a function of the distance between the planes. Various cases are considered, and the method feasibility is shown. This method is useful for reconstructing a two- or three-dimensional wave-number spectrum. In addition, it can be used to measure the magnetic field direction assuming that the parallel correlation length is longer than the perpendicular one. (14 refs.)
Evtikhin, V. A., I. E. Lyublinski, et al. (2002). "Lithium divertor concept and results of supporting experiments." Plasma Physics and Controlled Fusion
IAEA Technical Committee Meeting on Divertor Concepts 44(6): 955-77.
The ITER project development has shown that considerable difficulties are encountered when already known engineering solutions and materials are used for divertor and divertor plates for tokamaks of such a scale. We offer to use a Li capillary-pore system (CPS) as a plasma facing material for tokamak divertor. Evaporated Li serves as a gas target and redistributes thermal load. The heat flux from the plasma is transported to the first wall by Li radiation in the plasma periphery. This allows the divertor plate to reduce the heat flux. A solid CPS filled with liquid lithium has a high resistance to surface damage in the stationary mode and during plasma transitions (disruptions, ELMs, VDEs, runaways) to assure normal operation of the divertor target plates. These materials are not the sources of impurities giving rise to Z{sub eff} and they will not be collected as dust in the divertor area and in ducts. Experiments with lithium CPS in a steady-state mode (up to 25 MW m{sup -2}) and in plasma disruption simulation conditions (~5 MJ m{sup -2}, ~0.5 ms) have been performed. High stability of these systems have been shown. Li limiter tests on T-11M tokamak have revealed the lithium CPS compatibility with the edge plasma for energy loads of up to 10 MW m{sup -2}. In a stable discharge mode at lithium limiter temperature of 20-600 degrees C, no Li abnormal erosion and injection to plasma have been detected. A high sorption of D{sup +} and H{sup +} ions on the vessel walls was the main substantial result of the replacement of a graphite limiter by lithium one. He and D sorption was terminated by wall heating up to 50-100 degrees C and above 350 degrees C, respectively. T-11 tests on helium discharge allowed to reduce limiter heat load by a factor of two due to lithium radiation. All the experimental results have shown considerable progress in the development of lithium divertor. (45 refs.)
Goodman, G. E. (2002). "The "18 year" alternative: A business proposal to develop commercial fusion energy by 2021 - The Fusion Power Corporation." Journal of Fusion Energy 21(2): 117-120.
This policy essay asserts that the "35 year" plan recently adopted by the U.S. Department of Energy's Fusion Energy Sciences Advisory Committee is too risk averse and too costly. An alternative "18 year" schedule is proposed. All dollar amounts shown below are undiscounted, and are only intended to be indicative of approximate future costs.
Greenwald, M. (2002). "Density limits in toroidal plasmas." Plasma Physics and Controlled Fusion 44(8): R27-80.
In addition to the operational limits imposed by MHD stability on plasma current and pressure, an independent limit on plasma density is observed in confined toroidal plasmas. This review attempts to summarize recent work on the phenomenology and physics of the density limit. Perhaps the most surprising result is that all of the toroidal confinement devices considered operate in similar ranges of (suitably normalized) densities. The empirical scalings derived independently for tokamaks and reversed-field pinches are essentially identical, while stellarators appear to operate at somewhat higher densities with a different scaling. Dedicated density limit experiments have not been carried out for spheromaks and field-reversed configurations, however, 'optimized' discharges in these devices are also well characterized by the same empirical law. In tokamaks, where the most extensive studies have been conducted, there is strong evidence linking the limit to physics near the plasma boundary: thus, it is possible to extend the operational range for line-averaged density by operating with peaked density profiles. Additional particles in the plasma core apparently have no effect on density limit physics. While there is no widely accepted, first principles model for the density limit, research in this area has focussed on mechanisms which lead to strong edge cooling. Theoretical work has concentrated on the consequences of increased impurity radiation which may dominate power balance at high densities and low temperatures. These theories are not entirely satisfactory as they require assumptions about edge transport and make predictions for power and impurity scaling that may not be consistent with experimental results. A separate thread of research looks for the cause in collisionality enhanced turbulent transport. While there is experimental and theoretical support for this approach, understanding of the underlying mechanisms is only at a rudimentary stage and no predictive capability is yet available. (278 refs.)
Guo, H. Y., A. L. Hoffman, et al. (2002). "Formation and steady-state maintenance of field reversed configuration using rotating magnetic field current drive." Physics of Plasmas 9(1): 185-200.
Rotating magnetic fields (RMF) have been used to both form and maintain field reversed configurations (FRC) in quasisteady state. These experiments differ from steady-state rotamaks in that the FRCs are similar to those formed in theta-pinch devices, that is elongated and confined inside a flux conserver. The RMF creates an FRC by driving an azimuthal current which reverses an initial positive bias field. The FRC then expands radially, compressing the initial axial bias flux and raising the plasma density, until a balance is reached between the RMF drive force and the electron-ion friction. This generally results in a very high ratio of separatrix to flux conserver radius. The achievable final conditions are compared with simple analytic models to estimate the effective plasma resistivity. The RMF torque on the electrons is quickly transferred to the ions, but ion spin-up is limited in these low density experiments, presumably by ion-neutral friction, and does not influence the basic current drive process. However, the ion rotation can result in a rotating n=2 distortion if the separatrix radius is too far removed from the plasma tube wall. (22 refs.)
Heidbrink, W. W. (2002). "Alpha particle physics in a tokamak burning plasma experiment." Physics of Plasmas
43rd Annual Meeting of the Division of Plasma Physics of the American-Physical-Society 9(5): 2113-19.
Much is known about the behavior of energetic ions in tokamak devices but much remains to be understood. Single-particle effects are well understood and provide a firm basis for extrapolation to a burning plasma. In contrast, collective effects involving fast ions are more poorly understood and extrapolations are unreliable. Collective modes of concern include toroidicity-induced and ellipticity-induced Alfven eigenmodes, kinetic ballooning modes, and internal kink modes. In addition to these magnetohydrodynamic normal modes, there are also energetic particle modes characterized by strong dependence on the fast-ion distribution function. Although many issues are important areas of study in current experiments, five issues distinguish burning plasma experiments. First, the energetic alphas are not the dominant source of free energy for the instabilities unless the fusion power exceeds the heating power by a factor of 10. Second, the damping of the instabilities depends sensitively on mode coupling to other heavily-damped waves. The magnitude of this coupling is expected to depend on the normalized thermal gyroradius, which is much smaller in a reactor. Third, in a reactor, both the radial extent of the instabilities and the fast-ion orbit contract relative to current experiments, so the fast-ion transport will change. Fourth, when instability occurs, a larger number of modes are unstable, so the mechanism of nonlinear saturation could shift from fast-ion transport to mode coupling. Fifth, because of the extreme sensitivity of energetic particle modes to the distribution function, an isotropic alpha particle distribution function differs from anisotropic fast-ion populations. (47 refs.)
Intrator, T., M. Taccetti, et al. (2002). "Experimental measurements of a converging flux conserver suitable for compressing a field reversed configuration for magnetized target fusion." Nuclear Fusion 42(2): 211-22.
Data are presented that are part of a first step in establishing the scientific basis of magnetized target fusion (MTF) as a cost effective approach to fusion energy. A radially converging flux compressor shell with characteristics suitable for MTF is demonstrated to be feasible. The key scientific and engineering question for this experiment is whether the large radial force density required to uniformly pinch this cylindrical shell would do so without buckling or kinking its shape. The time evolution of the shell has been measured with several independent diagnostic methods. The uniformity, height to diameter ratio and radial convergence are all better than required to compress a high density field reversed configuration to fusion relevant temperature and density. (46 refs.)
Kaita, R., D. Johnson, et al. (2002). "NSTX diagnostics for fusion plasma science studies." IEEE Transactions on Plasma Science
Annual Meeting of the International Conference on Plasma Science/Pulsed Power Conference (PPPS 2001) 30(1): 219-226.
This paper will discuss how plasma science issues are addressed by the diagnostics for the National Spherical Torus Experiment (NSTX), the newest large-scale machine in the magnetic confinement fusion (MCF) program. The development of new schemes for plasma confinement involves the interplay of experimental results and theoretical interpretations. A fundamental requirement, for example, is a determination of the equilibria for these configurations. For MCF, this is well established in the solutions of the Grad-Shafranov equation. While it is simple to state its basis in the balance between the kinetic and magnetic pressures, what they Are as functions of space and time are often not easy to obtain. Quantities like the plasma pressure and current density are not directly measurable. They are derived from data that Are themselves complex products of more basic parameters. The same difficulties apply to the understanding of plasma instabilities. Not only are the needs for spatial and temporal resolution more stringent, but the wave parameters which characterize the instabilities are difficult to resolve. We will show how solutions to the problems of diagnostic design on NSTX, and the physics insight the data analysis provides, benefits both NSTX and the broader scientific community. (14 refs.)
Kline, J. L., E. E. Scime, et al. (2002). "Slow wave ion heating in the HELIX helicon source." Plasma Sources, Science and Technology 11(4): 413-25.
Ion temperature measurements have been made at multiple axial and radial locations in a helicon source for a range of magnetic field strengths and RF frequencies. The observed temperature gradient along the axis suggests limited thermal transport along the magnetic field. The radial profiles are flat near the axis and in some cases peak near the edge of the plasma. The ion temperature measurements combined with calculations of the perpendicular wave numbers for the slow wave or 'Trivelpiece-Gould' mode are consistent with ion heating due to ion Landau damping of the slow wave at the edge of the plasma. (40 refs.)
LaBombard, B. (2002). "An interpretation of fluctuation induced transport derived from electrostatic probe measurements." Physics of Plasmas 9(4): 1300-11.
Fluctuation-induced particle fluxes ( Gamma {sub n phi }) in the edge of Alcator C-Mod [Phys. Plasmas 1, 1511 (1994)] are inferred from a fast-scanning probe using standard analysis techniques. The magnitude and profile shape of Gamma {sub n phi } is clearly inconsistent with fluxes inferred from global particle and power balance. These differences are difficult to reconcile if Gamma {sub n phi } is interpreted as a measure of the particle flux in the unperturbed plasma. However, if Gamma {sub n phi } is reinterpreted as the particle flux which must "fill-in" the presheath zone formed by the probe, these inconsistencies are eliminated. In this case, an effective diffusivity in the presheath zone (D{sub ps}) can be estimated from Gamma {sub n phi }. D{sub ps} is found to be in the range of diffusivities inferred from global particle balance (D{sub SOL}), indirectly supporting the hypothesis. However, the profile of D{sub ps} and its dependency on discharge conditions are markedly different than D{sub SOL}, implying that D{sub ps} is also not simply related to transport in the unperturbed plasma. (59 refs.)
Lipschultz, B., B. LaBombard, et al. (2002). "Investigation of the origin of neutrals in the main chamber of Alcator C-Mod." Plasma Physics and Controlled Fusion
IAEA Technical Committee Meeting on Divertor Concepts 44(6): 733-48.
A series of experiments are described which are aimed at quantifying the relative contribution of divertor leakage and radial ion transport on neutral pressures surrounding the core plasma. Evidence is presented implying that cross-field transport competes with, or dominates, parallel transport in such a way that plasma exists far out in the scrape-off layer (SOL) in the shadow of limiters and recycles on main chamber surfaces. The transition from L- to H-mode core confinement does not affect the far SOL characteristics nor the scaling of pressures around the plasma. Variations in magnetic equilibrium are used to vary the divertor pressures independent of the core plasma. Based on the analysis of such experiments using a simple neutral flow model we estimate that neutrals escaping from leaks in the lower (closed) divertor during lower x-point operation contribute a smaller fraction (~10-30%) of the midplane pressure than main chamber recycling. The inferred leakage is much larger from the upper (open) divertor during upper x-point operation. Most neutrals escaping from either divertor do not directly travel to the midplane. Instead, they are redirected, most likely by some combination of ionization and/or collisions (elastic, charge exchange). (35 refs.)
Maingi, R., M. G. Bell, et al. (2002). "Characteristics of the first H-mode discharges in the National Spherical Torus Experiment." Physical Review Letters 88(3): 035003-4.
We report observations of the first low-to-high (L-H) confinement mode transitions in the National Spherical Torus Experiment. The H-mode energy confinement time increased over reference discharges transiently by 100-200%, as high as ~100 ms. This confinement time is ~2 times higher than predicted by a multimachine scaling. Thus the confinement time of spherical tori has been extended to a record high value, leading to an eventual revision of confinement scalings. Finally, the power threshold for H-mode access is >10* higher than predicted by an international scaling from conventional aspect-ratio tokamaks, which could lead to new understanding of H-mode transition dynamics. (21 refs.)
Majeski, R., G. Antar, et al. (2002). A toroidal liquid lithium limiter for CDX-U. 19th Symposium on Fusion Engineering (SOFE), ATLANTIC CITY, NEW JERSEY, I E E E.
Attention has focused recently on flowing liquid lithium as a first wall for a reactor because of its potentially attractive physics and engineering features. In order to test the suitability of liquid lithium as a plasma facing component, the Current Drive eXperiment - Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory has recently installed a fully toroidal liquid lithium limiter. CDX-U is a compact (R=34cm, a=22 cm, B-toroidal = 2 kG, I-p =100 kA, T-c(0)similar to100 eV, n(c)(0)similar to 5 X 10(19) m(-3)) short-pulse (<25 msec) spherical torus (ST) with extensive diagnostics. The limiter, which consists of a shallow circular stainless steel tray of radius 34 cm and width 10 cm, is filled with lithium to a depth of a few millimeters, and forms the lower limiting surface for the discharge. Heating elements beneath the tray are used to liquefy the lithium (melting point = 180.5degrees C) prior to the experiment. The total area of liquid lithium exposed to the plasma is approximately 2000 cm(2). The design of the limiter, modifications to CDX-U to accommodate in-vessel inventories of approximately 1 liter of liquid lithium, techniques for loading lithium onto the limiter, and other preparations will be described. CDX-U has previously been successfully operated with a smaller area (300 cm(2)) liquid lithium rail limiter. Diagnostics specific to lithium operations include muItichord spectrometry of the 135 Angstrom LiIII line in the core plasma, monitors for neutral lithium light at the lithium limiter, and a fast (10,000 frame per second) camera which monitors motion of the liquid during the discharge. First results of plasma operations with the toroidal liquid lithium limiter will also be given.
Maki, S., S. Masamune, et al. (2002). "Superthermal electron diffusion processes in a reversed field pinch." Journal of the Physical Society of Japan 71(7): 1680-3.
In reversed field pinches (RFPs), edge fast electrons play some important roles in the energy transport and RFP dynamo. There have been here detailed measurements of electron currents and floating potential with electrostatic probes at the edge region, showing that field-aligned current is carried by superthermal electrons, which results in reversed toroidal current outside the reversal surface. The time evolution of the radial electric field, which is estimated from that of floating potential measured with insertable probes with some assumptions, revealed that radially outward electric field is formed in about 100 mu s. Given that the radial electric field is a result of superthermal electron diffusion in a stochastic magnetic field line, then the estimated diffusion coefficient agrees well with the numerically calculated values, which were obtained by using Monte-Carlo particle orbit calculation in combination with a 3-dimensional nonlinear MHD simulation. It is concluded that the radial electric field is formed by stochastic diffusion of superthermal electrons, which also play important roles in RFP dynamo. (14 refs.)
Mantsinen, M. J., M. L. Mayoral, et al. (2002). "Alpha-tail production with ion-cyclotron-resonance heating of He-4-beam ions in JET plasmas - art. no. 105002." Physical Review Letters 88(10): 105002-5002.
Third-harmonic ion-cyclotron-resonance heating of He-4-beam ions has produced for the first time on the JET tokamak high-energy populations of He-4 ions to simulate 3.5 MeV fusion-born alpha (alpha) particles. Acceleration of He-4 ions to the MeV energy range is confirmed by gamma-ray emission from the nuclear reaction Be-9(alpha, ngamma) C-12 and excitation of Alfven eigenmodes. Concomitant electron heating and sawtooth stabilization are observed. The scheme could be used in next-step tokamaks to gain information on trapped alpha particles and to test a diagnostics in the early nonactivated phase of operation.
Marchenko, V. S. (2002). "Effect of slowing down on the ripple-induced transport of alpha particles." Physics of Plasmas 9(7): 3031-4.
The adiabatic convection of slowing-down alpha particles that are trapped in the resonance with toroidal field ripples is considered. It is shown that under certain conditions this transport mechanism dominates over the standard "superbanana" diffusion. In particular, it can be responsible for the "delayed" losses of partially thermalized alphas in the Tokamak Fusion Test Reactor [Proceedings of the 14th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Wurzburg, 1992 (IAEA, Vienna, 1992)] deuterium-tritium supershots [Herrmann et al., Nucl. Fusion 37, 293 (1997)]. (15 refs.)
