Improved Plasma Performance Research Articles

Progress towards steady state at low aspect ratio on the National Spherical Torus Experiment (NSTX)

Modifications to the plasma control capabilities and poloidal field coils of the National Spherical Torus Experiment (NSTX) have enabled a significant enhancement in shaping capability which has led to the transient achievement of a record shape factor (S ≡ q95 (Ip/a Bt)) of ∼41 (MA m−1 T−1) simultaneous with a record plasma elongation of κ ≡ b/a ∼ 3. This result was obtained using isoflux control and real-time equilibrium reconstruction. Achieving high shape factor together with tolerable divertor loading is an important result for future ST burning plasma experiments as exemplified by studies for future ST reactor concepts, as well as neutron producing devices, which rely on achieving high shape factors in order to achieve steady state operation while maintaining MHD stability. Statistical evidence is presented which demonstrates the expected correlation between increased shaping and improved plasma performance. Plasmas with high shape factor have been sustained for pulse lengths which correspond to τpulse = 1.6s ∼ 50 τE ∼ 5 τCR, where τCR is the current relaxation time and τE is the energy confinement time. Plasmas with higher βt ∼ 20% have been sustained for τpulse = 1.2s ∼ 25 τE ∼ 3 τCR with non-inductive current fractions fNI ∼ 50%, with ∼40% pressure driven current and ∼10% neutral beam driven current. An interesting feature of these discharges is the observation that the central value of the safety factor q(0) remains elevated and constant for several current diffusion times without sawteeth, similar to the ‘hybrid mode’. Calculations of the profiles of inductive and non-inductive current are compared with measurements of the total current profile and are shown to be in quantitative agreement. Results are shown from experiments which investigate the applicability of high harmonic fast waves (HHFWs) and electron Bernstein waves (EBWs) as current drive and heating sources, and the possibility of LHCD for future ST devices is raised. A calculated scenario which provides 100% non-inductive current drive is described. NSTX operates with peak divertor heat fluxes which are in the same range as those expected for the ITER device, i.e. with . High triangularity, high elongation plasmas on NSTX have been demonstrated to have reduced peak heat flux to the divertor plates to <3 MW m−2.

Active MHD control at high currents in RFX-mod

The modified RFX is a very flexible device used for a variety of control schemes for MHD instabilities and for advanced reverse field pinch scenarios. Relative to the previous machine, RFX-mod has a thin Cu shell with vertical field penetration time τS, lowered from 450 to 50 ms and shell/plasma proximity from b/a = 1.24–1.1. Toroidal equilibrium is feedback-controlled and new power supplies provide better Bϕ control. Newly designed graphite tiles protect the vessel from localized power deposition. A mesh of 192 external saddle coils, supervised by a digital feedback system, controls radial fields due to field errors and MHD modes. The paper presents an overview of the very encouraging results obtained using both new and ‘standard’ advanced operational modes in the current range 0.3–1 MA. A dramatic improvement of plasma performance was obtained by using the saddle coils to cancel all the radial field components, an operation mode dubbed virtual shell (VS). The toroidal voltage was lowered by more than 25% and the pulse length was tripled, up to 7 times the τS. Steady-state RFP pulses are now limited only by the applied volt-seconds. The improved magnetic boundary also has an effect on the tearing modes underlying the sustainment of the RFP, whose core amplitude is more than halved. The VS combined with new schemes for the active rotation of the MHD dynamo modes has allowed us to obtain reliable and well-controlled long RFP pulses in the MA current range. This results in a 100% increase in the particle and energy confinement time relative to the previous experiment and opens the possibility of exploring the machine performance in the 2 MA design range.

Power and Particle Handling and Wall Conditioning in NCSX

Plasma boundary control in stellarators has been shown to be very effective in improving plasma performance and, accordingly, is an important element from the very beginning of the National Compact Stellarator Experiment (NCSX) design. Studies of the magnetic field topology outside the last closed magnetic surface (LCMS) indicate the possibility of many toroidal revolutions of field lines launched within a couple of centimeters of the LCMS. Field line connection lengths, typically in the order of 100 m, should be sufficient to allow for the necessary separation of divertor and separatrix temperatures. In the top and bottom of the bean-shaped cross section (toroidal angle π = 0), a field expansion of >5 is observed, which will help to spread out the heat flux on limiters and divertor plates. Plasma-facing components (PFCs) will be developed systematically according to our respective understanding of the NCSX boundary; the phased PFC development will start out with a set of limiters and has the eventual goal to develop a divertor with all the benefits of impurity and neutrals control. Neutrals calculations have been started to investigate the effect of neutrals penetration at various plasma cross sections, especially at the location of π = 0 deg. Advanced wall conditioning techniques, as employed in other major fusion devices, will be incorporated in the NCSX operation.

