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CO 2 pumped FIR (Far Infrared) lasers have a number of valuable uses in the diagnostics arena of fusion research. To maintain lasing, the tuning of the two lasers must be tightly coupled. This coupling presents interesting control problems. A number of analog approaches have been employed to implement tightly coupled tuning control. Such implementations, however, suffer from a lack of flexibility. A new digital controller has been developed that greatly increases the range of control strategies that may be implemented. Of further interest is the controller's capability to provide 96dB of dynamic range when driving -200V piezoelectric transducers. This paper will discuss the algorithms and hardware used in this controller. Lessons learned from the development effort will presented.

The specific size and structure of the edge current profile has important effects on the magnetohydrodynamic stability and ultimate performance of many advanced tokamak (AT) operating modes. This is true for both bootstrap and externally driven currents that may be used to tailor the edge shear. Absent a direct local measurement of j(r), the best alternative is a determination of the poloidal field. Measurements of the precision (0.1°–0.01° in magnetic pitch angle and 1–10 ms) necessary to address issues of stability and control and provide constraints for EFIT are difficult to do in the region of interest (ρ=0.9–1.1). Using Zeeman polarization spectroscopy of the 2S–2P lithium resonance line emission from the DIII-D LIBEAM [D. M. Thomas, Rev. Sci. Instrum. 66, 806 (1995); D. M. Thomas, A. W. Hyatt, and M. P. Thomas, Rev. Sci. Instrum. 61, 340 (1990)] measurements of the various field components may be made to the necessary precision in exactly the region of interest to these studies. Because of the negli...

Sawtooth phenomena have been studied on DIII-D and the Tokamak Fusion Test Reactor (TFTR) [D. Meade and the TFTR Group, in 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]. In the experiments the sawtooth characteristics were studied with fast electron temperature (ECE) and soft x-ray diagnostics. For the first time, measurements of a strong ballistic electron heat pulse were made in a shaped tokamak (DIII-D) [J. Luxon and DIII-D Group, in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Kyoto (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] and the “ballistic effect” was stronger than was previously reported on TFTR. Evidence is presented in this paper that the ballistic effect is related to the fast growth phase of the sawtooth precursor. Fast, 2 ms interval, measurements on DIII-D were made of ...

Experimental evidence is reported of an internal kink instability driven by a new mechanism: barely trapped suprathermal electrons produced by off-axis electron cyclotron heating on the DIII-D tokamak. It occurs in plasmas with an evolving safety factor profile $q(r)$ when ${q}_{\mathrm{min}}$ approaches 1. This instability is most active when ECCD is applied on the high field side of the flux surface. It has a bursting behavior with poloidal/toroidal mode $\mathrm{number}\phantom{\rule{0ex}{0ex}}=\phantom{\rule{0ex}{0ex}}m/n\phantom{\rule{0ex}{0ex}}=\phantom{\rule{0ex}{0ex}}1/1$. In positive magnetic shear plasmas, this mode becomes the fishbone instability. This observation can be qualitatively explained by the drift reversal of the barely trapped suprathermal electrons.

A video camera system capable of imaging vacuum ultraviolet emission in the 120–160 nm wavelength range, from the entire divertor region in the DIII–D tokamak, was designed. The new system has a tangential view of the divertor similar to an existing tangential camera system [M. E. Fenstermacher et al., Rev. Sci. Instrum. 68, 974 (1997)] which has produced two-dimensional maps of visible line emission (400–800 nm) from deuterium and carbon in the divertor region. However, the overwhelming fraction of the power radiated by these elements is emitted by resonance transitions in the ultraviolet, namely the C IV line at 155.0 nm and Ly-α line at 121.6 nm. To image the ultraviolet light with an angular view including the inner wall and outer bias ring in DIII–D, a six-element optical system (f/8.9) was designed using a combination of reflective and refractive optics. This system will provide a spatial resolution of 1.2 cm in the object plane. An intermediate UV image formed in a secondary vacuum is converted to ...

Several important measurements in the International Thermonuclear Experimental Reactor (ITER) diagnostic mission, including the primary one of core helium ash density, are expected to be addressed using active spectroscopic techniques. These methods rely on the use of a dedicated diagnostic neutral beam (DNB) which has been optimized for the dual requirements of beam penetration and charge exchange cross section. For hydrogenic beams, this results in an optimal beam energy of ∼100 keV/amu. Signal-to-noise estimates using realistic geometries and the existing ITER profile and equilibrium data have confirmed the stringent requirements on beam quality and intensity to satisfy the stated ITER measurement precisions. In this article we consider the use of a neutral helium DNB for making active spectroscopic measurements on ITER, since helium beams offer better penetration in dense plasma for a given energy, and the prospects for given source performance may also be improved. Drawbacks include the more difficul...

A series of experiments was conducted on the DIII-D tokamak [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)] to investigate the physical processes which limit density in high confinement mode (H-mode) discharges. The typical H-mode to low confinement mode (L-mode) transition limit at high density near the empirical Greenwald density limit [M. Greenwald et al., Nucl. Fusion 28, 2199 (1988)] was avoided by divertor pumping, which reduced divertor neutral pressure and prevented formation of a high density, intense radiation zone (MARFE) near the X-point. It was determined that the density decay time after pellet injection was independent of density relative to the Greenwald limit and increased nonlinearly with the plasma current. Magnetohydrodynamic (MHD) activity in pellet-fueled plasmas was observed at all power levels, and often caused unacceptable confinement degradation, except when the neutral beam injected (NBI) power was ⩽3 MW. Formation of MARFEs on closed field lines was avoided with lo...