Martin, G. and M. Lipa (2002). "Li{sub 3}: first step studies of a liquid lithium limiter." Fusion Engineering and Design
6th International Symposium on Fusion Nuclear Technology 61/62: 237-43.
Carbon (graphite) is the most popular material in use for today's Tokamak first wall components. However, there are real concerns as to its ability to sustain reactor grade conditions, and alternative concepts are envisaged for future fusion reactors. One is the use of a liquid metal surface, which offers several attractive advantages: direct removal of the heat, high pumping capability, insensibility to neutron irradiation, self repairing in case of over-heating, etc. A preliminary study of a liquid lithium limiter has been carried out for Tore Supra, the only Tokamak that has been equipped from its first day of operation with actively cooled first wall elements (with carbon and CFC surfaces). This study is for a relatively small limiter module, with the aim of characterising the interaction of a high performance plasma with a liquid surface in fusion relevant conditions. Up to 5 l/s of liquid lithium flows over a 0.5 m{sup 2} surface, during 20-100 s. The heat removal capability of such a module could reach I MW steady state, with a maximum heat flux of 6 MW/m. (6 refs.)
Matsuda, T., T. Totsuka, et al. (2002). "Data processing and analysis systems for JT-60U." Fusion Science and Technology 42(2/3): 512-20.
The JT-60U data processing system is a large computer complex gradually modernized by utilizing progressive computer and network technology. A main computer using state-of-the-art CMOS technology can handle ~550 MB of data per discharge. A gigabit ethernet switch with FDDI ports has been introduced to cope with the increase of handling data. Workstation systems with VMEbus serial highway drivers for CAMAC have been developed and used to replace many minicomputer systems. VMEbus-based fast data acquisition systems have also been developed to enlarge and replace a minicomputer system for mass data. The JT-60U data analysis system is composed of a JT-60U database server and a JT-60U analysis server, which are distributed UNIX servers. The experimental database is stored in the 1TB RAID disk of the JT-60U database server and is composed of ZENKEI and diagnostic databases. Various data analysis tools are available on the JT-60U analysis server. For the remote collaboration, technical features of the data analysis system have been applied to the computer system to access JT-60U data via the Internet. Remote participation in JT-60U experiments has been successfully conducted since 1996. (33 refs.)
McLean, H. S., S. Woodruff, et al. (2002). "Suppression of MHD fluctuations leading to improved confinement in a gun-driven spheromak." Physical Review Letters 88(12): 125004-4.
Magnetic fluctuations have been reduced to ~1% during discharges on the Sustained Spheromak Physics Experiment by shaping the spatial distribution of the bias magnetic flux in the device. In the resulting quiescent regime, the safety factor profile is nearly flat in the plasma and the dominant ideal and resistive MHD modes are greatly reduced. During this period, the temperature profile is peaked at the magnetic axis and maps onto magnetic flux contours. Energy confinement time is improved over previous reports in a driven spheromak. (21 refs.)
Morita, S., Y. Shirai, et al. (2002). "Observation of ablation and acceleration of impurity pellets in the presence of energetic ions in the CHS heliotron/torsatron." Nuclear Fusion 42(7): 876-80.
Hydrocarbon pellets were injected into NBI and ECH plasmas of the Compact Helical System (CHS) heliotron/torsatron. The ablation of the pellet was observed from horizontal and vertical directions using two CCD cameras, and the velocities of the pellet during ablation were measured with an 11 channel fan array. It was observed that the pellet trajectory was strongly curved in NBI cases, whereas in ECH cases it was straight. When the toroidal direction of the tangential NBI was changed from clockwise to counter-clockwise, the direction of the curved trajectory completely changed. For the first time, it was also found that the pellet was accelerated during the ablation, for example from 270 to 600 m/s. The ablation of the pellet is simulated for the case with fast ions from NBI (40 keV). The results strongly support the idea that the pellet can be mainly ablated by collisions with fast ions. As a result, it is found that the pellet is accelerated by a jet of the ablation cloud resulting from one-side heating due to the toroidally circulating fast ions. (18 refs.)
Ongena, J. and A. M. Messiaen (2002). "Heating, confinement and extrapolation to reactors." Fusion Science and Technology
5th Carolus Magnus Eruo-Summer School on Plasma and Fusion Energy Physics 41(2T): 429-42.
This work discusses confinement in tokamak plasmas and extrapolation to a reactor, heating methods, plasma thermalization and heating and confinement results of direct interest to a future fusion reactor. (25 refs.)
Peterkin, R. E. (2002). "MHD modeling of plasma compression to high pressure with capacitive-driven solid shell implosions." IEEE Transactions on Plasma Science 30(2): 468-475.
Isotropic high-pressure plasma could be used to compress targets for equation of state studies, for generating fusion reactions, for accelerating projectiles to high speed and for other high-energy density endeavors. One way to generate such high-pressure plasma is to compress a modest pressure gas via implosion of a solid shell by a pulsed electrical discharge. The pressure of an ideal gas characterized by a constant ratio of specific heats Gamma that is undergoing adiabatic compression scales as rho {sup Gamma } where rho is the mass density. Whereas Gamma may not be constant for realistic compressing plasmas, it is certainly always greater than unity. Since the ratio of final to initial mass density can be greater for spherical compression than for cylindrical compression, spherical compression offers an opportunity to achieve substantial pressure increase. Quasi-spherical compression of a solid shell has been achieved in the laboratory and, in this paper, we discuss in detail numerical simulations that were used to guide and interpret such experiments. The numerical simulations are performed with MACH2; an unsteady resistive-MHD code for materials that may exist in any of the solid, liquid, gas and plasma states. These simulations illustrate how Mbar pressures can be achieved in the laboratory using presently available pulsed power technology. (21 refs.)
Peterson, B. J., N. Ashikawa, et al. (2002). "Analysis of images of radiation due to plasma-limiter interaction." IEEE Transactions on Plasma Science 30(1): 52-53.
Images of radiation due to plasma interaction with radiation due to plasma-limiter interaction limiting surfaces in the Large Helical Device are presented with an explanation of the diagnostic and the data analysis technique. The images are made using an 10*14 pixel infrared imaging video bolometer with a 15-Hz frame rate and a sensitivity of 0.5 mW/cm{sup 2}. Data from experiments using a wall limiter and an insertable limiter show the localization of the radiation source near the limiting surface. (4 refs.)
Pigarov, A. Y., S. I. Krasheninnikov, et al. (2002). "Tokamak edge plasma simulation including anomalous cross-field convective transport." Physics of Plasmas 9(4): 1287-99.
Multi-fluid two-dimensional transport models such as the UEDGE code model [T.D. Rognlien et al., J. Nucl. Mater. 196-198, 34 (1992)] are widely used in the simulation of tokamak edge plasmas. Usually these models are based on the assumption of anomalous plasma diffusion in the direction perpendicular to magnetic field lines. As will be shown, the pure diffusive cross-field transport model is inadequate and fails to match properly plasma parameters measured both in the scrape-off layer (SOL) and in the divertor of the DIII-D tokamak. Recently it has been suggested that specific nondiffusive transport occurs in the edge plasma [S.I. Krasheninnikov, Phys. Lett. A 283, 368 (2001)]. The nondiffusive transport is incorporated to the UEDGE model by adding the anomalous cross-field convective velocity for plasma species and by prescribing a specific two-dimensional profile to this velocity. A series of highly radiative discharges obtained on the DIII-D tokamak is analyzed using the UEDGE code with the hybrid, convective and diffusive, cross-field transport model. For these discharges, anomalous convective velocity profiles are adjusted until the simulated radial profiles agree with measurements in the SOL and in the divertor. It is found that in order to reproduce most of the extensive experimental data, anomalous plasma convection should play the dominant role in the outboard edge-plasma region. (33 refs.)
Rogister, A. L. (2002). "Drift wave based anomalous transport models." Fusion Science and Technology
5th Carolus Magnus Eruo-Summer School on Plasma and Fusion Energy Physics 41(2T): 251-67.
Energy and particle transport rates in magnetically confined plasmas are often larger than neo-classical transport owing to binary collisions would allow. Anomalous transport, a major road block on the path to an economic fusion reactor, is a consequence of electric and magnetic fluctuations driven to supra thermal levels by various instability mechanisms. The linearly excited modes saturate by inducing a relaxation of the equilibrium profiles towards the marginally stable state, on the one hand, and via various nonlinear interaction mechanisms, on the other hand. Specific instabilities, profile relaxation and nonlinear interaction models are described and their successes and drawbacks are analysed in the light of observed characteristics of plasma confinement. A rough evaluation of the nuclear heating power required to balance the anomalous losses in the International Tokamak Experimental Reactor (ITER) is derived on the basis of the very qualitative mixing length estimate applied to electrostatic drift wave turbulence. Results from large-scale gyro-kinetic simulation codes are discussed. (99 refs.)
Sabbagh, S. A., R. E. Bell, et al. (2002). "Beta-limiting instabilities and global mode stabilization in the National Spherical Torus Experiment." Physics of Plasmas
43rd Annual Meeting of the Division of Plasma Physics of the American-Physical-Society 9(5): 2085-92.
Research on the stability of spherical torus plasmas at and above the no-wall beta limit is being addressed on the National Spherical Torus Experiment [M. Ono et al., Nucl. Fusion 40, 557 (2000)], that has produced low aspect ratio plasmas, R/a~1.27 at plasma current exceeding 1.4 MA with high energy confinement (TauE/TauE_ITER89P>2). Toroidal and normalized beta have exceeded 25% and 4.3, respectively, in q~7 plasmas. The beta limit is observed to increase and then saturate with increasing l{sub i}. The stability factor beta {sub N}/l{sub i} has reached 6, limited by sudden beta collapses. Increased pressure peaking leads to a decrease in beta {sub N}. Ideal stability analysis of equilibria reconstructed with EFIT [L. L. Lao et al., Nucl. Fusion 25, 1611 (1985)] shows that the plasmas are at the no-wall beta limit for the n=1 kink/ballooning mode. Low aspect ratio and high edge q theoretically alter the plasma stability and mode structure compared to standard tokamak configurations. Below the no-wall limit, stability calculations show the perturbed radial field is maximized near the center column and mode stability is not highly effected by a nearby conducting wall due to the short poloidal wavelength in this region. In contrast, as beta reaches and exceeds the no-wall limit, the mode becomes strongly ballooning with long poloidal wavelength at large major radius and is highly wall stabilized. In this way, wall stabilization is more effective at higher beta in low aspect ratio geometry. The resistive wall mode has been observed in plasmas exceeding the ideal no-wall beta limit and leads to rapid toroidal rotation damping across the plasma core. (36 refs.)
Salzedas, F., S. Hokin, et al. (2002). "Evolution of electron temperature during the energy quench of a major plasma disruption." IEEE Transactions on Plasma Science 30(1): 80-81.
An electron cyclotron emission (ECE) heterodyne radiometer and a Thomson scattering (TS) system were used to measure the electron temperature (T{sub e}) evolution in plasmas of the Rijnhuizen Tokamak Project (RTP) during the abrupt loss of energy confinement in high-density limit disruptions with unprecedented high resolution. (4 refs.)
Schilham, A. M. R., G. M. D. Hogeweij, et al. (2002). "Application of the RTP transport model to the JET tokamak." Nuclear Fusion 42(5): 581-90.
The peculiarities of the electron temperature profile under dominant onand off-axis heating in the Rijnhuizen Tokamak Project (RTP) were successfully described by a 'shell model'. This model features electron transport barriers that are associated with simple rational values of the safety factor. In the present work, this model is tested against JET data. To this end, the model is incorporated into the JETTO code, which is used to simulate JET optimized shear discharges with pronounced electron transport barriers. It is found that the model is successful in reproducing the time and location of barrier formation, as well as the subsequent evolution of barrier location and strength. (33 refs.)
Skinner, C. H., H. W. Kugel, et al. (2002). "Effect of boronization on ohmic plasmas in NSTX." Nuclear Fusion 42(3): 329-32.
Boronization of NSTX has allowed access to higher density higher confinement plasmas. A glow discharge with 4 mtorr helium and 10% deuterated trimethyl boron deposited 1.7 g of boron on the plasma facing surfaces. Ion beam analysis of witness coupons showed a B+C areal density of 10{sup 18} cm{sup -2} corresponding to a film thickness of 100 nm. Subsequent ohmic discharges showed oxygen emission lines reduced by a factor of 15, carbon emission reduced by a factor of two and copper reduced to undetectable levels. After boronization, the plasma current flat-top time increased by 70% enabling access to higher density higher confinement plasmas. (22 refs.)
Smolyakov, A. I., E. Lazzaro, et al. (2002). "Role of the shear flow profile on the stability of magnetic islands." Physics of Plasmas 9(1): 371-4.
Plasma flow affects the stability of a magnetic island via modification of the ion inertial current. It is shown here that certain profiles of plasma velocity with shear may provide a stabilizing influence on the magnetic island. Such profiles of the plasma flow are characterized by finite plasma circulation inside a magnetic island. (14 refs.)
Sonato, P., V. Antoni, et al. (2002). "Particle control systems at the edge of RFP experiments." Plasma Physics and Controlled Fusion
IAEA Technical Committee Meeting on Divertor Concepts 44(6): 627-38.
Plasma performance in reversed field pinch (RFP) devices, as in the tokamak, is strongly affected by neutrals at the edge. So far only a few experiments have been dedicated to an active control of the neutral particle using conventional solutions of axisymmetric magnetic divertors or throat limiters. The alternative 'vented pump limiter' concept is more attractive for an RFP experiment due to the edge plasma and confinement properties of this magnetic configuration. In this paper, the application of a vented pump limiter to an RFP is discussed and the prototype module of the vented pump limiter designed for the RFX experiment is presented. Finally, the optimization of this concept for a next step RFP device is presented. (34 refs.)
Stotler, D. P., C. S. Pitcher, et al. (2002). "Understanding of neutral gas transport in the Alcator C-Mod tokamak divertor." Aip Conference Proceedings
13th APS Topical Conference on Atomic Processes in Plasmas 635: 251-260.
A series of experiments on the effect of divertor baffling on the Alcator C-Mod tokamak provides stringent tests on models of neutral gas transport in and around the divertor region. One attractive feature of these experiments is that a trial description of the background plasma can be constructed from experimental measurements using a simple model, allowing the neutral gas transport to be studied with a stand-alone code. The neutral-ion and neutral-neutral elastic scattering processes recently added to the DEGAS 2 Monte Carlo neutral transport code permit the neutral gas flow rates between the divertor and main chamber to be simulated more realistically than before. Nonetheless, the simulated neutral pressures are too low and the deuterium Balmer-alpha emission profiles differ qualitatively from those measured, indicating an incomplete understanding of the physical processes involved in the experiment. Some potential explanations are examined and opportunities for future exploration are highlighted. Improvements to atomic and surface physics data and models will play a role in the latter.
Sugie, T., T. Hatae, et al. (2002). "Diagnostics system of JT-60U." Fusion Science and Technology 42(2/3): 482-511.
The diagnostic system of JT-60U (JT-60 upgrade) is composed of about 50 individual diagnostic devices. Recently, the detailed radial profile measurements of plasma parameters have been improved, so that the internal structure of plasmas could be explored. The understanding of plasma confinement has been enhanced by density and temperature fluctuation measurements using a mm-wave reflectometer and electron cyclotron emission measurements, respectively. In addition, real-time control experiments of electron density, neutron yield, radiated power, and electron temperature gradient have been carried out successfully by corresponding diagnostic devices. These measurements and the real-time control contribute to improving plasma performance. Diagnostic devices for next generation fusion devices such as a CO{sub 2} laser interferometer/polarimeter and a CO{sub 2} laser collective Thomson scattering system have been developed. (175 refs.)
Synakowski, E. J., M. G. Bell, et al. (2002). "Initial studies of core and edge transport of NSTX plasmas." Plasma Physics and Controlled Fusion
8th IAEA Technical Committee Meeting on H-Mode Physics and Transport Barriers 44: A165-73.
