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 QDTeq 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.

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Internal transport barriers in tokamak plasmas

Recent experiments (Synakowski et al 2004 Nucl. Fusion 43 1648, Lloyd et al 2004 Plasma Phys. Control. Fusion 46 B477) on the Spherical Tokamak (or Spherical Torus, ST) (Peng 2000 Phys. Plasmas 7 1681) have discovered robust plasma conditions, easing shaping, stability limits, energy confinement, self-driven current and sustainment. This progress has encouraged an update of the plasma conditions and engineering of a Component Test Facility (CTF), (Cheng 1998 Fusion Eng. Des. 38 219) which is a very valuable step in the development of practical fusion energy. The testing conditions in a CTF are characterized by high fusion neutron fluxes Γn ≈ 8.8 × 1013 n s−1 cm−2 ('wall loading' WL ≈ 2 MW m−2), over size-scale >105 cm2 and depth-scale >50 cm, delivering >3 accumulated displacement per atom per year ('neutron fluence' >0.3 MW yr−1 m−2) (Abdou et al 1999 Fusion Technol. 29 1). Such conditions are estimated to be achievable in a CTF with R0 = 1.2 m, A = 1.5, elongation ~3, Ip ~ 12 MA, BT ~ 2.5 T, producing a driven fusion burn using 47 MW of combined neutral beam and RF heating power. A design concept that allows straight-line access via remote handling to all activated fusion core components is developed and presented. The ST CTF will test the lifetime of single-turn, copper alloy centre leg for the toroidal field coil without an induction solenoid and neutron shielding and require physics data on solenoid-free plasma current initiation, ramp-up to and sustainment at multiple megaampere level. A systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of relatively low cost CTF devices to suit a range of fusion engineering and technology test missions.

https://iopscience.iop.org/article/10.1088/0741-3335/47/12B/S20/pdf

A component test facility based on the spherical tokamak

A comparative analysis of the fundamental properties of fluctuations in the vicinity of boundaries in fusion plasmas and in plasmas of magnetospheric turbulent boundary layers (TBLs) shows the similarity of their basic statistical characteristics, including the scaling of the structure functions and mutifractal parameters. Important features observed include intermittent fluctuations and anomalous mass and momentum transport, due to sporadic plasma flow injections with large flow amplitudes occuring with a much higher probability than predicted for classical Gaussian diffusion. Turbulence in edge fusion plasmas and in TBLs exhibits general self-similarity in a wide range of scales extending to the dissipation scale. Experimental scalings obtained for plasma TBLs are compared with neutral fluid results, revealing the universal properties of developed turbulence. TBL scalings are described within the log-Poisson model, which takes quasi-one-dimensional dissipative structures into account. The time () dependence of the mean-square displacement obtained from the experimental parameters of the log-Poisson distribution takes the form with 1.2 – 1.8 and indicates the presence of superdiffusion in the TBLs studied. Determining the nature of the generalized diffusion process from available regular data is a necessary step toward the quantitative description of TBL transport.

https://iopscience.iop.org/article/10.3367/UFNe.0181.201109a.0905/pdf

Investigation of intermittency and generalized self-similarity of turbulent boundary layers in laboratory and magnetospheric plasmas: towards a quantitative definition of plasma transport features

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 Zeff 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.

/pdf/observations-concerning-the-injection-of-a-lithium-aerosol-1hpbvqynpl.pdf

Observations concerning the injection of a lithium aerosol into the edge of TFTR discharges

Saturated internal kink modes have been observed in many of the highest toroidal β discharges of the National Spherical Torus Experiment (NSTX). These modes often cause rotation flattening in the plasma core, can degrade fast-ion confinement and in some cases contribute to the complete loss of plasma angular momentum and stored energy. Characteristics of the modes are measured using soft x-ray, kinetic profile and magnetic diagnostics. Toroidal flows approaching Alfvenic speeds, island pressure peaking and enhanced viscous and diamagnetic effects associated with high-β may contribute to mode nonlinear stabilization. These saturation mechanisms are investigated for NSTX parameters and compared with experimental data.

/pdf/internal-kink-mode-dynamics-in-high-b-nstx-plasmas-4qxqkhwaim.pdf

Internal kink mode dynamics in high-β NSTX plasmas

The National Spherical Torus Experiment (NSTX) multipoint Thomson scattering system has been in operation for nearly two years and provides routine Te(R,t) and ne(R,t) measurements. The laser beams from two 30 Hz Nd:YAG lasers are imaged by a spherical mirror onto 36 fiber-optic bundles. In the present configuration, the output ends of 20 of these bundles are instrumented with filter polychromators and avalanche photodiode detectors. In this article, we discuss the laser implementation and the installed collection optics. We follow with examples of raw and analyzed data, and close with some comments about calibration.

/pdf/operation-of-the-nstx-thomson-scattering-system-55hhns9ouc.pdf

Operation of the NSTX Thomson scattering system

A new spectroscopic diagnostic was developed to measure poloidal velocity profiles of Tokamak fusion test reactor (TFTR) plasmas. Carbon poloidal velocities were measured using the Doppler shift of the C VI 5291 A impurity line of both intrinsic emission and charge exchange emission from neutral beams. Poloidal velocities are typically small (vθ⩽104 m/s) requiring small wavelength shifts (Δλ⩽0.2 A) to be measured. However, the high central ion temperatures in TFTR required the use of a low dispersion spectrometer to view the entire linewidth (full width at half maximum ⩽25 A). A very high throughput spectrometer/detector system was assembled to achieve the necessary precision in vθ. Statistical errors in the chord-averaged poloidal velocity less than 100 m/s have been obtained. The short focal length spectrometer features f/1.8 input optics, a transmission grating, and refractive optics. A thinned back-illuminated charge coupled device detector provided a high quantum efficiency (QE=75%). The diagnostic h...

Tokamak fusion test reactor poloidal rotation diagnostic (invited)

A new diagnostic for the National Spherical Torus Experiment is described whose function is to measure ion rotation and temperature at the plasma edge. The diagnostic is sensitive to C III, C IV, and He II intrinsic emission, covering a radial region of 15 cm at the extreme edge of the outboard midplane. Thirteen chords are distributed between toroidal and poloidal views, allowing the toroidal and poloidal rotation and temperature of the plasma edge to be simultaneously measured with 10 ms resolution. Combined with the local pressure gradient and EFIT reconstructed magnetic field profile, the edge flow gives a measure of the local radial electric field.

Edge rotation and temperature diagnostic on the National Spherical Torus Experiment

A new diagnostic for the National Spherical Torus Experiment (NSTX) is described whose function is to measure ion rotation and temperature at the plasma edge. The diagnostic is sensitive to C III, C IV, and He II intrinsic emission, covering a radial region of 15 cm at the extreme edge of the outboard midplane. Thirteen chords are distributed between toroidal and poloidal views, allowing the toroidal and poloidal rotation and temperature of the plasma edge to be simultaneously measured with 10 ms resolution. Combined with the local pressure gradient and the EFIT code reconstructed magnetic field profile, the edge flow gives a measure of the local radial electric field.

/pdf/an-edge-rotation-and-temperature-diagnostic-on-nstx-1dp1cgknd9.pdf

R. W. Palladino

Papers

Operation of the NSTX Thomson scattering system

Tokamak fusion test reactor poloidal rotation diagnostic (invited)

Edge rotation and temperature diagnostic on the National Spherical Torus Experiment

An Edge Rotation and Temperature Diagnostic on NSTX