A new reciprocating Langmuir probe was used to measure density and temperature profiles, ion flow, and potential fluctuation levels from the lower divertor floor up to the X point on the DIII-D Tokamak. This probe is designed to make fast (2 kHz swept, 20 kHz Mach, 500 kHz Vfloat) measurements with 2 mm spatial resolution in the region where the largest gradients on the plasma open flux tubes are found and therefore provide the best benchmarks for scrap-off layer and divertor numerical models. Profiles are constructed using the 300 ms time history of the probe measurements during the 25 cm reciprocating stroke. Both single and double null plasmas can be measured and compared with a 20 Hz divertor Thomson scattering system. The probe head is constructed of four different kinds of graphite to optimize the electrical and thermal characteristics. Electrically insulated pyrolytic graphite rings act as a heat shield to absorb the plasma heat flux on the probe shaft and are mounted on a carbon/carbon composite core for mechanical strength. The Langmuir probe sampling tips are made of a linear carbon fiber composite. The mechanical, electrical, data acquisition, and power supply systems will be described. Initial measurements will also be presented.

The next generation of tokamaks present unique challenges to plasma diagnostic design due to the physical size of the devices and the radiation environment. The need for a density measurement for density feedback control for a prototype reactor such as ITER is well established and several proposals for line average measurements have been put forward. In this article, a design for a line average density diagnostic for ITER using collinear interferometry and Faraday rotation measurements will be presented. Plasma effects on both types of measurements and density resolution will be discussed along with the possibility of combining the information from the two collinear measurements to improve the reliability and quality of the density profile. Survivability of the plasma facing mirrors, in particular the surface flatness and surface roughness, are critical issues and preliminary analysis suggests these may limit the wavelength of probing beams. Thermal and stress analysis of the plasma facing mirrors will be...

Strong shaping, favourable for stability and improved energy confinement, together with a significant expansion of the central region of improved confinement in negative central magnetic shear target plasmas, increased the maximum fusion power produced in DIII-D by a factor of 3. Using deuterium plasmas, the highest fusion power gain, the ratio of fusion power to input power, Q, was 0.0015, corresponding to an equivalent Q of 0.32 in a deuterium-tritium plasma, which is similar to values achieved in tokamaks of larger size and magnetic field. A simple transformation relating Q to the stability parameters is presented

Detailed analysis of recent high beta discharges in the DIII-D tokamak demonstrates that the resistive vacuum vessel can provide stabilization of low n magnetohydrodynamic (MHD) modes. The experimental beta values reaching up to {beta}{sub T} = 12.6% are more than 30% larger than the maximum stable beta calculated with no wall stabilization. Plasma rotation is essential for stabilization. When the plasma rotation slows sufficiently, unstable modes with the characteristics of the predicted {open_quotes}resistive wall{close_quotes} mode are observed. Through slowing of the plasma rotation between the q = 2 and q = 3 surfaces with the application of a non-axisymmetric field, the authors have determined that the rotation at the outer rational surfaces is most important, and that the critical rotation frequency is of the order of {Omega}/2{pi} = 1 kHz.

Externally applied, static, resonant low m=1, n=1 magnetic field perturbations have a strong effect on the behaviour of the sawtooth crash in the DIII-D tokamak. An external coil was used to produce magnetic field perturbations large enough to slow the core plasma rotation and greatly alter the radial electric field profiles but not large enough to induce a locked mode. In these cases the global plasma parameters and the density and temperature profiles remained unchanged; however, the sawtooth oscillation, in particular the sawtooth crash, changed character from resistive in nature to ideal as the static error field increased. The m=1, n=1 resistive MHD internal kink mode normally responsible for the sawtooth crash in DIII-D appears to be stabilized by resonance detuning caused by the combination of the applied static error field and the slower plasma toroidal rotation. The stabilization of the resistive internal kink allowed the plasma to evolve until an ideal mode caused a sawtooth crash. This effect may explain the differences in sawtooth behaviour reported between different tokamaks

Accurate equilibrium reconstruction and detailed stability analysis of a strongly shaped, double‐null, βT=11% discharge shows that the plasma core is in the second stable regime to ideal ballooning modes. The equilibrium reconstruction using all the available data (coil currents, poloidal magnetic loops, motional Stark effect data, the kinetic pressure profile, the magnetic axis location, and the location of the two q=1 surfaces) shows a region of negative magnetic shear near the magnetic axis, an outer positive shear region, and a low shear region connecting the two. The inner negative shear region allows a large positive shear region near the boundary, even at low q (q95=2.6), permitting a large outer region pressure gradient to be first regime stable. The inner region is in the second stable regime, consistent with the observed axial beta [βT(0)=44%]. In the low shear region p’ vanishes, consistent with Mercier stability. This is one way to extend the ballooning limit in shaped plasmas while maintainin...