Rapidly developing diagnostic, operational, and analysis capability is enabling the first detailed local physics studies to begin in high plasmas of the National Spherical Torus Experiment (NSTX). These studies are motivated in part by the observation of energy confinement times in neutral-beam-heated discharges that are favourable with respect to predictions from the ITER-89P scaling expression. For plasmas heated with neutral beam injection (NBI), analysis based on profile measurements suggests that electron heat conduction is the dominant thermal loss channel. Cases where early analysis indicates that ion thermal conduction may be exceptionally low is motivating studies of possible sources of ion heating not presently accounted for by classical collisional processes. Gyrokinetic microstability studies indicate that long wavelength turbulence with k{sub theta } rho {sub i}~0.1-1 may be stable or suppressed by E * B shear in these plasmas, while modes with k{sub theta } rho {sub i}~50 may be robust. High harmonic fast wave (HHFW) heating efficiently heats electrons on NSTX, and studies have begun using it to assess transport in the electron channel. Regarding edge transport, H-mode transitions occur with either NBI or HHFW heating. The power required for L- to H-mode transitions far exceeds that expected from empirical ELM-free H-mode scaling laws derived from moderate aspect ratio devices. Finally, initial fluctuation measurements made with two techniques are permitting the first characterizations of edge turbulence. (26 refs.)
Taccetti, J. M., T. P. Intrator, et al. (2002). "Magnetic field measurements inside a converging flux conserver for magnetized target fusion applications." Fusion Science and Technology 41(1): 13-23.
Two experiments showing continuous, real-time measurements of the radial convergence of a high-aspect-ratio aluminum flux conserver are presented. These results were obtained by measuring the compression of both axial and radial components of an internal low-intensity magnetic field. Repeatable flux conserver compressions of this type, uniform to 10:1 compression ratio, form a step toward achieving magnetized target fusion, where a plasma of appropriate temperature and density would be introduced into the flux conserver for compression to fusion conditions. While X radiographs show this compression ratio was achieved, the magnetic field probe signals were cut off earlier. Axial component measurements resulted in compression ratios of 7:1 and 6.3:1, for the first and second compressions, before the magnetic probe signals were lost. Radial component measurements disagree with the axial probe results. Although the discrepancy between axial and radial probe measurements is not completely understood, possible explanations are presented.
Takahashi, H., E. D. Fredrickson, et al. (2002). "Unusual low frequency magnetic perturbations in the TFTR tokamak." Nuclear Fusion 42(4): 448-85.
Low frequency magnetic perturbations (<or=30 kHz) observed in the Tokamak Fusion Test Reactor (TFTR) tokamak do not always conform to expectations from magnetohydrodynamic (MHD) modes. The discrepancy between observations and expectations arises from the existence of three classes of magnetic perturbation in TFTR: (1) edge originated magnetic perturbations (EOMPs), (2) kink-like modes (KLMs) and (3) tearing modes (TMs). The EOMP class shows unusual magnetic phenomena including up/down asymmetry in poloidal intensity variation that cannot be generated by MHD modes alone. The contributions of MHD modes in plasma edge regions are too small to explain the magnitude of the observed EOMP perturbations. At least two-thirds, and possibly nearly all, of the magnetic perturbations in a typical EOMP originate from sources other than MHD modes. An EOMP has a unity toroidal harmonic number and a poloidal harmonic number close to the q value of the discharge edge. It produces insignificant temperature fluctuations, except possibly in edge regions. The KLM class produces temperature fluctuations, mostly confined within the q=1 surface with an ideal-mode-like structure, but generates only insignificant external magnetic perturbations. The TM class generally conforms to expectations from MHD modes. It is proposed that current flowing in the scrape-off layer (SOL) plasma is a possible origin of EOMPs. (66 refs.)
Taylor, E. D., C. Cates, et al. (2002). "Effect of magnetic islands on the local plasma behavior in a tokamak experiment." Physics of Plasmas 9(9): 3938-45.
Experiments provide simultaneous, local measurements of the pressure and ion velocity perturbations from rotating m/n=2/1 magnetic islands using Mach probes in a tokamak. Measurements were made both inside and around the islands. Pressure perturbations followed the magnetic island motion for both naturally rotating and actively controlled islands. The toroidal ion velocity profile was sharply peaked near the center of the 2/1 magnetic island, and the magnitude of this peak was ~30% of the magnetic island velocity. Active rotation control experiments also successfully changed the ion fluid velocity. The acceleration of the ion fluid was ~20% of that experienced by the magnetic islands. Understanding the effect of magnetic islands on the pressure and ion velocity profiles is crucial for both fundamental plasma studies and the development of more efficient tokamak using advanced tokamak concepts. (48 refs.)
Tobita, K., Y. Kusama, et al. (2002). "Energetic particle experiments in JT-60U and their implications for a fusion reactor." Fusion Science and Technology 42(2/3): 315-26.
Energetic particle experiments in JT-60U are summarized, mainly covering ripple loss and Alfven eigenmodes (AE modes). Significant loss was observed for 85 keV neutral beam injected (NBI) ions and fusion-produced tritons as toroidal field ripple at the plasma surface increased, especially in a reversed shear plasma. Measurement of hot spots on the first wall due to ripple loss confirmed agreement with code predictions, validating the modeling incorporated in an orbit-following Monte Carlo code. A variety of AE modes were destabilized in ion cyclotron range of frequencies (ICRF) minority heating and negative-ion-based NBI (N-NBI) heating. Most of the observed modes are gap modes identified to be toroidicity-induced, ellipticity-induced, and triangularity-induced AE modes. An interesting finding is pulsating modes accompanying frequency sweep, which were destabilized by N-NBI and sometimes induced a beam ion loss of up to 25%. Also presented are energetic particle issues in auxiliary heating with ICRF and N-NBI. (77 refs.)
Vianello, N., M. Spolaore, et al. (2002). "Properties of the edge plasma in the rebuilt Extrap-T2R reversed field pinch experiment." Plasma Physics and Controlled Fusion 44(12): 2513-23.
The edge region of the rebuilt Extrap-T2R reversed field pinch experiment has been investigated using Langmuir probes. Radial profiles of main plasma parameters are obtained and compared with those of the previous device Extrap-T2. The spontaneous setting up of a double shear layer of E*B toroidal velocity is confirmed. The particle flux induced by electrostatic fluctuations is calculated and the resulting effective diffusion coefficient is consistent with the Bohm estimate. A close relationship between electrostatic fluctuations at the edge and non-linear coupling of MHD modes in the core is found. (34 refs.)
Wang, Z. H., C. W. Barnes, et al. (2002). "Large-amplitude electron density and H alpha fluctuations in the sustained spheromak physics experiment." Nuclear Fusion 42(6): 643-652.
New types of toroidally rotating fluctuations (toroidal mode numbers n = 1 and n = 2) of line-integrated electron density and H-alpha emission, with frequencies ranging from 10 to 100 kHz, are observed in the sustained spheromak physics experiment (SSPX). The rotating directions of these fluctuations are the same as the direction determined by E x B, while the E and B directions are determined by the gun voltage and gun magnetic flux polarities, respectively. These results take advantage of one distinctive signature of spheromaks, i.e. it is possible to observe toroidal MHD activity during decay and sustainment at any toroidal angle. A theoretical constraint on line-integrated measurement is proposed and is found to be consistent with experimental observations. Fluctuation analysis in the time and frequency domains indicates that the observed density and H-alpha fluctuations correlate with magnetic modes. Observation of H-alpha fluctuations correlating with magnetic fluctuations indicates that, at least in some cases, MHD n = 1 modes are due to the so-called 'dough-hook' current paths that connect the coaxial gun to the flux conserver, rather than internal kink instabilities. These results also show that electron density and H-alpha emission diagnostics complement other tools for spheromak mode study.
Warrier, M., R. Srinivasan, et al. (2002). "Analytical solutions for hole-like equilibria for arbitrary aspect ratio high beta tokamaks." Physics of Plasmas 9(7): 3075-81.
Analytical solutions for ultrahigh beta ( epsilon beta {sub p}>>1), high safety factor q [q>O(1)], hole-like equilibria with shaped boundary have been derived for arbitrary aspect ratio tokamaks. These solutions are similar to those derived by Cowley et al. [Phys. Fluids B 3, 2066 (1991)] in the large aspect ratio limit. A high beta, high q scaling (similar to that used by Hsu et al. [Phys. Plasmas 3, 266 (1996)]) is used to obtain the various features of finite aspect ratio tokamak equilibrium with a core region occupying most of the volume where psi identical to psi (R) and a boundary layer region of thickness 1/q. It is seen from the analytical solution that there exists a critical poloidal beta value ( beta {sub c}) beyond which one can have hole-like equilibria. The dependency of beta {sub c} on the inverse aspect ratio ( epsilon ) is also studied. (13 refs.)
Watkins, M. L. (2002). "Optimizing confinement in a tokamak." Plasma Physics and Controlled Fusion
29th EPS Conference on Plasma Physics and Controlled Fusion 44: B173-B192.
Significant progress has been made in recent years in achieving levels of energy confinement in existing tokamaks which scale to that required for ITER. In particular, high confinement is achieved routinely in the plasma edge and in the plasma core, leading to steep density and temperature gradients. These gradients can drive non-inductive currents which could reduce significantly the requirements for externally applied current drive in a steady state tokamak. However, high confinement can also lead to deleterious effects related to pressure and current driven magnetohydrodynamic instabilities and to impurity accumulation. Experimental effort is now turning to the real time control of plasma and current profiles to facilitate the achievement of high confinement and to prevent the deleterious effects which could otherwise inhibit the development of a fully coherent operating scenario for a reactor.
Xu, X. Q., W. M. Nevins, et al. (2002). "Dynamical simulations of boundary plasma turbulence in divertor geometry - art. no. 53." New Journal of Physics 4: 53.
Direct comparisons between numerical simulations and the measured plasma fluctuations and transport are presented by performing nonlinear two-fluid simulations with the BOUT code (Xu X Q and Cohen R H 1998 Contrib. Plasma Phys. 38 158). BOUT models boundary-plasma turbulence in a realistic divertor geometry using modified Braginskii equations for plasma vorticity, density (n(i)), electron and ion temperature (T-e, T-i) and parallel momenta. The BOUT code solves for the plasma fluid equations in a 3D toroidal segment, including the region somewhat inside the separatrix and extending into the scrape-off layer; the private flux region is also included. In this paper, the physics of resistive X-point turbulence and its relation to flow shear generation is discussed. We present comparisons between the boundary plasma turbulence observed in the BOUT code and experiments on DIII-D (Luxon J L et al 1986 Int. Conf. on Plasma Physics and Controlled Nuclear Fusion (Vienna: IAEA) p 159), the National Spherical Torus Experiment (Peng Y-K M 2000 Phys. Plasmas 7 1681), and C-Mod (Hutchinson I H et al 1994 Phys. Plasmas 1 1511). In an L-mode discharge in the DIII-D tokamak, both BOUT simulations and beam emission spectroscopy show a similar flow pattern and blob size across the last closed flux surface. In an L-mode discharge, both BOUT simulations and gas puff imaging show similar filament structures along the field line and similar frequency spectrum at the outboard midplane. In simulations of the quasi-coherent mode in the EDA regime of C-Mod, the particle flux measured from BOUT simulation is consistent with Langmuir probe measurements on C-Mod at the midplane near the separatrix. The qualitative comparisons thus indicate that BOUT contains much of the relevant physics for boundary plasma turbulence in the experimentally relevant X-point divertor geometry of present-day tokamaks and spherical tori.
Xu, X. Q., R. H. Cohen, et al. (2002). "Turbulence simulations of X point physics in the L-H transition." Nuclear Fusion 42(1): 21-7.
The resistive X point mode is shown to be the dominant mode in boundary plasmas in X point divertor geometry. The poloidal fluctuation phase velocity from the simulation results of the resistive X point turbulence shows a structure across the separatrix that is experimentally measured in many fusion devices. The fluctuation phase velocity is larger than the E*B velocity in both L and H mode phases. It is also demonstrated that there is a strong poloidal asymmetry of particle flux in the proximity of the separatrix. Turbulence suppression in the L-H transition results when sources of energy and particles drive sufficient gradients, as in the experiments. (44 refs.)
Yoo, S. J. and M. Kwon (2002). Periscope systems for infrared imaging on Korea superconducting Tokamak advanced research (KSTAR). International Conference on Advanced Diagnostics for Magnetic and Inertial Fusion, VARENNA, ITALY, KLUWER ACADEMIC PUBL.
Zhang, Q. M., Q. S. Wang, et al. (2002). "A new scintillating-fiber-array neutron detector." Nuclear Instruments & Methods in Physics Research Section A-Accelerators Spectrometers Detectors and Associated Equipment 486(3): 708-715.
A new scintillating-fiber-array neutron detector has been developed. The detector consists of a bee-hive-shaped lead absorber, a scintillating fiber array, a light guide, a filter and a photomultipfier tube. The experimental results show that the new detector's neuron-to-gamma sensitivity ratio is improved about six times compared to traditional plastic scintillation detectors to 2.5 MeV neutrons and 1.25 MeV gamma rays. Hence, the detector should be very useful in the measurements of pulsed neutrons from fission reactions in a neutron-gamma mixed field. (C) 2001 Elsevier Science B.V. All rights reserved.
Zweben, S. J., D. P. Stotler, et al. (2002). "Edge turbulence imaging in the Alcator C-Mod tokamak." Physics of Plasmas
43rd Annual Meeting of the Division of Plasma Physics of the American-Physical-Society 9(5): 1981-9.
The two-dimensional (2D) radial vs poloidal structure of edge turbulence in the Alcator C-Mod tokamak [I. H. Hutchinson, R. Boivin, P. T. Bonoli et al., Nucl. Fusion 41, 1391 (2001)] was measured using fast cameras and compared with three-dimensional numerical simulations of edge plasma turbulence. The main diagnostic is gas puff imaging, in which the visible D{sub alpha } emission from a localized D{sub 2} gas puff is viewed along a local magnetic field line. The observed D{sub 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 2D structure motion has also been measured using an ultrafast framing camera with 12 frames taken at 250 000 frames/s. 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. (32 refs.)
Afanasyev, V. I., A. Gondhalekar, et al. (2003). "Neutral particle analyzer/isotope separator for measurement of hydrogen isotope composition of JET plasmas." Review of Scientific Instruments 74(4): 2338-52.
This article describes a neutral particle analyzer/isotope separator (ISEP) developed for measurement of the relative hydrogen isotope composition of Joint European Torus (JET) plasmas. The ISEP deployed on the JET can be regarded as a prototype of an instrument proposed for measurement of the spatial profile of the ratio of the density of deuterium and tritium ions in the plasma, n{sub D}(r)/n{sub T}(r), in the International Thermonuclear Experimental Reactor (ITER). The ISEP makes simultaneous measurements of the energy distribution of efflux of hydrogen isotope atoms (H, D, and T) from the plasma. From such measurements it is possible to deduce the radial profile of the relative hydrogen isotope ion composition of the plasma and radial transport of ions of one isotope across the plasma of another isotope species. The main elements of the ISEP are (a) use of a thin carbon foil for reionization of the incident atoms, thereby eliminating gas stripping cells and gas sources of conventional neutral particle analyzers (NPAs), (b) acceleration of secondary ions in order to access the regime of higher detection efficiency of the NPA and to better separate ion pulses from neutron/ gamma -ray induced pulses in scintillator detectors, (c) E||B dispersion of the secondary ions in specially designed nonuniform magnetic and electric fields to provide focusing in the detector plane, increased throughput and greater contrast between neighboring isotopes, and (d) counting of energy and mass analyzed secondary ions using detectors consisting of thin [1<or=t ( mu m)<or=7] CsI(Tl) scintillators deposited directly on miniature thin window photomultiplier tubes mounted in vacuum. The ISEP has high contrast between atoms of neighboring masses (>or=10{sup 3} for E approximately=5 keV and much greater at higher energies), and high detection efficiency (0.06<or= epsilon <or=0.83 for atoms of 5<or=(keV)<or=150). ISEP detectors have very low sensitivity to neutrons and gamma rays (<or=10{sup -7} of ion sensitivity), making it feasible to use the ISEP in JET DT experiments without any shielding. Only a modest amount of neutron/ gamma -ray shielding would be required in the ITER for similar applications of the ISEP. The initial experiments on JET plasmas using the ISEP demonstrate well the capabilities of the instrument for measurement of the hydrogen isotope composition of the plasma and the energy distribution function of isotope ions. (19 refs.)
Akers, R. J., J. W. Ahn, et al. (2003). "Transport and confinement in the mega ampere spherical tokamak (MAST) plasma." Plasma Physics and Controlled Fusion
30th EPS Conference on Controlled Fusion and Plasma Physics 45: A175-A204.