Reduced divertor heat loads, plasma shape effects, and radial electric field structures in JFT-2M HRS H-mode plasmas

A key feature of the ‘high recycling steady’ (HRS) H-mode regimes is the reduction of the transient heat load to the divertor target due to large ELMs. The heat flux to the divertor targets measured by means of a Langmuir probe in the scrape-off layer on the target plate is about ∼0.3 MW m−2 in the typical HRS plasmas, which is very small in comparison with the transient heat load in the ELMy H-mode discharge. In addition, the effects of the plasma shaping on the access to and improved plasma performance of HRS H-mode regimes have been investigated. Furthermore, simultaneous measurements of both density fluctuations and potential profiles at the plasma edge region using a heavy ion beam probe are also presented.

Real-time determination of plasma confinement parameters in HL-2A

During the plasma discharge in HL-2A, the magnetic field at the plasma boundary is computed using currents filaments code. The Shafranov integrals S1，S2 and S3 are deduced; the poloidal beta, internal inductance, the horizontal position of the magnetic axis Rm and the plasma energy Wdia are calculated; the current center of plasma (Rc,Zc)is computed with the current moments. The real time determination of these quantities is of significance for understanding and adjusting the HL-2A discharge in time, and for improving plasma performance.

Modelling of minority ion cyclotron current drive during the activated phase of ITER

Neoclassical tearing modes, triggered by the long-period sawteeth expected in tokamaks with large non-thermal α-particle populations, may impose a severe β limit on experiments with large fusion yields and on reactors. Sawtooth destabilization by localized current drive could relax the β limit and improve plasma performance. 3He minority ion cyclotron current drive around the sawtooth inversion radius has been planned for ITER. Several ion species, including beam injected D ions and fusion born α particles, are however also resonant in the plasma and may represent a parasitic absorption of RF power.Modelling of minority ion cyclotron current drive in an ITER-FEAT-like plasma is presented, including the effects of ion trapping, finite ion drift orbit widths, wave-induced radial transport and the coupled evolution of wave fields and resonant ion distributions. The parasitic absorption of RF power by the other resonant species is concluded to be relatively small, but the 3He minority current drive is nevertheless negligible due to the strong collisionality of the 3He ions and the drag current by toroidally counter-rotating background ions and co-rotating electrons. H minority current drive is found to be a significantly more effective alternative.

Role of locked mode in the effectiveness of pulsed poloidal current drive regime in the reversed-field pinch

The close relationship between the locked mode (LM) and pulsed poloidal current drive (PPCD) regime in a reversed-field pinch device is described. If the modes lock close to the shell gap then the LM is enhanced and there is a high probability that the PPCD will not improve plasma performance. The physical reason will be discussed. If the position of the locked mode is far from the shell gap then the PPCD is effective with a high probability. LM energy is reduced and the more the LM energy decreases, the better the plasma performance.

Development of NSTX particle control techniques

Density control in NSTX, involves primarily controlling impurity influxes and recycling. We have compared boronization on hot and cold surfaces, varying helium glow discharge conditioning (HeGDC) durations, helium discharge cleaning, brief daily boronization, and between discharge boronization to reduce and control spontaneous density rises. Access to Ohmic H-modes was enabled by boronization on hot surfaces, however, the duration of the effectiveness of hot and cold boronization was comparable. A 15min HeGDC between discharges was needed for reproducible L–H transitions. He discharge conditioning yielded slower density rises than 15min of HeGDC. Brief daily boronization followed by a comparable duration of applied HeGDC restored and enhanced good conditions. Additional brief boronizations between discharges did not improve plasma performance, if conditions were already good. Between discharge boronization required increases in the NSTX duty cycle due to the need for additional HeGDC to remove codeposited D.

Multi-mode error field correction on the DIII-D tokamak

Error field optimization on DIII-D tokamak (Luxon J.L. 2002 Nucl. Fusion 42 814) plasma discharges has routinely been done for the last ten years with the use of the external `n = 1 coil' or the `C-coil'. The optimum level of correction coil current is determined by the ability to avoid the locked mode instability and access previously unstable parameter space at low densities. The locked mode typically has toroidal and poloidal mode numbers n = 1 and m = 2, respectively, and it is this component that initially determined the correction coil current and phase. Realization of the importance of nearby n = 1 mode components m = 1 and m = 3 has led to a revision of the error field correction algorithm. Viscous and toroidal mode coupling effects suggested the need for additional terms in the expression for the radial `penetration' field Bpen that can induce a locked mode. To incorporate these effects, the low density locked mode threshold database was expanded. A database of discharges at various toroidal fields, plasma currents, and safety factors was supplemented with data from an experiment in which the fields of the n = 1 coil and C-coil were combined, allowing the poloidal mode spectrum of the error field to be varied. Using this technique to optimize the error field reveals a new and clear low density limit determined by the onset of runaway electrons, not locked modes. A multi-variate regression analysis of this new low density locked mode database was done to determine the low density locked mode threshold scaling relationship ne∝BT−0.01q95−0.79Bpen and the coefficients of the poloidal mode components in the expression for Bpen. Improved plasma performance is achieved by optimizing Bpen by varying the applied correction coil currents.