There is a need for a real time, reliable density measurement compatible with the restricted access and radiation environment on International Thermonuclear Experimental Reactor (ITER). Due to the large plasma path length, high density and field, refraction, and Faraday rotation effects makes the use of contemporary long wavelength (≳50 μm) interferometers impractical. In this paper we consider the design of a short wavelength vibration compensated interferometer which allows operation without a prohibitively large vibration isolated structure and permits the optics to be conveniently mounted directly in or on the tokamak. A density interferometer design for ITER incorporating a 10.6 μm CO2 interferometer with vibration compensation provided by a 3.39 μm HeNe laser is discussed. The proposed interferometer design requires only a small intrusion into the ITER tokamak without a large support structure, refraction and Faraday rotation problems are avoided, and it provides a density resolution of at least 0.5%. Results are presented from an interferometer installed on the DIII‐D tokamak incorporating essential elements of the proposed ITER design including 10.6 and 3.39 μm lasers, a retroreflector mounted on the vacuum wall of the DIII‐D tokamak and real‐time density feedback control. In this paper we consider a short wavelength interferometer design that incorporates vibration compensation for use on ITER. Our primary concern is to develop an interferometer design that will produce a reliable real time density monitor. We use the ITER conceptual design activity report as the basis of the design.

This paper describes the design and operation of a 40 spatial channel Thomson scattering system that uses multiple 20‐Hz Nd:YAG lasers to measure the electron temperature and density profiles periodically throughout an entire plasma discharge. As many as eight lasers may be fired alternately for an average measurement frequency of 160 Hz, or they may be fired in rapid succession (

Electron-cyclotron-current-drive (ECCD) experiments performed in the DIII-D tokamak have produced rf-driven currents of up to 100 kA. The experimental results, which exceed predictions using linear theory, are enhanced by the presence of a residual, toroidal dc electric field. These ECCD experiments are performed with plasma conditions sufficient to result in strong localized deposition of the rf power and good confinement of the rf-generated current carriers. These improved conditions permit a test of theory under reactor relevant conditions. Theoretical predictions obtained using a Fokker-Planck code are in good agreement with the experimental results when effects due to electron trapping and the residual dc electric field are included.

The effect on H-mode energy confinement of operating at low values of the plasma safety factor q has been investigated experimentally on the DIII-D tokamak (see Plasma Physics and Controlled Nuclear Fusion Research 1986 (Poroc. 11th Int. Conf. Kyoto, 1986), Vol.1, IAEA, Vienna (1987) 159). Operationally, the linear increase of H-mode energy confinement with plasma current ceases when the ratio of the plasma current to the toroidal magnetic field is high than approximately=1 MA/T. Experimentally, it has been determined that this saturation of confinement is not the result of a ceiling imposed by saturated Ohmic confinement, by the plasma reaching a toroidal beta limit, by enhanced plasma radiation or by edge localized modes. The observed saturation of confinement is found to be quantitatively consistent with a model of global energy confinement degradation resulting primarily from the macroscopic phenomena of sawteeth. The experimental data are also qualitatively consistent with an alternative explanation for the confinement saturation which involves enhanced transport brought about by ion temperature gradient driven turbulence

A theoretical and experimental evaluation of axisymmetric stability and axisymmetric control has led to a modification of the vertical position control in the DIII‐D tokamak, which now allows operation to within a few percent of the ideal magnetohydrodynamic (MHD) n=0 limit. It is found that the onset the departure from rigid shift behavior in D‐shaped plasmas limits plasma elongation to 2.5 in DIII‐D. The possibility of avoiding the vertical instability in future tokamaks with highly elongated plasmas is discussed. Recent experiments have focused on utilizing this capability for axisymmetric control to construct plasma shapes optimized to increase the achievable beta. Operation near the axisymmetric stability limit allows an increase in the achieved normalized current Ip/aBT, where Ip is the total plasma current, a is the minor radius, and BT is the toroidal field. Based on stability calculations, an equilibrium was developed to achieve marginal stability simultaneously to axisymmetric, kink, and balloon...

One of the characteristic physical principles of our recently proposed theory on the x‐ray‐energy response of a semiconductor x‐ray detector is the effect of the three‐dimensional diffusion of x‐ray‐produced charges in a semiconductor field‐free substrate region. It is found and reported that the quantum efficiency of a semiconductor x‐ray detector is enhanced when this phenomenon is taken into account, and the data on the x‐ray‐energy response of the semiconductor detector using monochromatized synchrotron radiation at the Photon Factory are actually well fitted by the theory. In addition, we measure the diffusion length L of the detector using a precisely collimated x‐ray narrow beam and a multichannel detector array setting on a μm‐order position‐controllable stage. The verification of the consistency between this observed value of L and the predicted value of L from our theory on the quantum efficiency, has been achieved.

The energy response of a new semiconductor‐detector array in the DIII–D tokamak for plasma x‐ray diagnostics is characterized using synchrotron radiation at the Photon Factory (KEK) in Japan. This international cooperation research clarifies its good x‐ray response ranging up to about 10 keV. The variations in detector‐channel responses are also studied. Furthermore, new aluminum‐coated x‐ray semiconductor detectors are designed, fabricated, and characterized for the 10‐keV‐order DIII–D electron studies including the purpose of rf current‐drive investigations; a flat response (the quantum efficiency η∼1) up to around 10‐keV x rays, and a high efficiency (η≥0.3) even for the 20‐keV x rays, are achieved. A wide variety of x‐ray detection systems for the x‐ray‐energy range from a few tens eV to several hundreds keV in the GAMMA 10 tandem mirror is shown; that is, microchannel‐plate tomography systems for the 0.01–60‐keV x rays, a new ultra‐low‐ energy‐sensitive pure‐Ge detector (0.2–50 keV), a Si(Li) detecto...