A combination of recently installed state-of-the-art imaging and profile diagnostics, together with established plasma simulation codes, are providing for the first time on Mega Ampere Spherical Tokamak (MAST) the tools required for studying confinement and transport, from the core through to the plasma edge and scrape-off-layer (SOL). The H-mode edge transport barrier is now routinely turned on and off using a combination of poloidally localized fuelling and fine balancing of the X-points. Theory, supported by experiment, indicates that the edge radial electric field and toroidal flow velocity (thought to play an important role in H-mode access) are largest if gas fuelling is concentrated at the inboard side. H-mode plasmas show predominantly type III ELM characteristics, with confinement H-H factor (w.r.t. scaling law IPB98[y, 2]) around similar to1.0. Combining MAST H-mode data with the International Tokamak Physics Activities (ITPA) analyses, results in an L-H power threshold scaling proportional to plasma surface area (rather than P-LH similar to R-2). In addition, MAST favours an inverse aspect ratio scaling P-LH similar to epsilon(0.5). Similarly, the introduction of type III ELMing H-mode data to the pedestal energy regression analysis introduces a scaling W-ped similar to epsilon(-2.13) and modifies the exponents on R, B-T and kappa. Preliminary TRANSP simulations indicate that ion and electron thermal diffusivities in ELMing H-mode approach the ion-neoclassical level in the half-radius region of the plasma with momentum diffusivity a few times lower. Linear flux-tube ITG and ETG microstability calculations using GS2 offer explanations for the near-neoclassical ion diffusivity and significantly anomalous electron diffusivity seen on MAST. To complement the baseline quasi-steady-state H-mode, newly developed advanced regimes are being explored. In particular, 'broad' internal transport barriers (ITBs) have been formed using techniques developed at conventional aspect ratio. Electron and ion energy diffusivities are reduced towards the ion-neoclassical level in the ITB region of both co-and counter-injection NBI heated plasmas, with momentum diffusivity up to 10 times lower. Moving out to the edge and SOL, OSM2/EIRENE modelling is being used to extract edge perpendicular particle and heat diffusivities, results being consistent with the ballooning nature of power-flow seen in L-mode and reduction in outboard turbulence seen in ELM-free and inter-ELM H-mode. Modelling of parallel SOL transport requires the inclusion of the mirror force (similar to10 times higher in MAST than at the conventional aspect ratio) and B2SOLPS5.0 simulations show the edge electric field to be well modelled by neoclassical theory. Transient edge transport phenomena are being studied in detail using a variety of techniques (e.g. probability density function (PDF) and power spectrum analysis, differencing and rescaling methods). Intermittent transport is associated with a radial efflux at up to a tenth of the sound speed and up to 30 cm from the separatrix. Arguably, the most dramatic edge events seen in the plasma periphery are the ELMs. Recent results using fast, high-resolution visible imaging confirm the hypothesis that ELMs have both poloidal and toroidal structures (n similar to 10 at q = 4), consistent with recent theories of the non-linear evolution of ballooning modes.
Biewer, T. M., C. B. Forest, et al. (2003). "Electron heat transport measured in a stochastic magnetic field." Physical Review Letters 91(4): 045004-4.
New profile measurements have allowed the electron thermal diffusivity profile to be estimated from power balance in the Madison Symmetric Torus where magnetic islands overlap and field lines are stochastic. The measurements show that (1) the electron energy transport is conductive not convective, (2) the measured thermal diffusivities are in good agreement with numerical simulations of stochastic transport, and (3) transport is greatly reduced near the reversal surface where magnetic diffusion is small. (23 refs.)
Boedo, J. A., D. Rudakov, et al. (2003). "Intermittent convection and transport in the boundary of DIII-D." Revista Mexicana de Fisica
3rd Congress of the Venezolano-de-Fisica 49: 123-125.
It is found that 50% of the particle transport in the edge and scrape-off-layer (SOL) of the DIII-D tokamak is due to intermittent plasma objects (IPOs) that can possess higher density and temperature than the surrounding plasma and move radially and poloidally at up to 1500 m/s and up to 5000 m/s respectively.
Boedo, J. A., D. L. Rudakov, et al. (2003). "Transport by intermittency in the boundary of the DIII-D tokamak." Physics of Plasmas
44th Annual Meeting of the Division of Plasma of the American-Physical-Society 10(5): 1670-7.
Intermittent plasma objects (IPOs), featuring higher pressure than the surrounding plasma, are responsible for ~50% of the E*B{sub T} radial transport in the scrape off layer (SOL) of the Doublet III D (DIII-D) tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] in L- and H-mode discharges. Conditional averaging reveals that the IPOs are positively charged and feature internal poloidal electric fields of up to 4000 V/m. The IPOs move radially with E*B{sub T}/B{sup 2} velocities of ~2600 m/s near the last closed flux surface (LCFS), and ~330 m/s near the wall. The IPOs slow down as they shrink in size from 2 cm at the LCFS to 0.5 cm near the wall. The skewness (i.e., asymmetry of fluctuations from the average) of probe and beam emission spectroscopy data indicate IPO formation at or near the LCFS and the existence of positive and negative IPOs which move in opposite directions. The particle content of the IPOs at the LCFS is linearly dependent on the local density and decays over ~3 cm into the SOL while their temperature decays much faster (~1 cm). (53 refs.)
Bush, C. E., M. G. Bell, et al. (2003). "H-mode threshold and dynamics in the National Spherical Torus Experiment." Physics of Plasmas
44th Annual Meeting of the Division of Plasma of the American-Physical-Society 10(5): 1755-64.
Edge parameters play a critical role in high confinement mode (H-mode) access, which is a key component of discharge optimization in present day toroidal confinement experiments and the design of next generation devices. Because the edge magnetic topology of a spherical torus (ST) differs from a conventional aspect ratio tokamak, H-modes in STs exhibit important differences compared with tokamaks. The dependence of the National Spherical Torus Experiment (NSTX) [C. Neumeyer et al., Fusion Eng. Des. 54, 275 (2001)] edge plasma on heating power, including the low confinement mode (L-mode) to H-mode (L-H) transition requirements and the occurrence of edge-localized modes (ELMs), and on divertor configuration is quantified. Comparisons between good L-modes and H-modes show greater differences in the ion channel than the electron channel. The threshold power for the H-mode transition in NSTX is generally above the predictions of a recent International Tokamak Experimental Reactor (ITER) [ITER Physics Basis Editors, Nucl. Fusion 39, 2175 (1999)] scaling. Correlations of transition and ELM phenomena with turbulent fluctuations revealed by gas puff imaging and reflectometry are observed. In both single-null and double-null divertor discharges, the density peaks off-axis, sometimes developing prominent "ears" which can be sustained for many energy confinement times, tau {sub E}, in the absence of ELMs. A wide variety of ELM behavior is observed, and ELM characteristics depend on configuration and fueling. (28 refs.)
Chen, J. L., J. G. Li, et al. (2003). "The longer life and high performance of lithium containing coatings developed by ICRF in the HT-7 superconducting tokamak." Journal of Nuclear Materials
15th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI-15) 313/316: 140-3.
A new wall conditioning method, lithium containing silicon coatings ( alpha -C:H/Li-Si) in situ realized by means of ion cyclotron range of frequency plasma assisted chemical vapor deposition, has been successfully developed in the HT-7 superconducting tokamak, which leads to not only the effective suppression of carbon and oxygen impurities, but also lower hydrogen recycling than siliconization. After the wall conditioning, the impurity level in the vacuum vessel of HT-7 device measured by QMS and spectroscopy was largely reduced and obviously lower than siliconization and even approaches lithium in situ coatings. The depth profile of deposition was investigated by Auger surface analysis. The decreases of edge plasma temperature and electron density showed that the plasma confinement has been significantly improved comparing with siliconization, and is sustained for nearly 350 shots. (17 refs.)
Chu, D. L., B. Wu, et al. (2003). "Preliminary analysis of advanced equilibrium configuration for the fusion-driven subcritical system." Plasma Science & Technology 5(6): 2085-2092.
The Fusion-Driven Subcritical System (FDS) is a subcritical nuclear energy system drived by fusion neutron source. In this paper, an advanced plasma configuration for FDS system has been proposed, which aims at high beta, high bootstrap current and good confinement. A fixed-boundary equilibrium code has been used to obtain ideal equilibrium configuration. In order to determine the feasibility of FDS operation, a two-dimensional time-dependent free boundary simulation code has been adopted to simulate time-scale evolution of plasma current profile and boundary position. By analyses, the Reversed Shear mode as the most attractive one has been recommended for the FDS equilibrium configuration design.
Chudnovskiy, A. N., Y. Gvozdkov, et al. (2003). "MHD-mode locking by controlled a halo-current in the T-10 tokamak." Nuclear Fusion 43(8): 681-685.
Experiments on a non-disruptive halo-current influence on the m = 2 mode behaviour at the flat-top stage of a tokamak discharge are presented. The halo-current in the rail limiter-plasma-vacuum vessel-external circuit-rail limiter loop was used. An EMF source controlled with a preprogrammed signal or with a feedback m = 2 signal was introduced into the external part of the halo-current circuit. The EMF source has generated the oscillating halo-currents with an amplitude of up to 500 A in the frequency range 0-20 kHz. In the case of the preprogrammed control signal the switching on of the EMF source resulted in the shift of the m = 2 mode frequency to the frequency of the halo-current oscillations. In particular, the rotation of the m = 2 mode stopped under a pulse of zero-frequency halo-current. In the tokamak discharges, when the mode rotation spontaneously stopped before the switching on of the oscillating halo-current, the mode rotation was restored at the halo-current frequency. In the case of the halo-current feedback control by the m = 2 mode signal, the effect depended on the choice of the phase shift in the feedback loop. Some increase or decrease of the m = 2 mode amplitude as well as some variations of the mode frequency were observed at different values of the phase shift. The halo-current effect on the m = 2 mode behaviour can be attributed to a coupling between the m/n = 2/1 magnetic islands and the halo-current magnetic field. The experiment was simulated on the assumption that the tearing mode is affected by the halo-current magnetic field component with the same helicity. In the calculations for the T-10 conditions, the mode behaviour under the effect of the halo-current was similar to the experimental observations.
Coelho, R. and F. Porcelli (2003). "Ideal stability of an elliptical plasma column in the presence of external feedback currents." Physics of Plasmas 10(4): 930-3.
The magnetohydrodynamic stability of a plasma column with an elliptical cross section to axisymmetric, rigid plasma displacements is analyzed. The considered plasma equilibrium, characterized by the presence of a double-null magnetic separatrix, is found to be unstable. An application of the well-known Energy Principle [I. B. Bernstein, E. A. Frieman, M. D. Kruskal, and R. M. Kulsrud, Proc. R. Soc. London, Ser. A 244, 17 (1958)] is presented, taking into account the effect of external localized electric currents on the mode stability. It is shown that such currents can stabilize the axisymmetric displacements if their values exceed a critical threshold. (17 refs.)
Cuperman, S., C. Bruma, et al. (2003). "Structure and relative importance of ponderomotive forces and current drive generated by converted fast waves in pre-heated low aspect ratio tokamaks." Physics Letters A 311(2/3): 221-41.
The generation in low aspect ratio tokamaks (LARTs) of ponderomotive forces and non-inductive current drive by the resonant fast wave-plasma interaction with mode conversion to kinetic Alfven waves (KAWs) and subsequent deposition, mainly by resonant electron Landau damping, is considered. The calculations follow the rigorous solution of the full wave equations upon using a dielectric tensor operator consisting of (i) a parallel conductivity including both kinetic effects (collisionless Landau damping on passing electrons) and collisional damping on both trapped electrons and passing electrons + ions and (ii) perpendicular components provided by the resistive two-fluid model equations. The fast waves are launched by an antenna located on the low field side and extending +or-45 degrees about the equatorial plane. A parametric investigation of the structure and importance of the various components of the ponderomotive forces and current drive generated in START-like plasmas is carried out and their suitability for supplementing the required non-rf toroidal equilibrium current is demonstrated. (27 refs.)
Cuperman, S., C. Bruma, et al. (2003). "Solution of the resistive two-fluid wave equations for Alfvenic modes in spherical tokamak plasmas." Journal of Plasma Physics 69: 15-43.
Low aspect ratio tokamak (LART) configurations represent a promising approach to thermonuclear fusion. They require auxiliary non-ohmic power for heating, current drive and turbulent transport suppression. This paper pioneers the pre-requisite for the quantitative evaluation of these effects, namely the formulation and solution of the full wave equation for pre-heated (i.e. prior to the addition of any auxiliary heating or current drive) LARTs, the case in which mode-converted Alfven waves (omega<ω(ci)) are used as an additional non-ohmic power source. Arbitrary aspect ratio and magnetic shear tokamaks with non-circular cross-section are properly considered; this general approach includes, as a particular case, that of low aspect ratio (spherical) tokamaks. The problem is formulated in terms of the vector and scalar potentials (A, &UPhi;), thus avoiding the numerical pollution occurring in the case of the (E, B) formulation. Adequate boundary conditions at the vacuum-metallic wall interface and regularity conditions at the magnetic axis are enforced. For the pre-heated stage, a quite general two-fluid, resistive dielectric tensor operator able to describe the anisotropic plasma response in spherical tokamaks is derived and used; except for its linear character, no geometrical limitations are imposed on it. Consistent equilibrium profiles also including effects of neoclassical conductivity and bootstrap current are used (Wilson 1994). The wave equations are solved with the aid of a suitable 2(1)/(2) finite-element algorithm, taking advantage of the up-down symmetry of the problem. Illustrative solutions obtained for Culham's START device are presented and discussed; this includes a sensitivity study to the antenna parameters (geometry, frequency and wavenumbers). Finally, as one of the possible applications of these results, illustrative ensuing ponderomotive forces and the non-inductive current drive in a START-like device are calculated.
Dylla, H. F. (2003). "Development of ultrahigh and extreme high vacuum technology for physics research." Journal of Vacuum Science & Technology A (Vacuum, Surfaces, and Films) 21(5): S25-33.
Over the last 50 years increasingly large and more sophisticated devices have been designed and put into operation for the study of particle and nuclear physics, magnetic confinement of high temperature plasmas for thermonuclear fusion research, and gravity wave observatories based on laser interferometers. The evolution of these devices has generated many developments in ultrahigh and extreme high vacuum technology that were required for these devices to meet their operational goals. The technologies that were developed included unique ultrahigh vacuum vessel structures, ultrahigh vacuum compatible materials, surface conditioning techniques, specialized vacuum pumps, and vacuum diagnostics. Associated with these technological developments are scientific advancements in the understanding of outgassing limits of UHV-compatible materials and particle-induced desorption effects. (121 refs.)
Ejiri, A., S. Shiraiwa, et al. (2003). "Visible fluctuation measurements on the TST-2 spherical tokamak." Japanese Journal of Applied Physics, Part 1 (Regular Papers, Short Notes & Review Papers) 42(12): 7557-63.
The four-beam correlation method has a potential of measuring fine-scale fluctuations locally. In the method, many lines of sight are measured and the coherences among them are calculated. As a proof of principle experiment, a visible fluctuation measurement system has been fabricated and applied to the TST-2 spherical tokamak plasma. Using this system, fluctuations were measured and high-frequency components (100+or-12.5 kHz) were analyzed. Because of a long parallel correlation length, local information was not obtained directly. However, comparison of the experimental results and those of a model calculation suggests that (i) the fluctuating emission is relatively localized in the peripheral region of the plasma, (ii) the perpendicular correlation length is of the order of 1 cm, while the parallel one is much longer than it and (iii) the magnetic field angle in the peripheral region agrees with that estimated from magnetic measurements. (16 refs.)
Fredrickson, E. D., C. Z. Cheng, et al. (2003). "Wave driven fast ion loss in the national spherical torus experiment." Physics of Plasmas 10(7): 2852-62.
Spherical tokamaks have relatively low toroidal field which means that the fast-ion Larmor radius is relatively large ( rho {sub fi}>0.04 a{sub p}) and the fast ion velocity is much greater than the Alfven speed (V{sub fi}>2 V{sub Alfven}). This regime of large Larmor radius and low Alfven speed is a regime in which fast ion driven instabilities are potentially virulent. It is therefore an important goal of the present proof-of-principle spherical tokamaks to evaluate the role of fast ion driven instabilities in fast ion confinement. This paper presents the first observations of fast ion losses in a spherical tokamak resulting from energetic particle driven modes. Two classes of instabilities are responsible for the losses. Multiple, simultaneously bursting modes in the toroidal Alfven eigenmode frequency gap cause neutron drops of up to 15%. A bursting, chirping mode identified as precession and/or bounce resonance fishbone also causes significant neutron drops. Both modes are usually present when the losses are observed. (62 refs.)
Gao, Z., J. Q. Dong, et al. (2003). "Effects of flow shear on the ion temperature gradient modes in a high beta plasma slab." Physics of Plasmas 10(3): 774-781.