Recent results from the Large Helical Device

The most important finding in the Large Helical Device (LHD) experiments so far is that an inward-shifted configuration exhibits good plasma performance with a scaling similar to that of ELMy H-mode tokamaks. The inward-shifted configuration has been predicted to have unfavourable magnetohydrodynamic (MHD) properties, even though it has significantly better particle-orbit properties. However, no serious confinement degradation due to the MHD activities was observed, resolving favourably the potential conflict between stability and confinement. Neoclassical transport loss can be suppressed even in a low-collisionality regime, and in this way the inward-shifted configuration was shown to make the LHD plasma properties favourable. Then, it is very important to realize more improved plasma performance and higher temperature plasmas for extending the plasma-parameter regime in order to obtain data that can be extrapolated to a reactor. In the fifth campaign in 2001–2002, an increase in the heating power achieved an electron temperature Te of over 10 keV and an ion temperature Ti of 5 keV. A Te profile, which is characteristic of internal transport barriers, was also observed when the electron cyclotron resonance heating power was highly focused on the centre of the plasma sustained by neutral beam injection.

Plasma boundary considerations for the national compact stellarator experiment

The national compact stellarator experiment (NCSX) [EPS 2001, Madeira, Portugal, 18–22 June 2001] is a new fusion project located at Princeton Plasma Physics Laboratory, Princeton, NJ. Plasma boundary control in stellarators has been shown to be very effective in improving plasma performance [EPS 2001, Madeira, Portugal, 18–22 June 2001] and, accordingly, will be an important element from the very beginning of the NCSX design. Plasma-facing components will be developed systematically according to our understanding of the NCSX boundary, with the eventual goal to develop a divertor with all the benefits for impurity and neutrals control. Neutrals calculations have been started to investigate the effect of neutrals penetration at various cross-sections.

Diagnostics System of JT-60U

The diagnostic system of JT-60U (JT-60upgrade) 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 CO2 laser interferometer/polarimeter and a CO2 laser collective Thomson scattering system have been developed.

Improved plasma performance on Large Helical Device

Since the start of the Large Helical Device (LHD) experiment, various attempts have been made to achieve improved plasma performance in LHD [A. Iiyoshi et al., Nucl. Fusion 39, 1245 (1999)]. Recently, an inward-shifted configuration with a magnetic axis position Rax of 3.6 m has been found to exhibit much better plasma performance than the standard configuration with Rax of 3.75 m. A factor of 1.6 enhancement of energy confinement time was achieved over the International Stellarator Scaling 95. This configuration has been predicted to have unfavorable magnetohydrodynamic (MHD) properties, based on linear theory, even though it has significantly better particle-orbit properties, and hence lower neoclassical transport loss. However, no serious confinement degradation due to the MHD activities was observed, resolving favorably the potential conflict between stability and confinement at least up to the realized volume-averaged beta 〈β〉 of 2.4%. An improved radial profile of electron temperature was also achieved in the configuration with magnetic islands, minimized by an external perturbation coil system for the Local Island Divertor (LID). The LID has been proposed for remarkable improvement of plasma confinement like the high (H) mode in tokamaks, and the LID function was suggested in limiter experiments.

Plasma Control with Conducting Plates in the Anchor Region of the GAMMA 10 Tandem Mirror

This paper describes the effect of conducting plates installed in the transition region of the minimum-B anchor-cell of GAMMA 10 for the purpose of the improvement in plasma performance. The conducting plates are fixed closely to the plasma surface where the cross section of the plasma is flatly elongated in the transition region. Electrical characteristics of the conducting plates are investigated by changing the resister connected between the plates and the machine ground. Effects on the plasma parameters of the conducting plates are studied and it is found that the floating condition of the plates leads to the increase of the plasma density during potential formation. It is also find that the effect is reduced in the case that the space of main plates facing each other is widened. Existence of electric current in azimuthal direction of plasma cross section is discussed from the viewpoint of radial loss mechanism.

