S.M. Kaye

Scalings for tokamak energy confinement

On the basis of an analysis of the ITER L-mode energy confinement database, two new scaling expressions for tokamak L-mode energy confinement are proposed, namely a power law scaling and an offset-linear scaling. The analysis indicates that the present multiplicity of scaling expressions for the energy confinement time τE in tokamaks (Goldston, Kaye, Odajima-Shimomura, Rebut-Lallia, etc.) is due both to the lack of variation of a key parameter combination in the database, fs = 0.32 R a−0.75 k0.5 ~ A a0.25k0.5, and to variations in the dependence of τE on the physical parameters among the different tokamaks in the database. By combining multiples of fs and another factor, fq = 1.56 a2 kB/RIp = qeng/3.2, which partially reflects the tokamak to tokamak variation of the dependence of τE on q and therefore implicitly the dependence of τE on Ip and ne, the two proposed confinement scaling expressions can be transformed to forms very close to most of the common scaling expressions. To reduce the multiplicity of the scalings for energy confinement, the database must be improved by adding new data with significant variations in fs, and the physical reasons for the tokamak to tokamak variation of some of the dependences of the energy confinement time on tokamak parameters must be clarified.

A new look at density limits in tokamaks

While the results of early work on the density limit in tokamaks from the ORMAK and DITE groups have been useful over the years, results from recent experiments and the requirements for extrapolation to future experiments have prompted a new look at this subject. There are many physical processes which limit the attainable densities in tokamak plasmas. These processes include: (1) radiation from low Z impurities, convection, charge exchange and other losses at the plasma edge; (2) radiation from low or high Z impurities in the plasma core; (3) deterioration of particle confinement in the plasma core; and (4) inadequate fuelling, often exacerbated by strong pumping by walls, limiters or divertors. Depending upon the circumstances, any of these processes may dominate and determine a density limit. In general, these mechanisms do not show the same dependence on plasma parameters. The multiplicity of processes leading to density limits with a variety of scaling has led to some confusion when comparing density limits for different machines. The authors attempt to sort out the various limits and to extend the scaling law for one of them to include the important effects of plasma shaping, i.e. ;e = k, where ne is the line average electron density (1020 m−3), κ is the plasma elongation and (MAm−2) is the average plasma current density, defined as the total current divided by the plasma cross-sectional area. In a sense, this is the most important density limit since, together with the q-limit, it yields the maximum operating density for a tokamak plasma. It is shown that this limit may be caused by a dramatic deterioration in core particle confinement occurring as the density limit boundary is approached. This mechanism can help explain the disruptions and Marfes that are associated with the density limit.

Exploration of Spherical Torus Physics in the NSTX Device

The National Spherical Torus Experiment (NSTX) is being built at the Princeton Plasma Physics Laboratory to test the fusion physics principles for the Spherical Torus (ST) concept at the MA level. The NSTX nominal plasma parameters are R {sub 0} = 85 cm, a = 67 cm, R/a greater than or equal to 1.26, B {sub T} = 3 kG, I {sub p} = 1 MA, q {sub 95} = 14, elongation {kappa} less than or equal to 2.2, triangularity {delta} less than or equal to 0.5, and plasma pulse length of up to 5 sec. The plasma heating/current drive (CD) tools are High Harmonic Fast Wave (HHFW) (6 MW, 5 sec), Neutral Beam Injection (NBI) (5 MW, 80 keV, 5 sec), and Coaxial Helicity Injection (CHI). Theoretical calculations predict that NSTX should provide exciting possibilities for exploring a number of important new physics regimes including very high plasma beta, naturally high plasma elongation, high bootstrap current fraction, absolute magnetic well, and high pressure driven sheared flow. In addition, the NSTX program plans to explore fully noninductive plasma start-up, as well as a dispersive scrape-off layer for heat and particle flux handling.

Equilibrium properties of spherical torus plasmas in NSTX

Research in NSTX has been conducted to establish spherical torus plasmas to be used for high ?, auxiliary heated experiments. This device has a major radius R0 = 0.86?m and a midplane halfwidth of 0.7?m. It has been operated with toroidal magnetic field B0 ? 0.3?T and Ip ? 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 (?t = 17.8%) with neutral beam heating. A plasma elongation in the range 1.6 ? ? ? 2.0 and a triangularity in the range 0.25 ? ? ? 0.45 have been sustained, with values of ? = 2.6 and ? = 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 Ip, ? and li evolutions, have been observed. A low q limit was observed and is imposed by a low toroidal mode number kink instability.