The integral eigenmode equations derived previously for the study of drift instabilities in a sheared slab plasma with arbitrary beta (plasma pressure/magnetic pressure) are extended. These equations are used to investigate the effects of flow shear on the ion temperature gradient driven (ITG) modes. It is found that the destabilizing effect of a parallel velocity shear, V-0('), is weakened by the finite beta effect, especially in case of weak magnetic shear. However, the perpendicular velocity shear, V-E('), still effectively stabilizes these modes even in high beta regions. A large enough V-E(') can completely stabilize the ITG mode at arbitrary beta. In addition, the effect of V-E(') is highly enhanced in weak magnetic shear regions. When the parallel flow coexists with the perpendicular flow, the comprehensive flow effect depends on the relative sign of these velocity shears. The modes with higher growth rates may be stabilized by a smaller V-E(') for (V0VE')-V-'>0. (C) 2003 American Institute of Physics.
Garstka, G. D., S. J. Diem, et al. (2003). "Performance and stability of near-unity aspect ratio plasmas in the pegasus toroidal experiment." Physics of Plasmas
44th Annual Meeting of the Division of Plasma of the American-Physical-Society 10(5): 1705-11.
The Pegasus Toroidal Experiment [R. Fonck et al., Bull. Am. Phys. Soc. 41, 1400 (1996)] is a spherical torus designed to study the limits of plasma behavior as the aspect ratio A approaches unity. Access to near-unity A is achieved through the use of a novel high-stress reinforced solenoid magnet. High toroidal beta beta {sub t} is obtained in ohmically-heated plasmas by operation at low field with densities up to the Greenwald limit. Values of beta {sub t} up to 20% and normalized beta up to 5 have been obtained. The ratio of plasma current to toroidal field rod current, known as the toroidal field utilization, reaches values as large as 1 but appears to approach a "soft" boundary at that level related to both ohmic flux limitations and the onset of resistive magnetohydrodynamic (MHD) activity. The m/n=2/1 and 3/2 modes are most frequently observed, in agreement with the inferred safety factor profiles. Experiments are beginning to access the external kink stability boundary at edge safety factor q{sub 95}=5, which is significantly higher than that observed in conventional tokamaks. Calculations using the DCON code [A. H. Glasser and M. S. Chance, Bull. Am. Phys. Soc. 42, 1848 (1997)] confirm instability to the ideal kink. (18 refs.)
Gilmore, M., W. A. Peebles, et al. (2003). "Progress toward a practical magnetic field diagnostic for low-field fusion plasmas based on dual mode correlation reflectometry." Review of Scientific Instruments
14th Topical Conference on High-Temperature Plasma Diagnostics 74(3): 1469-72.
Previously, the proof of principle of measurement of magnetic field strength, |B| internal to a plasma by cross correlation of ordinary (O) and extraordinary (X) mode fluctuation reflectometer signals has been demonstrated in a linear plasma device. It was found that dual mode (O-X) reflectometry data could be interpreted by a one-dimensional numerical model to determine |B| in the vicinity of the reflectometer cutoff positions. Radial correlation properties of turbulence are also measured simultaneously with |B|. This technique is potentially well suited to measurement of |B| in low-field fusion devices (e.g., B<0.6 T) where standard magnetic field diagnostics are expected to be difficult to implement. However, transfer of this method to toroidal magnetic fusion devices presents a number of potential difficulties such as the effects of magnetic shear, steep density gradients, and limited time for cross correlations. In addition, recent simulations suggest that two-dimensional modeling may be required to interpret experimental data when density fluctuation levels are high. Dual mode correlation reflectometry experiments on the National Spherical Torus Experiment are presented, and progress toward the implementation of this technique as a practical diagnostic is discussed. (12 refs.)
Goncharov, P. R., T. Saida, et al. (2003). "Development and initial operation of the pellet charge exchange diagnostic on LHD heliotron." Review of Scientific Instruments
14th Topical Conference on High-Temperature Plasma Diagnostics 74(3): 1869-72.
An active corpuscular diagnostic with an artificially created localized target for the charge-exchange process has been developed and tested on a large helical device. The diagnostic is a combination of an impurity pellet injector and a natural diamond detector-based energy analyzer. High-energy particles neutralized at the pellet ablation cloud are detected while the pellet travels across the plasma column. Time-resolved atomic energy spectra translate into local measurements along the pellet trajectory. Thus, local parameters are obtained in the toroidally non-axis-symmetrical configuration. Physical basis, data interpretation, and the technical description of the diagnostic are given. The requirements to the analyzer geometry, electronics, and data acquisition needed to provide the desired spatial resolution are described. The initial experimental results are presented along with the other diagnostic data. The scope of experiments, possible comparisons with multichord passive charge-exchange measurements and future work on the diagnostic are also discussed. (12 refs.)
Grishanov, N. I., A. F. D. Loula, et al. (2003). "Radio-frequency wave dissipation by electron Landau damping in a low aspect ratio D-shaped tokamak." Plasma Physics and Controlled Fusion 45(9): 1791-1803.
Parallel permittivity elements are derived for radio-frequency waves in an axisymmetric tokamak with D-shaped transverse cross-sections of the magnetic surfaces under arbitrary aspect ratio, arbitrary elongation and small triangularity. The drift-kinetic equation is solved separately for untrapped (passing or circulating) and four groups of the trapped particles as a boundary-value problem. Periodicity of the perturbed distribution function over the poloidal angle is used for the untrapped particles, whereas the continuity of the perturbed distribution function at the reflection points (where the parallel velocity is equal to zero) is used for the trapped particles. The bounce resonances are taken into account. A coordinate system with the 'straight' magnetic field lines is used. The dielectric permittivity elements, evaluated in the paper, are suitable to estimate the wave dissipation by electron Landau damping (e.g. during the plasma heating and current drive generation) in the frequency range of Alfven and fast magnetosonic waves, for both the large and low aspect ratio tokamaks with circular, elliptic and D-shaped magnetic surfaces. The dissipated wave power is expressed by the summation of terms including the imaginary parts of both the diagonal and non-diagonal elements of the parallel permittivity. Contributions of the trapped and untrapped electrons to the imaginary part of the parallel permittivity elements have been estimated numerically for typical low aspect ratio tokamak parameters.
Gryaznevich, M., R. J. Akers, et al. (2003). "Next-step-targeted experiments on the mega-amp spherical Tokamak." Physics of Plasmas
44th Annual Meeting of the Division of Plasma of the American-Physical-Society 10(5): 1803-8.
Since its first physics campaign, the principal parameters on MAST (Mega-Amp Spherical Tokamak) [A. Sykes et al., Nuclear Fusion 41, 1423 (2001)] have been brought up towards their design values. Considerable advances have been made in a range of physics areas of direct relevance to the International Thermonuclear Experimental Reactor (ITER) [ITER Physics Basis, Nuclear Fusion 39, 2175 (1999)]. In this paper, results on H-mode access, global confinement and pedestal studies are presented and compared with conventional aspect ratio tokamak scalings. Physics and engineering requirements relevant to next step spherical tokamak devices are discussed, in particular the plasma formation, current ramp-up and sustainment, and plasma exhaust. Results of first experiments directly targeting these issues are presented: Plasma current up to 0.5 MA has been produced without use of the central solenoid flux, and current ramp-up and sustainment without use of the central solenoid flux has been demonstrated. Experiments on neutral beam heating and current drive (CD) demonstrate up to 50% bootstrap current fraction and good CD efficiency, and divertor power loading has been found to be tolerable and have a favorable outboard asymmetry. (23 refs.)
Hill, D. N., R. D. Wood, et al. (2003). "Wall conditioning and power balance for spheromak plasmas in SSPX." Journal of Nuclear Materials
15th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI-15) 313/316: 941-6.
We report here results from power balance measurements for ohmically heated plasmas in the sustained spheromak physics experiment. The plasma is formed inside a close-fitting tungsten-coated copper shell; wall conditioning by baking, glow discharge cleaning (GDC), Ti gettering, and helium shot conditioning produces clean plasmas (Z{sub eff} < 2.5) and reduces impurity radiation to a small fraction of the input energy, except when the molybdenum divertor plate has been overheated. We find that most of the input energy is lost by conduction to the walls (the divertor plate and the inner electrode in the coaxial source region). Recently, carborane was added during GDC to boronize the plasma-facing surfaces, but little benefit was obtained. (14 refs.)
Hoang, G. T., W. Horton, et al. (2003). "Analysis of the critical electron temperature gradient in Tore Supra." Physics of Plasmas 10(2): 405-12.
The Tore Supra [Equipe Tore Supra presented by R. Aymar in Plasma Physics and Controlled Nuclear Fusion Research, Proc. 12th Int. Conf., Nice, 1988 (IAEA, Vienna, 1989), Vol. 1, p. 9] database of fast wave electron heating discharges is analyzed with respect to the role of the critical electron temperature gradient. The experimental evidence for the critical gradient is presented from power balance thermal flux q{sub e}( Del T{sub e}) relation extrapolated to zero in the measured gradient Del T{sub e}. The fluctuation spectra are also known to extrapolate to low levels versus Del T{sub e}. The inferred critical gradient ( Del T{sub e}){sub c} is defined by the offset linear extrapolation of the thermal flux to zero values. Histograms of the anomalous fluxes before and after being normalized to the theoretical models are constructed. For a wide range of heating powers, the electron power balance radial heat flux is shown to be well described by T{sub e}{sup 3/2} (1/L{sub Te}-1/L{sub c}), where L{sub c}=(T{sub e}/ Del T{sub e}){sub c} is the gradient scale length from linear eigenmodes in a sheared magnetic field. The relative deviations of the normalized heat fluxes as a function of the excess of the temperature gradient beyond the critical gradient give one method of assessing the agreement between the models and the data. A second method is the predictive studies in which the transport codes are used to investigate the impact of the critical gradient at low and high radio frequency power levels with the model thermal diffusivities. The dominant dependence of the inferred critical temperature gradient is its proportionality to the magnetic shear as expected from the robust electron Landau damping on thermal electrons in electron temperature gradient turbulence. The evidence for the electrostatic density independent thermal diffusivity scaling versus the finite beta electromagnetic density dependent thermal diffusivity is shown in some detail. The results strongly favor the electromagnetic turbulent diffusivity formula. (37 refs.)
Hoffman, A. L., H. Y. Guo, et al. (2003). "Resistivity scaling of rotating magnetic field current drive in FRCs." Nuclear Fusion 43(10): 1091-100.
Rotating magnetic fields (RMFs) have been used to both form and sustain low density, prolate FRCs in the translation confinement and sustainment (TCS) facility. The two most important factors governing performance are the plasma resistivity, which sets the maximum density for which toroidal current can be maintained, and the energy loss rate, which sets the plasma temperature. The plasma resistivity has been determined by carefully measuring the amount of RMF power absorbed by the FRC. When the ratio of RMF magnitude, B{sub omega }, to external poloidal confinement field, B{sub c}, is high, this resistivity is very adversely affected by the RMF drive process. (10 refs.)
Johnson, D. and N. Team (2003). "Diagnostic development for ST plasmas on NSTX." Plasma Physics and Controlled Fusion 45(11): 1975-87.
Spherical tori (ST) have much lower aspect ratio (a/R) and lower toroidal magnetic field than conventional tokamaks and stellarators. This paper will highlight some of the challenges and opportunities these features pose in the diagnosis of ST plasmas in the National Spherical Torus Experiment (NSTX), and discuss some of the corresponding diagnostic development that is underway. The low aspect ratio necessitates a small centre stack, with tight space constraints and large thermal excursions, complicating the design of magnetic sensors in this region. The toroidal magnetic field on the NSTX is <or=0.6T, making it impossible to use ECE as a good monitor of electron temperature. A promising new development for diagnosing. electron temperature is electron Bernstein wave radiometry, which is currently being pursued on the NSTX. A high-resolution charge exchange recombination spectroscopy system is being installed. Since non-inductive current initiation and sustainment are top-level NSTX research goals, measurements of the current profile J(R) are essential to many planned experiments. On the NSTX several modifications are planned to adapt the motional Stark effect (MSE) technique to lower field, and two novel MSE systems are being prototyped. Several high-speed two-dimensional imaging techniques are being developed for viewing both visible and x-ray emission. The toroidal field is equivalent to the poloidal field at the outside plasma edge, producing a large field pitch (>50 degrees ) at the outer midplane. The large shear in pitch angle makes some fluctuation diagnostics like beam emission spectroscopy very difficult, while providing a means of achieving spatial localization for microwave scattering investigations of high-k turbulence, which are predicted to be virulent for NSTX plasmas. A brief description of several of these techniques will be given in the context of the current NSTX diagnostic set. (33 refs.)
Kaye, S. M., C. E. Bush, et al. (2003). "Low- to high-confinement mode transitions in the national spherical torus experiment." Physics of Plasmas 10(10): 3953-60.
Edge data from plasmas in the National Spherical Torus Experiment (NSTX) [S. Kaye et al., Fusion Technol. 36, 16 (1999)] have been compared to theories of transport suppression that have been used to develop a physics framework for low-confinement (L) to high-confinement (H) mode transitions. The NSTX data were obtained in low aspect ratio (R/a approximately=1.3) discharges taken from a variety of discharge phases, including L-modes, L-H transitions and H-modes with and without edge localized modes. The comparisons show that the group of points taken just before the L-H mode transition are well mixed with the purely L-mode group to within the measurement uncertainties, indicating that changes in these parameters leading up to the transition are subtle. One of the theory parameters, alpha {sub MHD}=-Rq{sup 2} d beta {sub t}/dr, does show a clear threshold ( alpha {sub MHD}=1 to 2) between the H-mode grouping of points and those remaining in the L-mode or taken just prior to the transition. Additionally, there is no evidence for an edge temperature threshold necessary for transitioning into the H-mode. NSTX data indicate further a possible connection between L-H transitions and nonambipolar beam ion losses. (28 refs.)
Krasheninnikov, S. I. and A. I. Smolyakov (2003). "On neutral wind and blob motion in linear devices." Physics of Plasmas 10(7): 3020-1.
It is shown that asymmetry of neutral components going to and from the wall results in a net force, inducing blob plasma polarization and radial blob motion. This mechanism may explain experimental observations of coherent plasma density structures in linear devices. Estimates of the radial blob velocity are in agreement with experimental data. (7 refs.)
Krasheninnikov, S. I., L. E. Zakharov, et al. (2003). "On lithium walls and the performance of magnetic fusion devices." Physics of Plasmas
44th Annual Meeting of the Division of Plasma of the American-Physical-Society 10(5): 1678-82.
It is shown that lithium walls resulting in zero-recycling regimes at the edge of magnetic fusion device can cause dramatic improvements of core plasma performance. The plasma temperature at the wall in these regimes is much larger than in conventional tokamaks. It reduces the core temperature gradient and, thus, related anomalous transport, allowing an increase in the achievable beta to the level ~20%, due to wall stabilization and second stability core. Fusion relevant plasma temperature over entire core and high beta results in a strong enhancement of fusion power density. Modeling of the International Thermonuclear Experimental Reactor performance in zero-recycling regimes shows so significant improvement that fusion power increases with no apparent limits due to elimination of the strong core temperature gradient and associated turbulent transport and due to expansion of the burning zone to the entire cross section. (10 refs.)
Kubota, S., W. A. Peebles, et al. (2003). "Automatic profile reconstruction for millimeter-wave frequency-modulated continuous-wave reflectometry on NSTX." Review of Scientific Instruments
14th Topical Conference on High-Temperature Plasma Diagnostics 74(3): 1477-80.
UCLA operates a set of millimeter-wave/microwave reflectometers on the National Spherical Torus Experiment (NSTX) for routine measurements of electron density profiles and fluctuations. The system has a combined frequency coverage of 12 to 50 GHz (in the bands 12-18, 20-32, and 33-50 GHz) or a corresponding ordinary-mode cutoff range of 1.8*10{sup 12} to 3.1*10{sup 13} cm{sup -3} to cover both the plasma core and edge. Profile measurements via frequency-modulated continuous-wave operation are typically made in O-mode reflectometry, with sweep times down to 50 mu s over the full band. Automated profile analysis of the reflectometry data is available with limited between-shot analysis and full batch analysis capabilities. The reconstruction algorithm uses complex demodulation with the short-time Fourier transform for signal processing. The unknown portion of the edge profile below the lowest cutoff density is modeled by fitting a family of polynomial density profiles to the experimental data. Uncertainties due to edge profile modeling and comparisons to Thomson scattering measurements are discussed. The reconstructed profiles have documented fast events such as L-H transitions and edge-localized modes in NSTX. (21 refs.)