Hydrogen isotope retention in beryllium for tokamak plasma-facing applications

Beryllium has been used as a plasma-facing material to effect substantial improvements in plasma performance in the Joint European Torus (JET), and it is planned as a plasma-facing material for the first wall (FW) and other components of the International Thermonuclear Experimental Reactor (ITER). The interaction of hydrogenic ions, and charge-exchange neutral atoms from plasmas, with beryllium has been studied in recent years with widely varying interpretations of results. In this paper we review experimental data regarding hydrogenic atom inventories in experiments pertinent to tokamak applications and show that with some very plausible assumptions, the experimental data appear to exhibit rather predictable trends. A phenomenon observed in high ion-flux experiments is the saturation of the beryllium surface such that inventories of implanted particles become insensitive to increased flux and to continued implantation fluence. Methods for modeling retention and release of implanted hydrogen in beryllium are reviewed and an adaptation is suggested for modeling the saturation effects. The TMAP4 code used with these modifications has succeeded in simulating experimental data taken under saturation conditions where codes without this feature have not. That implementation also works well under more routine conditions where the conventional recombination-limited release model is applicable. Calculations of tritium inventory and permeation in the ITER FW during the basic performance phase (BPP) using both the conventional recombination model and the saturation effects assumptions show a difference of several orders of magnitude in both inventory and permeation rate to the coolant.

Experiments on Li pellet injection into Heliotron E

Li pellets of large size were injected into electron cyclotron resonance (ECR) heated plasmas and neutral beam injection (NBI) heated plasmas of Heliotron E. The discharge behaviour, pellet ablation and wall conditioning were studied. The electron pressure is doubled after injection into the NBI plasma and remains unchanged in the case of ECR heating. This may be due to the energy exchange between the electrons and thermal ions with the fast ions from the neutral beam. The observed discrepancy between the experimental and modelled ablation rates may be caused by both the plasma cooling due to pellet ablatant and the ablation stimulated by the fast ions in the NBI-heated regime and by the fast electrons in the ECR-heated regime. In preliminary experiments on wall conditioning by Li pellet injection, no improvement of plasma performance after Li pellet injection was observed in the divertor or limiter configuration, with the limiter radii .

Improvement of plasma performance with wall conditioning in the HL-1M tokamak

Studies and selection of plasma facing materials continue to be a concern for future fusion devices, and ongoingefforts are being made in the HL-1M tokamak. The advanced methods of wall modification adopted on HL-1M are surveyed.Significant improvements in tokamak plasma performance have been obtained by using boron, silicon or lithium containingsubstances as a material for wall coatings. The siliconization technique is highlighted. Lithiumization, as the newesttechnique, will be investigated further.

Improved plasma performance in tokamaks with negative magnetic shear.

A tokamak plasma configuration is reported that simultaneously improves on the maximum stable plasma pressure, the bootstrap current contribution, and kinetic stability to temperature and density gradient driven modes in toroidal geometry. It is characterized by negative magnetic shear in the plasma interior and a peaked pressure profile. Stability to the ideal low-n external kink modes requires a conducting shell at 1.3 times the plasma minor radius. This novel plasma configuration is promising for improved plasma performance in advanced tokamak experiments.

Improved plasma performance in TEXTOR with silicon coated surfaces.

Coating of the walls of TEXTOR with silicon has led to improved tokamak plasma performance. Very low concentrations of C, B, and O are measured. Radiation from silicon is located at the plasma periphery at r/a\ensuremath{\ge}0.75 and decreases with increasing plasma density. Density limits are enhanced by 30% as compared to boronized conditions; large density gradients and low electron temperatures at the edge (10 eV) are obtained. The improved confinement regimes observed earlier at low densities [${\mathrm{\ensuremath{\tau}}}_{\mathit{E}}$=1.7\ifmmode\times\else\texttimes\fi{}ITER L 89-P International Thermonuclear Experimental Reactor)] were extended to significantly higher densities in spite of a high radiative power fraction.

A programmatic framework for the Tokamak Physics Experiment (TPX)

Significant advances have been made in the confinement of reactor-grade plasmas, so that we are now preparing for experiments at the “power breakeven” level in the JET and TFTR experiments. In ITER we will extend the performance of tokamaks into the burning plasma regime, develop the technology of fusion reactors, and produce over a gigawatt of fusion power. Besides taking these crucial steps toward the technical feasibility of fusion, we must also take steps to ensure its economic acceptability. The broad requirements for economically attractive tokamak reactors based on physics advancements have been set forth in a number of studies. An advanced physics data base is emerging from a physics program of concept improvement using existing tokamaks around the world. This concept improvements program is emerging as the primary focus of the U.S. domestic tokamak program, and a key element of that program is the proposed Tokamak Physics Experiment (TPX). With TPX we can develop the scientific data base for compact, continuously-operating fusion reactors, using advanced steady-state control techniques to improve plasma performance. We can develop operating techniques needed to ensure the success of ITER and provide first-time experience with several key fusion reactor technologies. This paper explains the relationships of TPX to the current U.S. fusion physics program, to the ITER program, and to the development of an attractive tokamak demonstration plant for this next stage in the fusion program.