Plasma response to lithium-coated plasma-facing components in the National Spherical Torus Experiment

Experiments in the National Spherical Torus Experiment (NSTX) have shown beneficial effects on the performance of divertor plasmas as a result of applying lithium coatings on the graphite and carbon-fiber-composite plasma-facing components. These coatings have mostly been applied by a pair of lithium evaporators mounted at the top of the vacuum vessel which inject collimated streams of lithium vapor toward the lower divertor. In neutral beam injection (NBI)-heated deuterium H-mode plasmas run immediately after the application of lithium, performance modifications included decreases in the plasma density, particularly in the edge, and inductive flux consumption, and increases in the electron and ion temperatures and the energy confinement time. Reductions in the number and amplitude of edge-localized modes (ELMs) were observed, including complete ELM suppression for periods of up to 1.2 s, apparently as a result of altering the stability of the edge. However, in the plasmas where ELMs were suppressed, there was a significant secular increase in the effective ion charge Zeff and the radiated power as a result of increases in the carbon and medium-Z metallic impurities, although not of lithium itself which remained at a very low level in the plasma core, <0.1%. The impurity buildup could be inhibited by repetitively triggering ELMs with the application of brief pulses of an n = 3 radial field perturbation. The reduction in the edge density by lithium also inhibited parasitic losses through the scrape-off-layer of ICRF power coupled to the plasma, enabling the waves to heat electrons in the core of H-mode plasmas produced by NBI. Lithium has also been introduced by injecting a stream of chemically stabilized, fine lithium powder directly into the scrape-off-layer of NBI-heated plasmas. The lithium was ionized in the SOL and appeared to flow along the magnetic field to the divertor plates. This method of coating produced similar effects to the evaporated lithium but at lower amounts.

Numerical study of tokamak equilibria with arbitrary flow

The effects of toroidal and poloidal flows on the equilibrium of tokamak plasmas are numerically investigated using the code FLOW. The code is used to determine the changes in the profiles induced by large toroidal flows on NSTX-like equilibria [with NSTX being the National Spherical Torus Experiment, M. Ono, S.M. Kaye, Y.-K.M. Peng et al., Nucl. Fusion 40, 557 (2000)] where flows exceeding the sound speed lead to a considerable outward shift of the plasma. The code is also used to study the effects of poloidal flow when the flow velocity profile varies from subsonic to supersonic with respect to the poloidal sound speed. It is found that pressure and density profiles develop a pedestal structure characterized by radial discontinuities at the transonic surface where the poloidal velocity abruptly jumps from subsonic to supersonic values. These results confirm the conclusions of the analytic theory of R. Betti and J. P. Freidberg [Phys. Plasmas 7, 2439 (2000)], derived for a low-β, large aspect ratio tokam...

Edge turbulence measurements in NSTX by gas puff imaging

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...

Physics design of the national spherical torus experiment

The mission of the National Spherical Torus Experiment (NSTX) is to prove the principles of spherical torus physics by producing high-beta toroidal plasmas that are non-inductively sustained, and whose current profiles are in steady-state. NSTX will be one of the first ultra low a[P(input) up to 11 MW] in order to produce high-beta toroidal (25 to 40%), low collisionality, high bootstrap fraction (less than or equal to 70%) discharges. Both radio-frequency (RF) and neutral-beam (NB) heating and current drive will be employed. Built into NSTX is sufficient configurational flexibility to study a range of operating space and the resulting dependences of the confinement, micro- and MHD stability, and particle and power handling properties. NSTX research will be carried out by a nationally based science team.

Resistive wall stabilized operation in rotating high beta NSTX plasmas

The National Spherical Torus Experiment (NSTX) has demonstrated the advantages of low aspect ratio geometry in accessing high toroidal and normalized plasma beta, and βN ≡ 10 8〈βt〉 aB0/Ip. Experiments have reached βt = 39% and βN = 7.2 through boundary and profile optimization. High βN plasmas can exceed the ideal no-wall stability limit, βNno-wall, for periods much greater than the wall eddy current decay time. Resistive wall mode (RWM) physics is studied to understand mode stabilization in these plasmas. The toroidal mode spectrum of unstable RWMs has been measured with mode number n up to 3. The critical rotation frequency of Bondeson-Chu, Ωcrit = ωA/(4q2), describes well the RWM stability of NSTX plasmas when applied over the entire rotation profile and in conjunction with the ideal stability criterion. Rotation damping and global rotation collapse observed in plasmas exceeding βNno-wall differs from the damping observed during tearing mode activity and can be described qualitatively by drag due to neoclassical toroidal viscosity in the helically perturbed field of an ideal displacement. Resonant field amplification of an applied n = 1 field perturbation has been measured and increases with increasing βN. Equilibria are reconstructed including measured ion and electron pressure, toroidal rotation and flux isotherm constraint in plasmas with core rotation ω/ωA up to 0.48. Peak pressure shifts of 18% of the minor radius from the magnetic axis have been reconstructed.

Initial results from coaxial helicity injection experiments in NSTX

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.