Liu, Y., A. Y. Kostrioukov, et al. (2003). "Plasma emission tomographic reconstruction in the large helical device." Review of Scientific Instruments 74(4): 2312-17.
Both one-dimensional (1D) and two-dimensional (2D) tomographic reconstruction of the total radiation power distribution from the large helical device (LHD) was based on the data obtained from absolute extreme ultraviolet silicon photodiodes. Two arrays (16 and 19 channels) installed in the normal LHD cross section (constant toroidal angle) provided simple and reliable 1D poloidally symmetric radiation profile reconstruction. The data obtained from two other arrays (20 and 20 channels) were used for 2D reconstruction of the radiation distribution in a semitangential plasma cross section. Using a 2D peeling away algorithm, improved by a feedback procedure, enabled reconstruction of several Fourier harmonics at each magnetic flux surface. These measurement and analysis techniques have enabled us to visualize asymmetries in plasma emission due to pellet and gas fueling. (13 refs.)
Maingi, R., M. G. Bell, et al. (2003). "H-mode research in NSTX." Nuclear Fusion
3rd International-Atomic-Energy-Agency Technical Meeting 43(9): 969-74.
H-modes are routinely obtained in the National Spherical Torus Experiment (NSTX) and have become a standard operational scenario. L-H transitions triggered by NBI heating have been obtained over a wide parameter range in I{sub p}, B{sub t}, and n{sub e} in either lower-single-null (LSN) or double-null (DN) diverted discharges. Edge localized modes are observed in both configurations but the characteristics differ between DN and LSN, which also have different triangularities ( delta ). An H-mode duration of 500 ms was obtained in LSN, with a total pulse length of ~1 s. Preliminary power threshold studies indicate that the L-H threshold is between 600 kW and 1.2 MW, depending on the target parameters. Gas injector fuelling from the centre stack (i.e. the high toroidal field side) has enabled routine H-mode access, and comparisons with low-field side (LFS) fuelled H-mode discharges show that the LFS fuelling delays the L-H transition and alters the pre-transition plasma profiles. Gas puff imaging and reflectometry show that the H-mode edge is usually more quiescent than the L-mode edge. Divertor infrared camera measurements indicate up to 70% of available power flows to the divertor targets in quiescent H-mode discharges. (26 refs.)
Maqueda, R. J., G. A. Wurden, et al. (2003). "Gas puff imaging of edge turbulence (invited)." Review of Scientific Instruments
14th Topical Conference on High-Temperature Plasma Diagnostics 74(3): 2020-2026.
The gas puff imaging (GPI) diagnostic can be used to study the turbulence present at the edge of magnetically confined plasmas. In this diagnostic the instantaneous two-dimensional (2D) radial vs poloidal structure of the turbulence is measured using fast-gated cameras and discrete fast chords. By imaging a controlled neutral gas puffs of typically helium or deuterium, the brightness and contrast of the turbulent emission fluctuations are increased and the structure can be measured independently of natural gas recycling. In addition, recent advances. in ultrafast framing cameras allow the turbulence to be followed in time. The gas puff itself does not perturb the edge turbulence and the neutral gas does not introduce fluctuations in the emission that could possibly arise from, a nonsmooth (turbulent) neutral gas puff. Results from neutral transport and atomic physics simulations using the DEGAS 2 code are discussed showing that the observed line emission is sensitive to modulations in both the electron density and the electron temperature. The GPI diagnostic implementation in the National Spherical Torus Experiment (NSTX) and Alcator C-Mod, tokamak is presented together with example results from these two experiments. (C) 2003 American Institute of Physics.
Menard, J. E., M. G. Bell, et al. (2003). "beta -limiting MHD instabilities in improved-performance NSTX spherical torus plasmas." Nuclear Fusion 43(5): 330-40.
Global magnetohydrodynamic (MHD) stability limits in the National Spherical Torus Experiment (NSTX) have increased significantly recently due to a combination of device and operational improvements. First, more routine H-mode operation with broadened pressure profiles allows access to higher normalized and lower internal inductance. Second, the correction of a poloidal field coil induced error-field has largely eliminated locked tearing modes during normal operation and increased the maximum achievable. As a result of these improvements, peak values have reached (not simultaneously) beta {sub T}=35 %, beta {sub N}=6.4, < beta {sub N}>=4.5, beta {sub N}/l{sub i}=10, and beta {sub p}=1.4. High beta {sub p} operation with reduced tearing activity has allowed a doubling of discharge pulse-length to just over 1 s with sustained periods of beta {sub N} approximately=6 above the ideal no-wall limit and near the with-wall limit. Details of the beta -limit scalings and beta -limiting instabilities in various operating regimes are described. (37 refs.)
Mirnov, S. V., V. B. Lazarev, et al. (2003). "Li-CPS limiter in tokamak T-11M." Fusion Engineering and Design
International Workshop on Innovative Concepts for Plasma-Interactive Components 65(3): 455-65.
This paper is a review of the experimental investigations of Li-behavior as limiter material in real tokamak. All experiments were performed in tokamak T-11M with main parameters: plasma current, I{sub p} = 100 kA; duration of the discharge 0.1 s, toroidal magnetic field, B = 1 T; major radius, R = 0.7 m; minor radius, a = 0.19-0.23 m; average electron density, n{sub c} ~ (1 5) * 10{sup 19} m {sup 3} and electron temperature, T{sub c}(0) ~ 0.3-0.5 keV. Two moving limiters with similar geometry were installed for comparison in chamber-the conventional graphite-boron limiter and Li-limiter on the basis of capillary pore system (CPS). The Li-influx to plasma was measured by visible, UV and SX plasma emission. The main experimental results were: (a) no dramatic bursts of lithium injection at heat load close to the tokamak-reactor level, ~ 10 MW/m{sup 2}, were observed, (b) the total lithium erosion from limiter can be explained by deuterium plus lithium ions sputtering (self-sputtering), (c) high lithium radiation during disruptions prevent Li-limiter from high power load, and (d) the solid basis of CPS limiter had no damages after more than 200 shots with disruptions. The main effect of lithium use in T-11M was the rise of the first wall getter properties, i.e. recycling reduction for not only H{sub 2} and D{sub 2} but for He too. The temperature of hydrogen isotopes desorption was 350-400 degrees C and He desorption was 50-100 C. (29 refs.)
Mueller, D., M. Ono, et al. (2003). "Results of NSTX heating experiments." IEEE Transactions on Plasma Science
29th IEEE International Conference on Plasma Science 31(1): 60-7.
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 beta {sub T} of 32%. A variety of MHD phenomena have been observed to limit beta . NSTX has now begun addressing tau {sub E} scaling, 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, 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 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 beta {sub p}. (21 refs.)
Nazikian, R., G. J. Kramer, et al. (2003). "New interpretation of alpha-particle-driven instabilities in deuterium-tritium experiments on the Tokamak Fusion Test Reactor." Physical Review Letters 91(12): 125003-4.
The original description of alpha particle driven instabilities in the Tokamak Fusion Test Reactor in terms of toroidal Alfven eigenmodes (TAEs) remained inconsistent with three fundamental characteristics of the observations: (i) the variation of the mode frequency with toroidal mode number, (ii) the chirping of the mode frequency for a given toroidal mode number, and (iii) the antiballooning density perturbation of the modes. It is now shown that these characteristics can be explained by observing that cylindrical-like modes can exist in the weak magnetic shear region of the plasma that then make a transition to TAEs as the central safety factor decreases in time. (20 refs.)
Nieto, M., D. N. Ruzic, et al. (2003). "Experimental measurements of helium retention in flowing liquid lithium." Fusion Science and Technology
15th Topical Meeting on the Technology of Fusion Energy 44(1): 232-6.
The Flowing Liquid Surface Retention Experiment (FLIRE) has been built and designed at the University of Illinois at Urbana-Champaign to provide fundamental experimental data on the retention and pumping of He, H and other species in flowing liquid surfaces. These measurements are critical to the development of advanced plasma-facing components (PFCs) that apply flowing liquid metals to mitigate high heat loads encountered in the divertor region of next-step fusion reactors. The FLIRE facility currently uses an ion beam source, which injects ions into a flowing stream of liquid lithium. Its design allows the liquid lithium to flow between two vacuum chambers that become isolated from each other when the lithium flows. Recent results show retention of helium inflowing liquid lithium at 250-300 degrees C to be of the order 10{sup -4} and diffusivities of 10{sup -4} to 10{sup -3} cm{sup 2}/sec. (14 refs.)
Okada, S., F. Kodera, et al. (2003). "Additional control experiments on field reversed configuration plasma." Fusion Science and Technology
4th International Conference on Open Magnetic Systems for Plasma Confinement 43(1T): 295-8.
Plasmas with field reversed configuration (FRC) are confined in open systems and have extremely high beta value of about 100% and they are one of candidates for an attractive reactor. But, in many cases they are produced in theta pinch apparatus and accessibility of additional heating facilities is poor. In order to solve this problem and to realize density appropriate for neutral beam injection, technology of translation is useful. By the translation, an FRC plasma is ejected out from theta pinch formation region and is translocated into a confinement region. With this translation, experiments related to sustain and control the FRC plasma become to be accomplished. Actually, axial magnetic compression, neutral beam heating and low frequency RF wave heating experiments are carried out on the FRC Injection Experiment (FIX) apparatus. (15 refs.)
Paul, S. F. (2003). "Real-time plasma rotation diagnostic for measuring small Doppler shifts." Review of Scientific Instruments
14th Topical Conference on High-Temperature Plasma Diagnostics 74(3): 2098-102.
A noninvasive velocity measurement has been developed for use in high-temperature plasma confinement devices, having particular use where multi-kilohertz measurement of plasma rotation is needed. The most interesting aspect of the technique is that the Doppler shift is determined from the ratio of the light intensity from two detectors rather than by resolving the emission line with a spectrometer. One detector views the plasma through an interference filter whose passband has a negative slope, and the other channel views the identical volume of plasma through a positive-slope filter. The signal ratio varies as the line is shifted across the passbands, but is not sensitive to changes in plasma emission. For interference filters with linear passbands, that is, constant slopes, the ratio is not sensitive to ion temperature, and the shifted wavelength reduces to a simple function of the signal ratio, the channels' relative responsivity, and the two filters' transmission curves. Filters with sharp passband slopes (100%-200% per nm) and approximately=0.1% deviation from linearity have been fabricated. For an uncertainty in the signal ratio of +or-1%, the expected error in the wavelength shift is +or-0.001 nm ( Delta v{sub phi } approximately=+or-1 km/s). (9 refs.)
Peterson, B. J., A. Y. Kostrioukov, et al. (2003). "Bolometer diagnostics for one- and two-dimensional measurements of radiated power on the Large Helical Device." Plasma Physics and Controlled Fusion 45(7): 1167-82.
Bolometer diagnostics are installed on the Large Helical Device (LHD) to provide measurements of the total (broad spectrum) electromagnetic radiation emitted by the plasma. Three types of detectors are used: resistive metal film bolometers, absolute extreme ultraviolet photodiodes and infrared imaging video bolometers. Details of the installation in LHD are given for each type of detector. The detector calibration, data analysis and tomography techniques are described and compared and sample results are shown. (34 refs.)
Peterson, B. J., A. Y. Kostrioukov, et al. (2003). "Calibration and sensitivity of the infrared imaging video bolometer." Review of Scientific Instruments
14th Topical Conference on High-Temperature Plasma Diagnostics 74(3): 2040-3.
The infrared (IR) imaging video bolometer (IRVB) is an imaging bolometer which uses a large (9 cm*9 cm) thin (1 mu m) gold foil and an IR camera to provide images of radiation from the plasma. Calibration of the IRVB using a lamp has been performed to compensate for any nonuniformities in the foil's thickness and its thermal properties due to blackening of the foil with graphite to improve the IR emissivity. This calibration revealed close to expected values for the calibration coefficient proportional to the product of the thermal conductivity and the foil thickness in the central region of the foil, while these values were anomalously high near the foil edge. The calibration coefficient proportional to the thermal diffusivity is a factor of 2 smaller than the expected value at the center and drops further at the edge of the foil. Using a derived expression for the IRVB noise equivalent power, a sensitivity comparison shows the IRVB using current IR technologies to be ~200 times less sensitive than an equivalent conventional resistive bolometer operating under ideal conditions. (14 refs.)
Peterson, B. J., K. Sato, et al. (2003). "Impurity radiation during 'breathing'-like oscillations in LHD discharges using a wall limiter." Journal of Nuclear Materials
15th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI-15) 313/316: 1178-82.
By changing the distribution of the coil currents in the helical coils of the Large Helical Device (LHD) the plasma could be scraped off on the stainless steel inboard wall at the last closed flux surface, eliminating the ergodic edge region and helical divertor (HD) plasmas. During such NBI-heated discharges slow oscillations (1-2.5 Hz) have been observed in the major global parameters. These oscillations exhibit many similarities to the 'breathing' relaxation phenomenon observed during long pulse experiments in LHD when the divertor material was stainless steel. These similarities include a large core radiation fraction, significant radiation from the metallic impurities and dependence of the oscillation frequency on density. In contrast, discharges using the graphite HD show hollow radiation profiles and reduced levels of radiation from heavy impurities. Modeling shows that differences in the density dependence of the oscillation frequency between the wall limiter and HD cases can be attributed to differences in the impurity diffusion between the core and ergodic edge regions. (17 refs.)
Petrov, Y. and T. S. Huang (2003). "Effect of a large-gradient radial electric field on plasma current in a spherical torus." Nuclear Fusion 43(2): 130-41.
A description of the guiding field line (GFL) model is presented for spherical torus geometry. The model is extended to cover small banana orbits that are trapped either above or below the midplane. Comparison of the model with the fluid and neoclassical theories demonstrates correctness of the model, though it shows that the model is not exactly equivalent to them. The GFL model provides a more general description for the orbit-squeezing effect resulting from the large gradient of the radial electric field. It shows that not only does the deformation of banana orbits, but also the change of the passing particles' orbits, affect the current in plasma. Besides, it appears that the change of the orbits' shape, as projected on the magnetic surface, contributes to modification of the current along with the radial squeezing/expanding. Distortion of the current by the gradient of the electric field is found to be predominantly in the direction parallel to the magnetic field, while almost no effect is seen for the perpendicular component. (12 refs.)
Pustovitov, V. D. (2003). "Beta-limiting MHD phenomena in toroidal systems." Plasma Physics Reports 29(10): 816-825.
Equilibrium effects, neoclassical tearing modes, and resistive wall modes are discussed as phenomena limiting attainable plasma pressure, with emphasis on the current progress in theoretical studies at the Kurchatov Institute. The review is based on the results presented at the 11th International Congress on Plasma Physics (Sydney, 2002). (C) 2003 MAIK.
Raman, R., H. W. Kugel, et al. (2003). "Fast neutral pressure measurements in NSTX." Review of Scientific Instruments
14th Topical Conference on High-Temperature Plasma Diagnostics 74(3): 1900-4.
Several fast neutral pressure gauges have been installed on the National Spherical Torus Experiment (NSTX) to measure the vessel and divertor pressure during inductive and coaxial helicity injected plasma operations. Modified, Princeton divertor experiment-type Penning gauges have been installed on the upper and lower divertors. Neutral pressure measurements during plasma operations from these and from two shielded fast microion gauges at different toroidal locations on the vessel midplane are described. A new unshielded ion gauge, referred to as the in-vessel neutral pressure gauge is under development. (6 refs.)
Reichle, R., D. Guilhem, et al. (2003). "Status of power balance monitoring for long pulse operation at Tore Supra." Nuclear Fusion
3rd International-Atomic-Energy-Agency Technical Meeting 43(9): 797-804.
A short review of diagnostics and techniques for active power exhaust control is given from the point of view of experiences gained at Tore Supra. The physical and technological aspects are presented with the aim of illustrating strengths, difficulties and potential for future long pulse and steady state fusion devices such as W7X and ITER-FEAT. One of the main features of the Tore Supra approach to power exhaust control is the reliance on infrared (IR) thermography to survey nearly 100% of the high flux target zones. This is applied to the largest elements, the toroidal pump limiter and the heating antennae, by a number of IR endoscopes. For high flux target areas inaccessible to direct view, optical fibres in the IR and the near-IR range are used. The large wall panels, which evacuate the rest of the power, are controlled by calorimetry. Bolometry is used to achieve a complete power balance. These measurements can be inserted in control loops acting typically on the injected heating power or the radiation cooling. (25 refs.)
Ricci, P., G. Lapenta, et al. (2003). "Electron acceleration and heating in collisionless magnetic reconnection." Physics of Plasmas 10(9): 3554-60.