Silicon diodes are commonly used for x‐ray measurements in the soft‐x‐ray region between a few hundred eV and 20 keV. Recent work by Cho et al. [J. Appl. Phys. 72, 8 (1992)] has shown that the charge collecting region in an underbiased silicon detector is the depletion depth plus some contribution from a region near the depleted region due to charge diffusion. The depletion depth can be fully characterized as a function of the applied bias voltage and is roughly proportional to the square root of the bias voltage. We propose a technique to exploit this effect to use the silicon within the detector as an actively controlled x‐ray filter. With reasonable silicon manufacturing methods, a silicon diode detector can be constructed in which the sensitivity of the collected charge to the impinging photon energy spectrum can be changed dynamically in the visible to above the 20‐keV range. This type of detector could be used to measure the electron temperature in, for example, a tokamak plasma by sweeping the appl...

The authors describe the design and operation of a 40 spatial channel Thomson scattering system that uses multiple 20-Hz Nd:YAG lasers to measure the electron temperature and density profiles periodically throughout an entire plasma discharge. Interference filter polychromators disperse the scattered light, which is detected by silicon avalanche photodiodes. The measurable temperature range is from 10 eV to 20 keV, and the minimum detectable density is about 2*10/sup 18/ m/sup -3/. Laser control and data acquisition are performed in real time by a VME-based microcomputer. Data analysis is performed by a MicroVAX 3400. Unique features of this system include burst mode operation, where multiple lasers are fired in rapid succession ( >

A silicon avalanche photodiode with an internal gain of about 50 to 100 is used in a temperature‐controlled environment to measure the Nd:YAG laser Thomson scattered spectrum in the wavelength range from 700 to 1150 nm. A charge‐sensitive preamplifier has been developed for minimizing the noise contribution from the detector electronics. Signal levels as low as 20 photoelectrons (S/N=1) can be detected. Measurements show that both the signal and the variance of the signal vary linearly with the input light level over the range of interest, indicating Poisson statistics. The signal is processed using a 100 ns delay line and a differential amplifier which subtracts the low‐frequency background light component. The background signal is amplified with a computer‐controlled variable gain amplifier and is used for an estimate of the measurement error, calibration, and Zeff measurements of the plasma. The signal processing has been analyzed using a theoretical model to aid the system design and establish the pro...

A VME‐based real‐time computer system for laser control, data acquisition, and analysis for the DIII‐D multipulse Thomson scattering diagnostic is described. The laser control task requires precise timing of up to eight Nd:YAG lasers, each with an average firing rate of 20 Hz. A cpu module in a real‐time multiprocessing computer system will operate the lasers with evenly staggered laser pulses or in a ‘‘burst mode,’’ where all available (fully charged) lasers can be fired at 50–100 μs intervals upon receipt of an external event trigger signal. One or more cpu modules, along with a LeCroy FERA (fast encoding and readout ADC) system, will perform real‐time data acquisition and analysis. Partial electron temperature and density profiles will be available for plasma feedback control within 1 ms following each laser pulse. The VME‐based computer system consists of two or more target processor modules (25 MHz Motorola 68030) running the VMEexec real‐time operating system connected to a Unix‐based host system (also a 68030). All real‐time software is fully interrupt driven to maximize system efficiency. Operator interaction and (non‐real‐time) data analysis takes place on a MicroVAX 3400 connected via DECnet.

A multilaser system is being developed for the DIII‐D Thomson scattering diagnostic. This system combines the beams from up to 8 Nd:YAG lasers onto a common beamline in which the beams are nearly parallel and are all focused into a small, common area within the desired scattering volume. Each laser can be fired at a constant rate (20 Hz per laser) for a high average repetition rate, or together in a ‘‘burst,’’ which will give very high sampling rates (10–20 kHz) for short periods. The burst mode will be triggerable by plasma events, which will allow for study of transient phenomena, but will require nonperiodic firing of the lasers. Beamline diagnostics include position sensitive detectors for computer controlled feedback alignment of the 35 m beamline, an image position detection system for monitoring the alignment of the collection lens to the scattering volume, and a 1‐D reticon camera for divergence monitoring. The effects of the nonperiodic firing of the lasers will be monitored with the reticon camera.

A seven channel polychromator, utilizing high performance interference filters (transmission ≳80%, rejection at laser wavelength>105), has been tested for use in the multi‐Nd:YAG laser Thomson scattering system for the DIII‐D tokamak. Unique features of this polychromator are the combination of high throughput, easy alignment, flexibility, compact size, and low cost when compared with other alternatives. Light is introduced to the polychromator (f/1.75) via a fiber optic bundle which permits the use of small (3.0 cm diam) optics and leads to a compact (44×24×8 cm) design, an important design consideration for multiple polychromator systems. The light is cascaded through a series of different bandpass interference filters and relay lenses which are mounted on two precision parallel rails in such a way that alignment is trivial. The relay lenses are positioned directly in front of the filters so that light reflected from the filter passes through the lens twice. This leads to an efficient, compact design an...

