This book provides a systematic introduction to the physics behind measurements on plasmas. It develops from first principles the concepts needed to plan, execute, and interpret plasma diagnostics. The book is therefore accessible to graduate students and professionals with little specific plasma physics background, but is also a valuable reference for seasoned plasma physicists. Most of the examples are taken from laboratory plasma research, but the focus on principles makes the treatment useful to all experimental and theoretical plasma physicists, including those interested in space and astrophysical applications. This second edition is thoroughly revised and updated, with new sections and chapters covering recent developments in the field. Specific areas of added coverage include neutral-beam-based diagnostics, flow measurement with mach probes, equilibrium of strongly shaped plasmas and fusion product diagnostics.

A series of experiments, examining the confinement properties of ICRF heated H-mode plasmas, has been carried out on the C-Mod tokamak. C-Mod is a compact tokamak which operates at high particle, power, and current densities at toroidal fields up to 8T. Under these conditions the plasma is essentially thermal with very little contribution to the stored energy from energetic ions (typically no more than 5%) and with Ti~Te. Most of the data were taken with the machine in a single-null "closed" divertor configuration with the plasma facing components clad in molybdenum tiles. The data include those taken both before and after the first wall surfaces were coated with boron, with emphasis on the latter. H-modes obtained from plasmas run on boronized walls typically had lower impurity content and radiated power and attained higher stored energy than those run on bare molybdenum. Confinement enhancement, the energy confinement time normalized to L-mode scaling, for discharges with boronized walls, ranged from 1.6 to 2.4. The unique operating regime of the C-Mod device provided a means for extending the 1 tests of global scaling laws to parameter ranges not previously accessible. For example, the C-Mod ELMfree data was found to be 1.1-1.6 times the ITERH93 scaling and the ELMy data almost 2.0-2.8 times the ITERH92 ELMy scaling law, suggesting that the size scaling in both scalings may be too strong. While both ELMfree and ELMy discharges were produced, the ELM characteristics were not easily compared to observations on other devices. No large, low frequency ELMs were seen despite the very high edge pressure and temperature gradients that were attained. For all of our H-mode discharges, a clear linear relationship between the edge temperature pedestal and the temperature gradient in the core plasma was observed; the discharges with the "best" transport barriers also showing the greatest improvement in core confinement.

High-resolution charge-exchange recombination spectroscopic measurements of B5+ ions have enabled the first spatially resolved calculations of the radial electric field (Er) in the Alcator C-Mod pedestal region [E. S. Marmar, Fusion Sci. Technol. 51, 261 (2006)]. These observations offer new challenges for theory and simulation and provide for important comparisons with other devices. Qualitatively, the field structure observed on C-Mod is similar to that on other tokamaks. However, the narrow high-confinement mode (H-mode) Er well widths (5 mm) observed on C-Mod suggest a scaling with machine size, while the observed depths (up to 300 kV/m) are unprecedented. Due to the strong ion-electron thermal coupling in the C-Mod pedestal, it is possible to infer information about the main ion population in this region. The results indicate that in H-mode the main ion pressure gradient is the dominant contributor to the Er well and that the main ions have significant edge flow. C-Mod H-mode data show a clear correlation between deeper Er wells, higher confinement plasmas, and higher electron temperature pedestal heights. However, improved L-mode (I-mode) plasmas exhibit energy confinement equivalent to that observed in similar H-mode discharges, but with significantly shallower Er wells. I-mode plasmas are characterized by H-mode-like energy barriers, but with L-mode-like particle barriers. The decoupling of energy and particle barrier formation makes the I-mode an interesting regime for fusion research and provides for a low collisionality pedestal without edge localized modes.