Electron acceleration and heating during collisionless magnetic reconnection is discussed by using the results of implicit kinetic simulations of Harris current sheets. Simulations up to the physical mass ratio are performed to study and compare electron dynamics in plasmas with different beta values. The attention is focused on the typical trajectory of electrons passing through the reconnection region, on the electron velocity, in particular on the out-of-plane velocity, and on the electron heating along the in-plane and out-of-plane directions. (57 refs.)
Roh, Y., C. W. Domier, et al. (2003). "Ultrashort pulse reflectometry for density profile and fluctuation measurements on SSPX." Review of Scientific Instruments
14th Topical Conference on High-Temperature Plasma Diagnostics 74(3): 1518-21.
A broadband, multichannel ultrashort pulse reflectometry (USPR) diagnostic on the Sustained Spheromak Physics eXperiment device has recently undergone a number of system upgrades, which has resulted in significant improvements in the signal-to-noise ratio of the USPR signals and a dramatic reduction in the number of "lost signals" in which the amplitude of the reflected wave form drops below detection threshold. This has greatly enhanced the ability of USPR to study relatively fast density profile modifications, and allows the simultaneous monitoring of multiple density layers with as short as a 3 mu s pulse repetition rate. This article provides details of the upgraded USPR system together with density profiles and fluctuation data. (9 refs.)
Ryutov, D. D., D. C. Barnes, et al. (2003). "Particle and heat transport in a dense wall-confined MTF plasma (theory and simulations)." Nuclear Fusion
3rd International-Atomic-Energy-Agency Technical Meeting 43(9): 955-60.
Plasma beta in magnetized target fusion systems is sometimes much greater than 1, and the plasma may be in direct contact with the imploding liner. Plasma processes are strongly dominated by inter-particle collisions. Under such conditions, the plasma microturbulence, behaviour of alpha -particles, and plasma equilibria are very different from conventional fusion systems. This paper contains the most comprehensive analysis of the corresponding phenomena to date. Two-dimensional numerical simulations of plasma convection in the targets of a diffuse pinch type demonstrate an onset of convection in this configuration. (40 refs.)
Ryutov, D. D., R. H. Cohen, et al. (2003). "A phenomenological model of the current filamentation instability driven by cathode processes in the Livermore spheromk." Plasma Physics Reports 29(7): 605-617.
The current density on the open field lines of the Livermore spheromak (SSPX) typically exceeds the saturation current density of the bulk plasma. We assume that the mechanism that provides conditions for that is associated with the formation of a thin layer near the cathode surface, where both the plasma and the neutral density are higher than in the bulk plasma and where intense ionization occurs. The ions formed in this layer fall back onto the cathode, whereas electrons contribute to the high current density in the bulk plasma. The particle balance in the ionizing layer is determined by the recycling coefficient, which, in turn, depends on the cathode temperature and the sheath voltage. As it turns out, these dependences give rise to an instability that leads to the current filamentation and the formation of hot spots on the cathode surface. The instability can be characterized in a phenomenological manner without going into the details of the structure of the ionizing layer, whose effect on the instability shows up in the form of a couple of numerical coefficients of the order of one. We predict the characteristic size and the shape of the filaments (and the hot spots), which are in a general agreement with discoloration patterns on the surface of the cathode in the SSPX. If the magnetic field is tilted to the surface, the footpoints of the filaments move with a significant velocity, whose direction depends on the ratio of the ion gyroradius and the thickness of the ionizing layer. This instability, although primarily considered in conjunction with the SSPX experiment, may play a role in spherical tokamaks and other systems with coaxial helicity injection. (C) 2003 MAIK "Nauka/Interperiodica".
Ryzhkov, S. V., V. I. Khvesyuk, et al. (2003). "Progress in an alternate confinement system called a FRC." Fusion Science and Technology
4th International Conference on Open Magnetic Systems for Plasma Confinement 43(1T): 304-8.
The high fusion power density resulting from high beta (the ratio of the plasma to magnetic energy density) and natural divertor make the field-reversed configuration (FRC) a prime candidate for fusion reactor other than tokamak, the so-called alternate concept. Brief review of the simple compact system with natural advantages and reactor potential is given. Theoretical and experimental results over the last seven years are discussed. (45 refs.)
Samm, U. (2003). "Controlled thermonuclear fusion at the beginning of a new era." Contemporary Physics 44(3): 203-17.
Fusion research has come to a stage at which the newly designed International Thermonuclear Experimental Reactor (ITER) can be build-a machine which will demonstrate the feasibility of generating a significant surplus of energy with a thermonuclear plasma. This machine shall provide the last step towards the first fusion power plant, the construction of which could start in about two decades and would open to mankind the elements deuterium and lithium as a new safe and clean primary energy source. This paper describes the success story of the tokamak-concept of magnetic confinement on which ITER is based. But it also discusses the ongoing research which concentrates on the most critical problems to be solved for the power plant beyond ITER for which a continuously and efficiently operating fusion reactor is required. (38 refs.)
Schuller, F. C. (2003). "Conference summary: experiments in confinement and wave-plasma interaction." Nuclear Fusion 43(12): 1720-39.
This paper summarizes the results presented at the IAEA Fusion Energy Conference 2002 with respect to the performance and confinement of tokamaks, spherical tokamaks, stellarators, reversed field pinches and mirrors. Transport, internal transport barriers, methods to control pressure and current density profiles by auxiliary heating and non-inductive current drive is discussed and compared to the predictions for ITER. Based on the results presented at this conference it can be concluded that one has confidence that the ITER performance specifications will be achieved in the future. Marked progress on alternative confinement concepts can be noted as well. (52 refs.)
Sen, A. K., M. Nagashima, et al. (2003). "Optimal control of tokamak resistive wall modes in the presence of noise." Physics of Plasmas 10(11): 4350-7.
A scheme of optimal control based on "state" feedback for the resistive wall modes in tokamaks is described. The important effects of ever present magnetohydrodynamic noises, which have been neglected in the past, are included in the stochastic formulation of the basic equations. The optimality of the system is obtained both in terms of minimization of the stabilized fluctuation levels of the modes and the control energy. It is found that the stabilized fluctuation level and control signal level are of the order of the noise level of the system in the steady state. (14 refs.)
Soukhanovskii, V. A., A. L. Roquemore, et al. (2003). "High-resolution spectroscopic diagnostic for divertor and scrape-off layer neutral and impurity emission measurements in the national spherical torus experiment." Review of Scientific Instruments
14th Topical Conference on High-Temperature Plasma Diagnostics 74(3): 2094-7.
The National Spherical Torus Experiment boundary physics program presently focuses on edge power and particle flow optimization and control in beta >or=25% long pulse L- and H-mode plasmas with high harmonic fast wave heating power up to 6 MW and neutral beam injection power up to 5 MW, with the emphasis on the edge physics implications resulting from the low aspect ratio geometry. To address the particle flux measurements in the divertor and plasma scrape-off layer (SOL), two spectrally filtered one-dimensional charge coupled device cameras have been fielded. The cameras utilize 2048 pixel 12-bit Dalsa CL-C6 arrays, f=85 mm lenses, and Delta lambda =1.5 nm bandpass interference filters. Both cameras provide mm spatial resolution, sub-ms temporal resolution and are spatially and photometrically calibrated. Midplane SOL and divertor brightness profiles of C III and deuterium species have been obtained in the L- and H-mode phases of center stack limited and diverted plasmas. The equilibria reconstructed by the EFIT code are found in agreement with the optical and infra-red measurements. In-out asymmetries in divertor recycling and carbon fluxes have been observed in L- and H-mode plasmas. The analysis of neutral recycling and impurity fluxes using the two-dimensional multifluid code UEDGE is in progress. (11 refs.)
Stotler, D. P., B. LaBombard, et al. (2003). "Neutral transport simulations of gas puff imaging experiments." Journal of Nuclear Materials
15th International Conference on Plasma-Surface Interactions in Controlled Fusion Devices (PSI-15) 313/316: 1066-70.
Visible imaging of gas puffs has been used on the Alcator C-Mod tokamak to characterize edge plasma turbulence, yielding data that can be compared with plasma turbulence codes. Simulations of these experiments with the DEGAS 2 Monte Carlo neutral transport code have been carried out to explore the relationship between the plasma fluctuations and the observed light emission. By imposing two-dimensional modulations on the measured time-average plasma density and temperature profiles, we demonstrate that the spatial structure of the emission cloud reflects that of the underlying turbulence. However, the photon emission rate depends on the plasma density and temperature in a complicated way, and no simple scheme for inferring the plasma parameters directly from the light emission patterns is apparent. The simulations indicate that excited atoms generated by molecular dissociation are a significant source of photons, further complicating interpretation of the gas puff imaging results. (12 refs.)
Stutman, D., M. Finkenthal, et al. (2003). "Impurity transport measurements in beam heated low-confinement mode discharges in the national spherical torus experiment." Physics of Plasmas 10(11): 4387-95.
Impurity injection experiments were performed in the National Spherical Torus Experiment [NSTX, M. Ono et al., Nucl. Fusion 40, 557 (2000)] for a first assessment of low-Z impurity transport in a low field, low-aspect ratio device. Short neon puffs were injected in beam heated, co-rotating L-mode (low confinement) discharges and the radial penetration of the successive neon charge states has been monitored with arrays of filtered ultrasoft x-ray diodes. Time-dependent modeling of the neon emissivity in several spectral bands indicates a core diffusion coefficient in the neoclassical range (<1 m{sup 2}/s) in these strongly rotating plasmas, consistent with the low thermal ion transport inferred from the power balance analysis. By contrast, due to the large Larmor radii in NSTX turbulent diffusion would reach tens of m{sup 2}/s if tokamak-like instability growth rates were assumed. The much lower experimental diffusivity then suggests that turbulent ion transport must be largely suppressed in the NSTX core. It is not yet clear in what measure this is due to the large E*B flow shear characteristic of beam heated NSTX discharges, or to the long wavelength instability growth rates being intrinsically low in the NSTX core, as recently predicted. (40 refs.)
Stutman, D., M. Finkenthal, et al. (2003). "Integrated impurity diagnostic package for magnetic fusion experiments." Review of Scientific Instruments
14th Topical Conference on High-Temperature Plasma Diagnostics 74(3): 1982-7.
We develop an integrated instrumental and computational package for the diagnosis of impurity content, Z{sub eff} profile, particle transport, and magnetohydrodynamics activity in magnetic fusion experiment plasmas. The package includes broadband filtered arrays of absolute (AXUV) photodiodes, a transmission grating imaging spectrometer measuring up to 20 chords across the discharge and having a few angstrom spectral resolution, together with an atomic physics database coupled with an impurity transport code. The atomic physics database is based on ab initio computations with the Hebrew University Lawrence Livermore atomic code. The package is designed for the diagnostic of sub-keV plasmas having predominantly low-Z impurities (C, B, and O) together with trace metals. A preliminary version is being tested on the National Spherical Torus Experiment spherical torus at the Princeton Plasma Physics Laboratory, using the ultrasoft x-ray imaging system and a grazing incidence spectrometer. Representative results from non-H-mode discharges are presented. (16 refs.)
Sudo, S., Y. Nagayama, et al. (2003). "Recent diagnostic developments on LHD." Plasma Physics and Controlled Fusion
30th EPS Conference on Controlled Fusion and Plasma Physics 45: A425-A443.
Standard diagnostics for fundamental plasma parameters and for plasma physics are routinely utilized for daily operation and physics studies in the large helical device (LHD) with high reliability. Diagnostics for steady-state plasma are under intensive development, especially for T-e, n(e) (yttrium-aluminium garnet (YAG) laser Thomson, CO2 laser polarimeter), data acquisition in steady-state and heat-resistant probes. To clarify the plasma properties of the helical structure, two- or three-dimensional diagnostics are being aggressively developed: tangential cameras (fast SX TV, photon counting CCD, H-alpha TV); tomography (tangential SX CCD, bolometer); imaging (bolometer, ECE, reflectometer). Divertor and edge physics are important key issues for steady-state operation. Diagnostics for neutral flux (H-alpha array, Zeeman spectroscopy) and n(e) (fast scanning probe, Li beam probe, pulsed radar reflectometer) are also in advanced stages of development. In addition to these, advanced diagnostics are being intensively developed in LHD through domestic and international collaborations.
Suzuki, T., A. Ito, et al. (2003). "Statistical model of current filaments in a turbulent plasma." Fluid Dynamics Research 32(6): 247-60.
Local currents in a turbulent toroidal discharge plasma fluctuate intermittently, suggesting that the current density has a strongly inhomogeneous distribution. A canonical statistical model of current filaments accounts well for the spectral structure of the fluctuations. An innovative aspect of the model is that the ensemble is defined by the total current, not by the energy. Filamentation of the current density is a possible mechanism of producing resistance anomaly in a turbulent plasma. (20 refs.)
Swain, D. W., M. D. Carter, et al. (2003). "Loading and asymmetry measurements and modeling for the National Spherical Torus Experiment ion cyclotron range of frequencies system." Fusion Science and Technology 43(4): 503-13.
The ion cyclotron heating and current drive system on the National Spherical Torus Experiment (NSTX) has delivered over 3 MW reliably for pulse lengths over 100 ms with various phasings of the antennas. A circuit model of the system that includes the 12 coupled antennas and six radio-frequency sources has been developed that gives good agreement with vacuum measurements. When it is used to experimentally determine the S-matrix of the system under different plasma conditions, pronounced asymmetries in the off-diagonal values of the S-matrix are seen. The S-matrix in the presence of plasma has been calculated with the RANT3D code using measured edge density profiles in front of the antenna; these agree remarkably well with the measurements. The asymmetry is caused primarily by the large pitch angle of the magnetic field in front of the antenna, coupled with the gradients in the plasma edge. (14 refs.)
Synakowski, E. J., M. G. Bell, et al. (2003). "The National Spherical Torus Experiment (NSTX) research programme and progress towards high beta, long pulse operating scenarios." Nuclear Fusion
Fusion Energy Conference (FEC-2002) 43(12): 1653-64.
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 the 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 beta {sub T} identical to <p>/(B{sub T0}{sup 2}/2{sub mu 0}) 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 fueling 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 comparison to the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emission 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{sup -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 start-up studies have focused on coaxial exactly injection. With this technique, trial currents up to 400 kA have been driven, and studies to assess flux closure and coupling to the other current drive techniques have begun. (47 refs.)
Taccetti, J. M., T. P. Intrator, et al. (2003). "FRX-L: a field-reversed configuration plasma injector for magnetized target fusion." Review of Scientific Instruments 74(10): 4314-23.
We describe the experiment and technology leading to a target plasma for the magnetized target fusion research effort, an approach to fusion wherein a plasma with embedded magnetic fields is formed and subsequently adiabatically compressed to fusion conditions. The target plasmas under consideration, field-reversed configurations (FRCs), have the required closed-field-line topology and are translatable and compressible. Our goal is to form high-density (10{sup 17} cm{sup -3}) FRCs on the field-reversed experiment-liner (FRX-L) device, inside a 36 cm long, 6.2 cm radius theta coil, with 5 T peak magnetic field and an azimuthal electric field as high as 1 kV/cm. FRCs have been formed with an equilibrium density n{sub e} approximately=(1 to 2)*10{sup 16} cm{sup -3}, T{sub e}+T{sub i} approximately=250 eV, and excluded flux approximately=2 to 3 mWb. (43 refs.)
Taylor, G., P. C. Efthimion, et al. (2003). "Enhanced conversion of thermal electron Bernstein waves to the extraordinary electromagnetic mode on the National Spherical Torus Experiment." Physics of Plasmas 10(5): 1395-401.
A fourfold increase in the conversion of thermal electron Bernstein waves (EBW) to the extraordinary mode (X mode) was measured when the density scale length (L{sub n}) was progressively shortened by a local boron nitride limiter in the scrape-off of an Ohmically heated National Spherical Torus Experiment plasma [M. Ono, S. Kaye, M. Peng et al., Proceedings of the 17th IAEA Fusion Energy Conference (IAEA, Vienna, 1999), vol. 3, p. 1135]. The maximum conversion efficiency approached 50% when L{sub n} was reduced to 0.7 cm, in agreement with theoretical predictions that used locally measured L{sub n}. Calculations indicate that it is possible to establish L{sub n}<0.3 cm with a local limiter, a value predicted to attain ~100% EBW conversion to the X mode in support of proposed EBW heating and current drive scenarios. (21 refs.)
Terry, J. L., S. J. Zweben, et al. (2003). "Observations of the turbulence in the scrape-off-layer of Alcator C-mod and comparisons with simulation." Physics of Plasmas
44th Annual Meeting of the Division of Plasma of the American-Physical-Society 10(5): 1739-47.