The measurement of low-Z impurities, in particular He ash, in the core of International Thermonuclear Experimental Reactor (ITER) remains an outstanding diagnostic issue. The only credible candidate at present is active charge exchange recombination spectroscopy (CER) utilizing a diagnostic neutral beam (DNB) optimized for the dual requirements of beam penetration and charge exchange cross section, resulting in beam energies of 2∼100 keV/AMU. Using the existing ITER parameter profile and equilibrium data files and reasonable assumptions regarding viewing optics and DNB performance, we have employed a benchmarked multistep beam penetration code to yield signal-to-noise estimates for possible core helium concentration measurements. These studies confirm the importance of precise determination of beam intensities via accurate modeling and independent measurement, as well as the need for beam modulation, to satisfy the stated measurement precisions needed for ITER. Comparable calculations have been done for an intense pulsed neutral beam based on ion diode technology, as well as other candidate He-CER wavelengths, to assess the relative advantages of these techniques. Since any DNB-based diagnostic system actually deployed on ITER will likely be used for a variety of purposes, signal-to-noise calculations for the related active CER measurement of ion temperatures have also been performed and will be presented.

It is generally agreed that the primary scientific goal of the first near Sun fly-by mission is to identify the source of momentum responsible for driving the solar wind and to understand the physics of solar coronal heating. In this paper, we present an innovative concept for obtaining a unique set of data with which to better understand the physics of solar wind acceleration and coronal heating. The instrument, called the Solar Event Correlation Analyzer (SECA), operates on the principle that the topology of the open magnetic field lines can be used to causally correlate the evolution of small scale structures (events) in the Sun’s transition region with in situ solar probe measurements. These small-scale events are generally viewed as the mechanisms providing energy and momentum for the solar corona and wind, an assumption that will be conclusively tested with the SECA instrument. This paper describes the SECA concept and its scientific justification.

Author(s): Lazarus, EA; Navratil, GA; Greenfield, CM; Strait, EJ; Austin, ME; Burrell, KH; Casper, TA; Baker, DR; DeBoo, JC; Doyle, EJ; Durst, R; Ferron, JR; Forest, CB; Gohil, P; Groebner, RJ; Heidbrink, WW; Hong, R; Houlberg, WA; Howald, AW; Hsieh, C; Hyatt, AW; Jackson, GL; Kim, J; Lao, LL; Lasnier, CJ; Leonard, AW; Lohr, J; La Haye RJ; Maingi, R; Miller, RL; Murakami, M; Osborne, TH; Perkins, LJ; Petty, CC; Rettig, CL; Rhodes, TL; Rice, BW; Sabbagh, SA; Schissel, DP; Scoville, JT; Snider, RT; Staebler, GM; Stallard, BW; Stambaugh, RD; St John HE; Stockdale, RE; Taylor, PL; Thomas, DM; Turnbull, AD; Wade, MR; Wood, R; Whyte, D | Abstract: Fusion power has been increased by a factor of 3 in DIII-D by tailoring the pressure profile to avoid the kink instability in H-mode plasmas. The resulting plasmas are found to have neoclassical ion confinement. This reduction in transport losses in beam-heated plasmas with negative central shear is correlated with a dramatic reduction in density fluctuations. Improved magnetohydrodynamic stability is achieved by controlling the plasma pressure profile width. In deuterium plasmas the highest gain Q (the ratio of fusion power to input power), was 0.0015, corresponding to an equivalent Q of 0.32 in a deuterium-tritium plasma. © 1996 The American Physical Society.

There is a need for real time, reliable density measurement for tokamak plasma density control, compatible with the restricted access and radiation environment on ITER. Line average density measurements using microwave or laser interferometry techniques have proven to be robust and reliable for density control on contemporary tokamaks. In ITER, the large path length, high density and high density gradients, limit the wavelength of a probing beam to shorter then about 50 µm due to refraction effects. In this paper we consider the design of short wavelength vibration compensated interferometers and Faraday rotation techniques for density measurements on ITER. These techniques allow operation of the diagnostics without a prohibitively large vibration isolated structure and permit the optics to be mounted directly on the radial port plugs on ITER. A beam path designed for 10.6 µm (CO2 laser) with a tangential path through the plasma allows both an interferometer and a Faraday rotation measurement of the line average density with good density resolution while avoiding refraction problems. Plasma effects on the probing beams and design tradeoffs will be discussed along with radiation and long pulse issues. A proposed layout of the diagnostic for ITER will be presented.

Statistical analysis of a large set of current profile data obtained for sawtoothing discharges in the DIII-D tokamak indicates that the safety factor (inverse rotational transform) in the plasma center, ${\mathit{q}}_{0}$, is close to 1.0. The increase of ${\mathit{q}}_{0}$ associated with the sawtooth crash is found to be of the order of 0.05. The results of time resolved measurements of the current profile and the soft x-ray emissions support a model of the sawtooth instability based on total reconnection of the m=1 magnetic island.

Neutral beams were injected into low field (B = 0.7-1.0 T) deuterium plasmas in an attempt to destabilize toroidicity induced Alfvén eigenmodes (TAE modes). When the parallel beam velocity approached the Alfvén velocity and the volume averaged beam beta exceeded 2%, localized, propagating modes with n = 2-10 were observed. As much as 45% of the beam power was lost as a result of the modes. The threshold for TAE instability is at least one order of magnitude higher than that predicted by Fu and VanDam (Phys. Fluids B 1 (1989) 1949). © 1991, IOP publishing Ltd.