A set of external coils (A-coils) capable of producing nonaxisymmetric, predominantly n=1, fields with different toroidal phase and a range of poloidal mode m spectra has been used to determine the threshold amplitude for mode locking over a range of plasma parameters in Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda, P. Bonoli, S. Fairfax, C. Fiore, J. Goetz, S. Golovato, R. Granetz, M. Greenwald et al., Phys. Plasmas 1, 1511 (1994)]. The threshold perturbations and parametric scalings, expressed in terms of (B21∕BT), are similar to those observed on larger, lower field devices. The threshold is roughly linear in density, with typical magnitudes of order 10−4. This result implies that locked modes should not be significantly more problematic for the International Thermonuclear Experimental Reactor [I. P. B. Editors, Nucl. Fusion 39, 2286 (1999)] than for existing devices. Coordinated nondimensional identity experiments on the Joint European Torus [Fusion Technol. 11, 13 (1987)], DIII-D [Fusion Technol. 8, 441 (1985)], and C-Mod, with matching applied mode spectra, have been carried out to determine more definitively the field and size scalings. Locked modes on C-Mod are observed to result in braking of core toroidal rotation, modification of sawtooth activity, and significant reduction in energy and particle confinement, frequently leading to disruptions. Intrinsic error fields inferred from the threshold studies are found to be consistent in amplitude and phase with a comprehensive model of the sources of field errors based on “as-built” coil and bus-work details and coil imperfections inferred from measurements using in situ magnetic diagnostics on dedicated test pulses. Use of the A-coils to largely cancel the 2∕1 component of the intrinsic nonaxisymmetric field has led to expansion of the accessible operating space in C-Mod, including operation up to 2 MA plasma current at 8 T.

As part of the ITER Design Review, the physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation.

IntroductionRotation plays an important role in the transition from L-to H-mode [1][2][3][4].Poloidal rotation in the edge plasma region has been closely associated with the L-H transition [5-7], and toroidal momentum confinement is well correlated with energy confinement [8-11]. While there have been several diagnostic systems designed to measure impurity toroidal rotation in tokamak plasmas [8-33], most of the observations have been made in plasmas with an external momentum source, usually provided by neutral beams. Toroidal impurity rotation in ohmic plasmas (no net momentum input) is consistent with neoclassical predictions [24,33,34]; in ohmic L-mode discharges, impurities rotate in the direction opposite to the plasma current, so in general the assumption that the majority ions and impurities (which are most often measured) rotate in the same way might be questionable [34]. In neutral beam heated plasmas, which have substantial direct momentum input, the toroidal momentum confinement time is much shorter than the neoclassical predic- Experiment DescriptionThe observations presented here were obtained from the Alcator C- Mod [35] tokamak, a compact (major radius R = 67 cm, typical minor radius of 22 cm, and Observations of Toroidal Rotation and ScalingsShown in Fig clockwise, and the solid spectrum is blue-shifted, the argon is rotating clockwise, in same direction as the plasma current during the RF pulse. This is in the direction opposite to the rotation of impurities in ohmic L-mode plasmas [8,24,33,34]. The magnitude of the shift is -. 5 mA, which yields a toroidal rotation velocity of (.5 mA/3731.1 mA) x c = 4.0 x 106 cm/s, in the co-current direction. Since the major radius of Alcator C-Mod is 67 cm, this corresponds to an angular rotation speed of 60 kRad/sec.The toroidal rotation may also be determined from the heliumlike argon forbid-4 den line. Shown in Fig.3a and 3b are spectra recorded from a spectrometer viewing the plasma mid-plane, at an angle of 60 from a major radius, in a slight counter clockwise view. The spectra of Fig.3a were obtained from a 1.1 MA discharge (clockwise current), and there is a blue-shift of .13±.01 ml during the H-mode phase (solid spectrum), when the plasma stored energy increased by .12 MJ. This corresponds to a toroidal rotation velocity increase of (.13 mA/3994.3 mX)/sin(6*)x c = 1.0 x 107 cm/s (150 kRad/sec), in the co-current direction. The spectra of Fig.3b were obtained with the same view from a 1.0 MA discharge with the plasma current in the counter clockwise direction, and there is a red-shift of .07t.01 mA during the 2.7 MW ICRF pulse, indicating that the argon is rotating in the counter clockwise direction, again co-current. This L-mode discharge had a stored energy increase of 45 kJ, when the magnitude of the toroidal rotation velocity increased by 5.3 x 106 cm/s, and still opposite to the rotation direction in ohmic discharges.The phasing of the RF antennas for this case was the same as in Fig.3a. Shown in Fig.3c are spectra recorded from a 6* clockwise ...