The intermittent turbulent transport in the scrape-off-layer (SOL) of Alcator C-Mod [I.H. Hutchinson, R. Boivin, P.T. Bonoli et al., Nucl. Fusion 41, 1391 (2001)] is studied experimentally by imaging with a very high density of spatial measurements. The two-dimensional structure and dynamics of emission from a localized gas puff are observed, and intermittent features (also sometimes called "filaments" or "blobs") are typically seen. The characteristics of the spatial structure of the turbulence and their relationship to the time-averaged SOL profiles are discussed and compared with those measured on the National Spherical Torus Experiment [M. Ono, S. M. Kaye, Y.-K. M. Pong et al., Nucl. Fusion 40, 557 (2000)]. The experimental observations are compared also with three-dimensional nonlinear numerical simulations of edge turbulence. Radial profiles of the poloidal wave number spectra and the poloidal scale length from the simulations are in reasonable agreement with those obtained from the experimental images, once the response of the optical system is accounted for. The resistive ballooning mode is the dominant linear instability in the simulations. The ballooning character of the turbulence is also consistent with fluctuation measurements made at the inboard and outboard midplane, where normalized fluctuation levels are found to be about 10 times smaller on the inboard side. For discharges near the density limit, turbulent structures are seen on closed flux surfaces. (36 refs.)
Waller, V., B. Pegourie, et al. (2003). "Investigation of current-density modification during magnetic reconnection by analysis of hydrogen-pellet deflection." Physical Review Letters 91(20): 205002-4.
A pellet penetrating the inner region of a tokamak discharge, where the safety factor drops below unity, triggers an instability analogous to a sawtooth crash. Because of the simultaneity of the crash and pellet crossing, the latter is an appropriate probe for investigating the current distribution during reconnection. In this Letter, pellet deflection is used to characterize the associated electron distribution function. The perturbation compatible with the observed trajectory requires a negative current layer on the q=1 magnetic surface between 3 and 12 times the equilibrium current density and an expulsion of high energy electrons from the plasma core. (18 refs.)
Wang, Z. H. and G. A. Wurden (2003). "Hypervelocity dust beam injection for internal magnetic field mapping." Review of Scientific Instruments
14th Topical Conference on High-Temperature Plasma Diagnostics 74(3): 1887-1891.
Injecting neutral atoms into high-temperature plasmas forms the basis for several important diagnostics, such as motional Stark effect and charge exchange recombination spectroscopy. We describe an alternative approach to seeding the plasma with neutrals, via "hypervelocity dust beam injection" (HDBI), using micron-sized dusts. Among its many potential applications, HDBI mapping of two-dimensional internal magnetic fields inside medium-sized (50-500 eV) plasmas is discussed in detail. Electrostatic acceleration at similar to100-200 kV will launch a stream of (0.2-10 Am-sized) dust grains of lithium or carbon to hypervelocities (1-10 km/s). Each dust grain, acting. as a "microcomet" in the plasma, forming a plume (tail), which if photographed, will reveal the direction of the local magnetic field, with anywhere from 10-100 microcomets in the plasma at any time, a full profile of the B-field direction could be obtained per high resolution image. Due to the small dust grain size, the perturbation to the plasma will be minimal. HDBI could be a simple low cost approach to obtain internal magnetic field information in plasmas with magnetic field structures that are significantly different than vacuum fields, such as in spherical tokamaks, FRC's, RFP's, and spheromaks. (C) 2003 American Institute of Physics.
Wang, Z. H. and J. L. Kline (2003). "Electrostatic method to accelerate nanoshells to extreme hypervelocity." Applied Physics Letters 83(8): 1662-1664.
Using an acceleration voltage of less than a few hundred kilovolts, it is unlikely that a charged solid object larger than a few micron (10(-6) m) in all three dimensions can be accelerated to more than 10 km/s. Quasi-two-dimensional (Q2D) objects are unique forms of matter with two macroscopic dimensions, while the third approaches atomic dimensions. Well-known examples of Q2D objects are thin films. Another example of a Q2D object will be a sphere with a nm thick shell (nanoshell). In this letter, it is predicted that nanoshells can be accelerated to 100 km/s (extreme hypervelocity, or EHV) and above using the electrostatic method. The maximum velocity is limited by field emission and material strength. The two limits only allow a certain number of charges on a nanoshell before it starts to emit ions or electrons, or to break. "Table-top" EHV nanoshell beams can be used for high-temperature plasma diagnostics and fueling. EHV nanoshells can also be used to study hypervelocity-impact phenomena in a momentum space not accessible in the past. (C) 2003 American Institute of Physics.
Wolf, R. C. (2003). "Internal transport barriers in tokamak plasmas." Plasma Physics and Controlled Fusion 45(1): R1-91.
Internal transport barriers in tokamak plasmas are explored in order to improve confinement and stability beyond the reference scenario, used for the ITER extrapolation, and to achieve higher bootstrap current fractions as an essential part of non-inductive current drive. Internal transport barriers are produced by modifications of the current profile using external heating and current drive effects, often combined with partial freezing of the initial skin current profile. Thus, formerly inaccessible ion temperatures and Q{sub DT}{sup eq} values have been (transiently) achieved. The present paper reviews the state of the art of these techniques and their effects on plasma transport in view of optimizing the confinement properties. Implications and limits for possible steady state operations and extrapolation to burning plasmas are discussed. (368 refs.)
Woodruff, S., D. N. Hill, et al. (2003). "New mode of operating a magnetized coaxial plasma gun for injecting magnetic helicity into a spheromak." Physical Review Letters 90(9): 095001-4.
By operating a magnetized coaxial plasma gun continuously with just sufficient current to enable plasma ejection, large gun-voltage spikes (~1 kV) are produced, giving the highest sustained voltage ~500 V and highest sustained helicity injection rate observed in the Sustained Spheromak Physics Experiment. The spheromak magnetic field increases monotonically with time, exhibiting the lowest fluctuation levels observed during formation of any spheromak (B/B >or= 2%). The results suggest an important mechanism for field generation by helicity injection, namely, the merging of helicity-carrying filaments. (23 refs.)
Xu, G. S., B. N. Wan, et al. (2003). "Bursty events and incremental diffusion in a local diffusion and multi-scale convection system." Chinese Physics 12(2): 189-197.
A one-dimensional cellular automaton is defined without the critical gradient rule (Deltah > Deltah(c)) which is essential to the existence of avalanches in self-organized criticality (SOC) models. Instead, only the local diffusion rule is used, however, the characteristics of SOC, such as the bursty behaviour, power-law decay in fluctuation spectra, self-similarity over a broad range of scales and long-time correlations, are still observed in these numerical experiments. This numerical model is established to suggest that the bursty events and the incremental diffusion observed universally in fusion experiments do not necessarily imply the submarginal dynamics.
Yagi, Y., Y. Maejima, et al. (2003). "Characteristics of global confinement properties in TPE-series reversed-field pinch devices." Nuclear Fusion 43(12): 1787-800.
A confinement database of the reversed-field pinch (RFP) devices of the toroidal pinch experiment (TPE) series has been established. The TPE database contains global confinement properties under various experimental conditions for four different devices and seven different cases, which cover two decades of experiments. Characteristics of the TPE database have been extracted under consistent experimental conditions of conventional RFP discharges, and scaling laws of the poloidal beta, beta {sub p} and energy confinement time, tau {sub E}, are given against independent parameters that include the size of the plasma and the pinch parameter Theta . Scaling laws of beta {sub p} ~ a{sup -0.23}I{sub p}{sup -0.05}(I/N){sup -0.56} Theta {sup 1.47} and tau {sub E} ~ a{sup 1.63} I{sub p}{sup 0.78}(I/N){sup 0.33} Theta {sup 2.97} are obtained for a set of scans, where a is the plasma minor radius, I{sub p} the plasma current and N the column density. The scaling of tau {sub E} on a, I{sub p}, and I/N is similar to that dominated by the transport under tearing instability. Another scaling shows that tau {sub E} is inversely proportional to the Joule input power. The scaling for the apparent effective charge shows a positive correlation both with I{sub p} and I/N. (46 refs.)
Zhang, Q. M., Q. S. Wang, et al. (2003). "Detection of fast neutrons using a new scintillating-fiber-array neutron detector." Nuclear Instruments & Methods in Physics Research Section A-Accelerators Spectrometers Detectors and Associated Equipment 496(1): 228-232.
A new scintillating-fiber-array neutron detector was used in the fast neutron detection and spectroscopy. Two pulse height spectra for 2.5- and 14-MeV neutrons were obtained and compared. Differences of the pulse height spectra for 14 MeV neutrons at different applied voltages were measured and analyzed as well. Other two feasible methods are proposed in the calibration of the absolute sensitivity for the detector. The experimental results show that compared to conventional scintillating fiber detectors, this detector exhibits two distinct properties: (1) high neutron-to-gamma sensitivity ratio, (2) apparent peak in the pulse height spectrum for 14 MeV neutrons. (C) 2002 Elsevier Science B.V. All rights reserved.
Zhang, Q. M., Q. S. Wang, et al. (2003). "Study on the energy response to neutrons for a new scintillating-fiber-array neutron detector." Nuclear Instruments & Methods in Physics Research Section A-Accelerators Spectrometers Detectors and Associated Equipment 496(1): 146-153.
The energy response of a new scintillating-fiber-array neutron detector to neutrons in the energy range 0.01 MeV less than or equal to E-n less than or equal to 14 MeV was modeled by combining a simplified Monte Carlo model and the MCNP 4b code. In order to test the model and get the absolute sensitivity of the detector to neutrons., one experiment was carried out for 2.5 and 14MeV neutrons from T(p,n)(3) He and T(d,n)(4) He reactions at the Neutron Generator Laboratory at the Institute of Modern Physics, the Chinese Academy of Science. The absolute neutron fluence was obtained with a relative standard uncertainty 4.5% or 2.0% by monitoring the associated protons or He-4 particles, respectively. Another experiment was carried out for 0.5, 1.0 1.5, 2.0, 2.5 MeV neutrons from T(p,n)He-3 reaction, and for 3.28, 3.50, 4.83, 5.74MeV neutrons from D(d,n)He-3 reaction on the Model 5SDH-2 accelerator at China Institute of Atomic Energy. The absolute neutron fluence was obtained with a relative standard uncertainty 5.0% by using a calibrated long counter. The results show that the calculated absolute sensitivity from the model agrees with the experimental data considering the expanded relative uncertainty 15% in the experiments. (C) 2002 Elsevier Science B.V. All rights reserved.
Gray, D. S., S. C. Luckhardt, et al. (2004). "Time resolved radiated power during tokamak disruptions and spectral averaging of AXUV photodiode response in DIII-D." Review of Scientific Instruments 75(2): 376-81.
Silicon absolute extreme ultraviolet (AXUV) photodiodes have been employed in a disruption radiometer diagnostic for measurement of radiant power in the DIII-D tokamak with a 170 kHz bandwidth. This is motivated by a need to improve the understanding of radiative processes in tokamak disruptions. The diagnostic described in this article has a single line of sight though the central plasma. Accounting for the photon energy dependence of the AXUV photodiode responsivity is made possible by optical filtering, with the aid of spectra from an extreme ultraviolet survey spectrometer. The appropriate effective responsivity for interpretation of the data is lower than the nominal value typically used for the detector. In the current quench phase of disruptions, it is less than half the nominal value. Comparisons with results from a foil bolometer find good agreement. (16 refs.)
Malmberg, J. A., J. Brzozowski, et al. (2004). "Mode- and plasma rotation in a resistive shell reversed-field pinch." Physics of Plasmas 11(2): 647-58.
Mode rotation studies in a resistive shell reversed-field pinch, EXTRAP T2R [P. R. Brunsell et al., Plasma Phys. Control. Fusion 43, 1 (2001)] are presented. The phase relations and nonlinear coupling of the resonant modes are characterized and compared with that expected from modeling based on the hypothesis that mode dynamics can be described by a quasi stationary force balance including electromagnetic and viscous forces. Both m=0 and m=1 resonant modes are studied. The m=1 modes have rotation velocities corresponding to the plasma flow velocity (20-60 km/s) in the core region. The rotation velocity decreases towards the end of the discharge, although the plasma flow velocity does not decrease. A rotating phase locked m=1 structure is observed with a velocity of about 60 km/s. The m=0 modes accelerate throughout the discharges and reach velocities as high as 150-250 km/s. The observed m=0 phase locking is consistent with theory for certain conditions, but there are several conditions when the dynamics are not described. This is not unexpected because the assumption of quasi stationarity for the mode spectra is not fulfilled for many conditions. Localized m=0 perturbations are formed in correlation with highly transient discrete dynamo events. These perturbations form at the location of the m=1 phase locked structure, but rotate with a different velocity as they spread out in the toroidal direction. (32 refs.)
Menard, J. E., M. G. Bell, et al. (2004). "Aspect ratio scaling of ideal no-wall stability limits in high bootstrap fraction tokamak plasmas." Physics of Plasmas 11(2): 639-46.
Recent experiments in the low aspect ratio National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 (2000)] have achieved normalized beta values twice the conventional tokamak limit at low internal inductance and with significant bootstrap current. These experimental results have motivated a computational re-examination of the plasma aspect ratio dependence of ideal no-wall magnetohydrodynamic stability limits. These calculations find that the profile-optimized no-wall stability limit in high bootstrap fraction regimes is well described by a nearly aspect ratio invariant normalized beta parameter utilizing the total magnetic field energy density inside the plasma. However, the scaling of normalized beta with internal inductance is found to be strongly aspect ratio dependent at sufficiently low aspect ratio. These calculations and detailed stability analyses of experimental equilibria indicate that the nonrotating plasma no-wall stability limit has been exceeded by as much as 30% in NSTX in a high bootstrap fraction regime. (53 refs.)
Milroy, R. D. and K. E. Miller (2004). "Edge-driven rotating magnetic field current drive of field-reversed configurations." Physics of Plasmas 11(2): 633-8.
Field-reversed configurations (FRCs) are created and sustained using a rotating magnetic field (RMF) in the Translation Confinement and Sustainment experiment. Normally this experiment is operated in a manner where the RMF only partially penetrates the plasma column. This method of operation may have significant advantages in producing less disturbances to the bulk of the FRC, but requires driving an overall radially inward flow to maintain E{sub theta }(r)=0 everywhere (through the V{sub r}B{sub z} term in the generalized Ohm's law). However, some RMF penetration is still required at the field null R, where B{sub z}=0. For some experimental conditions it appears that the RMF does not even penetrate as far as the null, raising the question as to how E{sub theta }(r=R) can be maintained at zero despite a finite eta {sub perpendicular to }j{sub theta }(r=R). Numerical simulations with a resistivity profile that is sharply peaked near the plasma edge yield similar profiles, and provide insight into this physical process. An inner magnetic structure forms, which rotates at a much lower frequency than the RMF. A tearing and reconnection process produces a torque transfer from the outer RMF to the inner structure, allowing it to act as an RMF downshifted to a lower frequency, and thus provide current drive to the inner region of the FRC. This mode of RMF current drive is being called "edge-driven mode". (9 refs.)
Ongena, J. and A. M. Messiaen (2004). "Heating, confinement and extrapolation to reactors." Fusion Science and Technology
6th Carolus Magnus Euro-Summer School on Plasma and Fusion Energy Physics 45(2T): 453-466.
Ono, M., M. Peng, et al. (2004). "Next-step spherical torus experiment and spherical torus strategy in the course of development of fusion energy." Nuclear Fusion 44(3): 452-463.
A spherical torus (ST) fusion energy development path which is complementary to the proposed tokamak burning plasma experiments such as ITER is described. The ST strategy focuses on a compact component test facility (CTF) and high performance advanced regimes leading to more attractive Demo and power plant scale reactors. To provide the physical basis for the CTF an intermediate step needs to be taken, which we refer to as the 'next-step spherical torus' (NSST) device and which we examine in some detail herein. NSST is a 'performance extension' stage ST with a plasma current of 5-10 MA, R = 1.5 m, B-T less than or equal to 2.6 T and the possibility of varying physical parameters. The mission of NSST is to (1) provide a sufficient physical basis for the design of a CTF; (2) explore advanced operating scenarios with high bootstrap current fraction and high performance which can be utilized by CTF, Demo, and power plants; and (3) contribute to the general science of high beta toroidal plasmas. The NSST is designed to utilize a TFTR-like site to minimize the cost and time required for design and construction.
Zweben, S. J., R. J. Maqueda, et al. (2004). "High-speed imaging of edge turbulence in NSTX." Nuclear Fusion 44(1): 134-53.
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, Land 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 approximately=10{sup 5} cm s{sup -1}, and sometimes show spatially periodic features. (62 refs.)