Recent results from the DIII-D tokamak have provided significant contributions to the understanding of many of the elements of tokamak physics and the application of this understanding to the design of next generation devices including ITER and CIT. The limitations of magnetohydrodynamic stability on the values of plasma beta (the ratio of kinetic pressure to the containing pressure of the magnetic field) that can be attained has been experimentally demonstrated and found to be described by existing theory. Values of beta (10.7%) well in excess of those required for proposed devices (ITER and CIT) have been demonstrated. Regimes of confinement (H-mode) have been established that scale favorably to proposed next generation devices, and experiments demonstrating the dependence of the energy confinement on plasma size have been completed. Understanding of confinement is rapidly developing especially in the areas of bulk transport and the role of turbulence in the plasma edge. Key experimental results in areas of plasma transport and edge plasma phenomena are in agreement with theories based on short wavelength turbulence. Control of the divertor heat loads and impurity influx has been demonstrated, and new progress has been made in the understanding of plasma edge phenomena. Experiments with ion Bernstein wave heating have not found regimes in which these waves can produce effective central ion heating. Electron cyclotron current drive experiments have demonstrated 70 kA of driven current in 400 kA discharges.

We propose use of a fast Alfven wave interferometer and reflectometer as a tokamak diagnostic to measure the plasma mass density, D–T species mix profile, and density fluctuations. Utilizing the property that the phase velocity of the fast wave propagating across the magnetic field is the Alfven speed with thermal correction, this fast wave interferometer on the DIII-D tokamak was successfully used to obtain the line integrated density. Since the position of the ion–ion hybrid cutoff in tokamaks is uniquely determined by the species mix ratio and wave frequency, the reflectometer arrangement finds the species mix profile. The inversion method of reflectometry is discussed. The multiple chord interferometer also measures the mass density fluctuation profile.

The injection of Ar into the region of the DIII-D divertor is a promising technique for energy dissipation (through radiation and collisions) and consequently for reduction of the heat load on the plates. An important problem related to this technique, is the inherent poisoning of the core plasma by migrating Ar. The Ar core contamination seems also to improve the thermal transport in an advanced operating mode of the tokamak. It is therefore of great importance to measure the evolution of the impurity concentration profile within the core plasma. This goal could be achieved by using the Ross filter method in conjunction with the existing x-ray diagnostics on DIII-D. A basic Ross filter system consists of two identical detectors placed behind two different x-ray absorbing foils looking at the same plasma volume. The foils are made of different elements or compounds with adjacent or nearly adjacent atomic numbers. Their accurate thickness causes the x-ray transmission curves of the two foils to be effectiv...

Line density measurements on ITER are expected to be provided by a conventional vibration-compensated CO2 laser interferometer. This type of interferometer relies on maintaining an accurate measurement of hundreds of vibration fringes during the entire time history of the plasma. Because of the very long pulse lengths (1000 s) and the possibility of fringe counting errors, a second method that measures the Faraday rotation of a laser beam with a tangential laser beam path through the plasma will be used to provide a history-independent measurement of the line density with somewhat reduced accuracy. These two measurements can be accomplished in a combined interferometer/polarimeter system that uses a single CO2 laser and detector. We have constructed a prototype of this type of system, which uses two acousto-optic modulators to generate two beams, shifted from the laser frequency by 40 and 45 MHz. These probe beams are converted to left- and right-hand circular polarization and combined into a single probe...

Techniques to reduce the heat flux to the divertor plates in tokamak power plants and the consequent erosion of, and subsequent damage to the divertor target plates include the injection of impurities such as argon, that can dissipate the energy (through radiative or collisional processes) before it reaches the target plates. An important issue with this type of scheme is poisoning of the plasma core by the impurities introduced in the divertor region. Subsequently, there is a desire to measure the profiles of the injected impurities in the core. X-ray Ross filters with an effective narrow band pass centered on the argon Kα line at 3.2 keV, have been installed on two of the existing x-ray arrays on DIII–D in order to help determine the argon concentration profiles. Emissivity profiles of the Kα lines and the emissivity profiles for the argon enhanced continuum can be inferred from the inverted filtered x-ray brightness signals if Te, ne, and Ar18+ profiles are known. The MISTReference 1 code is used to co...

Neutral beam heated DIII-D expanded boundary divertor discharges have exhibited ASDEX-like H-mode behaviour over a wide parameter range. The deuterium H-mode energy confinement of 120 ms remained near the Ohmic value for up to 6 MW of neutral beam heating, where it was 2-2.5 times higher than the L-mode value at a plasma current of 1 MA. The hydrogen and helium H-mode energy confinement times were similar and substantially below the deuterium H-mode confinement time. The H-mode confinement times decreased with increasing neutral beam power and were only 30% better than the L-mode confinement times at 5 MW. In an H-mode with a mixture of hydrogen and deuterium ([H]/[H+D] 40%), the confinement time was in between the values obtained in the pure hydrogen and deuterium cases, increased linearly with plasma current for q95 > 3.2, and decreased with increasing neutral beam power. The confinement quality in these plasmas was 85 ms per MA at a heating power of 5.6 MW. The lower energy confinement in the non-deuterium H-modes and the degradation of energy confinement with neutral beam power were both accompanied by an increase in the edge localized mode (ELM) amplitude and frequency. The changing ELM characteristics make a determination of the intrinsic isotopic and neutral beam effect on confinement difficult. For values of BT < 0.9 T and q95 < 3, the confinement quality in the deuterium and hydrogen/deuterium H-modes deteriorated to values near the L-mode level. This deterioration in energy confinement was not related to operation at high beta but instead appears to be due to a combined action between sawteeth and ELMs that becomes more pronounced at low q and low BT. L-mode energy confinement was independent of ion species and in good agreement with Kaye-Goldston scaling. Odajima-Shimomura scaling disagreed with the present L-mode τE data in terms of isotopic mass dependence; their prediction for the hydrogen L-mode exceeds the present measurements by a factor of two.