Divertor detachment may be essential to reduce heat loads to magnetic fusion tokamak reactor divertor surfaces. Yet in experiments it is difficult to control the extent of the detached, low pressure, plasma region. At maximum extent the front edge of the detached region reaches the X-point and can lead to degradation of core plasma properties. We define the ‘detachment window’ in a given position control variable C (for example, the upstream plasma density) as the range in C within which the front location can be stably held at any position from the target to the X-point; increased detachment window corresponds to better control. We extend a 1D analytic model [] to determine the detachment window for the following control variables: the upstream plasma density, the impurity concentration and the power entering the scrape-off layer (SOL). We find that variations in magnetic configuration can have strong effects; increasing the ratio of the total magnetic field at the X-point to that at the target, , (total flux expansion, as in the super-x divertor configuration) strongly increases the detachment window for all control variables studied, thus strongly improving detachment front control and the capability of the divertor plasma to passively accommodate transients while still staying detached. Increasing flux tube length and thus volume in the divertor, through poloidal flux expansion (as in the snowflake or x-divertor configurations) or length of the divertor, also increases the detachment window, but less than the total flux expansion does. The sensitivity of the detachment front location, zh, to each control variable, C, defined as , depends on the magnetic configuration. The size of the radiating volume and the total divertor radiation increase and , respectively, but not by increasing divertor poloidal flux expansion or field line length. We believe this model is applicable more generally to any thermal fronts in flux tubes with varying magnetic field, and similar sources and sinks, such as detachment fronts in stellarator divertors and solar prominences in coronal loops.

I IntroductionAlcator C-MOD', the third high-field compact tokamak in the Alcator line, has been operating tokamak plasmas since May 1993. Its design capability includes toroidal field, BT = 9 T, plasma current I, up to 3 MA, in plasmas with major radius R = 0.67 m, minor radius a = 0.21 m, with elongation up to n = 1.8. Divertor operation can be either into its closed, baffled, divertor chamber or to open flat plates. The magnetic configuration is rather similar to that presently envisaged for the International Thermonuclear Experimental Reactor, ITER, except that it is about a factor of ten smaller.The high particle-, current-and power-densities characteristic of such compact tokamaks lead to edge conditions that are in many respects comparable to those expected in ITER, and offer the opportunity to investigate so-called dissipative divertor operation, in which the power scraped off into the divertor is exhausted through a combination of neutral and radiative processes rather than through plasma conduction direct to the divertor plates.Alcator C-MOD offers excellent port access to the plasma for diagnostic and heating purposes. Its present complement of diagnostics includes full magnetics for equilibrium reconstruction, electron temperature profiles from electron cyclotron emission (ECE), density profiles from a ten-channel CO 2 laser interferometer, ion temperature profiles from high-resolution x-ray doppler measurements, neutron emission, and fast neutral particle analysis, various spectroscopic measurements such as visible bremsstrahlung, H. arrays, and vacuum ultraviolet impurity measurements, bolometer arrays, and x-ray and UV tomography. In addition, detailed edge, scrape-off-layer and divertor diagnosis based on probes and spectroscopy is available.The primary auxiliary heating method in the short term is ICRF, and two transmitters are available, providing a total 4 MW at 80 MHz. Thus far, experiments have concentrated on plasma coupling studies using a movable monopole antenna. Good power coupling into high density plasmas has been obtained, with loading resistance in the range of 5 to 15 Q, 2 in reasonable agreement with the theoretical calculations.So far the magnetic field has been limited to about 5.3 T awaiting power systems upgrades that will enable full-field operation next year. Even so, plasma currents up to 1 MA have been obtained, and durations over 1 second. Peak electron densities up to 9 x 1020 m-3, and temperatures up to T = 2.6, Ti = 1.6 keV have been achieved. Energy confinement is observed to exceed Neo-Alcator scaling.In section II we review some MHD and operational characteristics of the plasma.Section III discusses divertor experiments, section IV the confinement results, and section V the first ICRF coupling studies. II MHD and OperationA unique feature of the design of Alcator C-MOD is its thick stainless-steel vacuum vessel and structure. For reasons of mechanical strength, these have no insulating breaks and thus constitute 'shorted turns' on the ohmic transformer and the eddy ...