Observation of the intensity of the recycling particle flux at the main plasma edge for various limiter and divertor discharges indicates that the gross energy confinement of beam-heated discharges is closely related to the intensity of the edge particle flux. In limiter discharges, the global particle confinement time and the energy confinement time τE show many similarities: 1) linear Ip dependence at Ip < 600 kA, 2) no BT dependence, and 3) deterioration against injection power. Improvement of τE by increasing Ip, for example, is associated with high temperatures at the plasma edge region accompanied by reduced particle recycling. – Divertor discharges with low particle recycling around the main plasma show better energy confinement than limiter discharges at high plasma densities. The improvement of τE is primarily originated in the reduction of heat transport at the main plasma edge region, which is associated with the reduction of recycling particle flux at the main plasma edge. Under certain operation condition, for example, excessive cold-gas puffing, the discharge shows relatively high scrape-off plasma density and strong particle recycling between the main plasma and the limiter. The energy confinement time of these discharges degrades somewhat or reduces completely to that of the limiter discharge. – In low-recycling divertor discharges, the central electron and ion temperature is proportional to the injection power, and the plasma stored energy is proportional to ePabs (scales as INTOR scaling). With ≈ 4 MW beam injection, high-temperature and high-density plasmas were obtained (stored energy up to 280 kJ, Te(0) ≈ Ti(0) ≈ 2.5–3.0 keV at e ≈ (6–7) × 1013 cm−3, τE* ≈ 70 ms).

The DIII-D high and low beta stability results have established the basic feasibility of the divertor and H-mode configurations up to elongation 2.0 for next generation tokamaks. The 6.8% beta T achieved has already exceeded projected operating requirements of next generation devices. beta T > 6% has been sustained for 800 ms. Stability calculations and patterns of MHD mode behavior suggest a central expanding zone of ballooning instabilities leads ultimately to unstable m/n=2/1 modes which cause beta collapse or disruption. The pressure gradient at the plasma edge just reaches the first regime ballooning limit prior to ELMs. ECRH has proven effective for generating H-mode, sawtooth suppression, and ELM suppression.

Results are presented for multiple deuterium pellet injection into both the Ohmic phase and the beam heated phase of Doublet III discharges. Pellet injection experiments were carried out using an injector, built at Oak Ridge National Laboratory, which has the capability of firing up to 40 hydrogen or deuterium pellets per second at a speed of 800 ms−1. For the deuterium pellets used in this experiment, one pellet raises the line average density to 1.3 × 1013 cm−3. In the Ohmic phase, the pellets penetrate deep into the plasma, but with 2.2 MW of neutral beam heating they penetrate only half as deep. This difference is mostly due to the increased ablation in the outer 25% of the plasma during beam heating. Ablation calculations using the neutral shielding model agree well with experimental results during Ohmic heating, but they predict a shallower pellet penetration than is actually observed during neutral beam heating. About 90–100% of the pellet mass can be accounted for after injection. Electron temperature profile relaxation accompanies pellet injection, but on a time-scale that is shorter than the pellet lifetime. A transient increase in the energy confinement time τE is possible, provided pellet penetration is sufficient to sustain a peaked density profile. In discharges with one or more multifractured pellets, which ablate rapidly at the plasma edge and thus broaden the density profile, no increase in τE is seen. These results indicate that continuous τE improvement may be possible if a peaked density profile can be maintained by sustained deep pellet penetration.

A neutral beam heated, β maximizing discharge at low qa in Doublet III ending in disruption is studied and correlated with theoretical models. This discharge achieved MHD β-values close to the theoretical Troyon-Sykes-Wesson limit in its evolution. The MHD phenomena of this discharge are analysed. The sequence of events leading to the high β disruptions is hypothesized as follows: the current and pressure profiles are broad-ened continuously by neutral beam injection. A last sawtooth internal disruption initiates an (m/n = 2/1) island through current profile steepening around the q = 2 surface. The loss of plasma through stochastic field lines slows the island rotation and enhances its interaction with the limiter. The resultant enhanced island growth through island cooling or profile change enlarged the edge stochastic region. The overlapping of the edge stochastic region with the sawtooth mixing region precipitated the pressure disruption. Thus, in our hypothetical model for this discharge, β increase by neutral beam heating does not directly cause the disruption but ushers the plasma indirectly towards it through the profile broadening process and contributes to the destabilization of the 1/1 and 2/1 tearing modes.

In DIII-D H-mode plasmas, large amplitude modulation of steep edge electron pressure gradients are observed during the occurrences of giant edge-localized modes (ELM's). For a considerable range of plasma currents (1 to 2 MA), the experimental pressure gradient before a giant ELM is close to the theoretically predicted marginal stability limit for the first regime of the ideal ballooning mode. These results, together with observations showing that ELM's originate in the region of bad curvature, strongly suggest that ELM's are triggered by ballooning mode instabilities.

Electron‐cyclotron heating, using second harmonic waves launched from the low field side, has a strong effect on sawteeth in the DIII‐D tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (IAEA, Vienna, 1987), Vol. I, p. 159]. There are two distinct methods and mechanisms for sawtooth period extension. In the first case a factor of 6 increase in sawtooth period, compared to that during Ohmic heating, occurs with electron‐cyclotron heating (ECH) localized near the q=1 surface. The sawtooth period increases faster than linearly with increasing ECH power. Qualitatively, there is agreement with sawtooth suppression models, although details of the accompanying m/n=1/1 oscillation appear inconsistent. The second case is a technique for complete sawtooth suppression in which impurities and off‐axis heating broaden the current profile and drive q0>1. It appears that the temperature and the current and q profiles inside q=1 are determined by the magnetohydrodynamic activity, which is modified by E...

Limiter, single‐null, and double‐null divertor plasmas have been heated by hydrogen neutral beam injection in the direction opposite to that of the plasma current. The H‐mode power threshold is reduced by approximately 25% for counterinjection. Sawtooth suppression was possible only for high neutral beam power and for high values of the safety factor. No confinement improvement over similar coinjection discharges was observed in H‐mode, L‐mode, or limiter discharges. In plasmas without sawteeth, metallic impurity buildup in the central region of the plasma eventually leads to a reduction in confinement. The parametric dependence of energy confinement for both L‐and H‐mode plasmas was found to behave in the same manner during similar coinjection experiments. H‐mode confinement increased linearly with increasing plasma current (65 msec/MA at 9 MW). Both L‐mode and H‐mode energy confinement decreased with increasing neutral beam power. At 1.25 MA, the L‐mode plasmas had a confinement time that was less than that predicted by Kaye–Goldston scaling.

Neutral-beam-heated expanded-boundary (XB) divertor discharges have been obtained in Doublet III with high heating efficiency for wide ranges of plasma parameters (Ip: 300–800 kA; BT: 8–24kG; e :(2–10) × 1013cm−3; Pb 3 MW), a mild deterioration of the energy confinement time has been observed.

In high-power beam heated divertor discharges with P IN ~ 7.5 MW , the central radiative power density from nickel can be suppressed to less than 10% of the heating power density. The radiative loss in the divertor increased as the main electron density was increased. 24% of the absorbed power was observed to be radiated in the divertor in high-density, H-mode discharges. During a density scan, the radiative power from the main plasma remained constant. The radiative power can be increased by injection of argon gas to 75% of the absorbed power which resulted in a significant decrease in the heat load of the divertor plate. The electron temperature and the particle flux were measured at the divertor plate. With the increase of the main plasma density, the electron temperature decreased to 8 eV (4.5 MW absorbed power). The electron density at the divertor plate was 3×10 14 cm −3 . All these experimental results indicate the feasibility of the single-null open divertor to a fusion reactor in suppressing central radiative loss, reducing the heat load and erosion of the divertor plate and facilitating ash exhaust.

The authors' data indicate that the L-mode to H-mode transition in the DIII-D tokamak is associated with the sudden reduction in anomalous, fluctuation-connected transport across the outer midplane of the plasma. In addition to the reduction in edge density and magnetic fluctuations observed at the transition, the edge radial electric field becomes more negative after the transition. They have determined the scaling of the H-mode power threshold with various plasma parameters; the roughly linear increase with plasma density and toroidal field are particularly significant. Control of the ELM frequency and duration by adjusting neutral beam input power has allowed us to produce H-mode plasmas with constant impurity levels and durations up to 5 s. Energy confinement time in ohmic H-mode plasmas and in deuterium H-mode plasmas with deuterium beam injection can exceed saturated ohmic confinement times by at least a factor of two. Energy confinement times above 0.3 s have been achieved in these beam-heated plasmas with plasma currents in the range of 2.0 to 2.5 MA. Local transport studies have shown that electron and ion thermal diffusivities and angular momentum diffusivity are comparable in magnitude and all decrease with increasing plasma current.

Analysis of the expanded boundary divertor in Doublet III reveals a medium density operating regime ( n e ≲ 4 × 10 13 cm -3 ), in which (1) the divertor density ne(Div) rises with the cube of the main plasma line average density, n e , and (2) the neutral pressure at the exhaust end of the divertor has a quadratic dependence on n e (with divertor plasma temperatures ≲ 15 eV). In this regime, injection of argon into the discharge results in a steady state concentration of argon, which is two orders of magnitude higher in the divertor plasma than in the main discharge. This trapping effect suggests that the argon concentration in the divertor region may be useful as an externally controllable parameter; that is, injected argon effectively cools the plasma boundary, yet is not deleterious to the plasma as a whole. A model which successfully describes the behavior of the edge and divertor plasmas in ohmically heated discharges suggests that these desirable trapping and radiative features may be maintained in the high powered neutral beam heating phase.

Tokamak discharges using the expanded boundary divertor in the DIII-D device exhibit H-mode confinement. With neutral-beam power up to 6 MW, energy confinement remains comparable to the Ohmic value at a plasma current of 1 MA. Confinement is also independent of plasma density and toroidal field. Confinement increases with plasma current, but the exact functional dependence is, as yet, uncertain. These results show that the H mode can be achieved in a reactor-compatible open divertor configuration.