Wall stabilization of high beta plasmas in DIII-D (original) (raw)
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Resistive wall stabilization of high-beta plasmas in DIII–D
Nuclear Fusion, 2003
Recent DIII-D experiments show that ideal kink modes can be stabilized at high beta by a resistive wall, with sufficient plasma rotation. However, the resonant response by a marginally stable resistive wall mode to static magnetic field asymmetries can lead to strong damping of the rotation. Careful reduction of such asymmetries has allowed plasmas with beta well above the ideal MHD nowall limit, and approaching the ideal-wall limit, to be sustained for durations exceeding one second. Feedback control can improve plasma stability by direct stabilization of the resistive wall mode or by reducing magnetic field asymmetry. Assisted by plasma rotation, direct feedback control of resistive wall modes with growth rates more than 5 times faster than the characteristic wall time has been observed. These results open a new regime of tokamak operation above the free-boundary stability limit, accessible by a combination of plasma rotation and feedback control.
Progress Toward Fully Noninductive, High Beta Discharges in DIII–D
2003
Advanced Tokamak (AT) research in DIII-D focuses on developing a scientific basis for steady-state, high performance operation. For optimal performance, these experiments routinely operate with β above the n=1 no-wall limit, enabled by active feedback control. The ideal wall β limit is optimized by modifying the plasma shape, current and pressure profile. Present DIII-D AT experiments operate with f BS ≈50%-60%, with a longterm goal of ~90%. Additional current is provided by neutral beam and electron cyclotron current drive, the latter being localized well away from the magnetic axis (ρ≈0.4-0.5). Guided by integrated modeling, recent experiments have produced discharges with β≈3%, β Ν ≈3, f BS ≈55% and noninductive fraction f NI ≈90%. Additional control is anticipated using fast wave current drive to control the central current density.
Long pulse high performance discharges in the DIII-D tokamak
Nuclear Fusion, 2001
Significant progress in obtaining high performance discharges for many energy confinement times in the DIII-D tokamak has been realized since the previous IAEA meeting. In relation to previous discharges, normalized performance ~10 has been sustained for >5 τ E with q min >1.5. (The normalized performance is measured by the product β N H 89 indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H-mode discharges have an ELMing edge and β < 5%. The limit to increasing β is a resistive wall mode, rather than the tearing modes previously observed. Confinement remains good despite the increase in q. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility in the next two years. Measurement of the current density and loop voltage profiles indicate ~75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H-mode discharges with β N H 89 ~ 7 for up to 6.3 s or ~34 τ E . These discharges appear to be in resistive equilibrium with q min ~ 1.05, in agreement with the current profile relaxation time of 1.8 s.
Nuclear Fusion, 2001
The dependence of edge stability on plasma shape and local pressure gradients, P′, in the DIII-D and JT-60U tokamaks is studied. The stronger plasma shaping in DIII-D allows the edge region of DIII-D discharges with Type I ("giant") ELMs to have access to the second region of stability for ideal ballooning modes and larger edge P′ than JT-60U Type I ELM discharges. These JT-60U discharges are near the ballooning mode first regime stability limit. DIII-D results support an ideal stability based working model of Type I ELMs as low to intermediate toroidal mode number, n, MHD modes. Results from stability analysis of JT-60U Type I ELM discharges indicate that predictions from this model are also consistent with JT-60U edge stability observations.
Recent results from DIII-D and their implications for next generation tokamaks
Plasma Physics and Controlled Fusion, 1990
Recent results from the DIII-D tokamak have provided significant contributions to the understanding of many of the elements 0: 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. k egimes of confinement (H-mode) 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 found to be 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. have been establis h ed that scale favorably to proposed next generation devices, and experiments 869 870
Beta limits in long-pulse tokamak discharges
Physics of Plasmas, 1997
The maximum normalized beta achieved in long-pulse tokamak discharges at low collisionality falls significantly below both that observed in short pulse discharges and that predicted by the ideal MHD theory. Recent long-pulse experiments, in particular those simulating the International Thermonuclear Experimental Reactor ͑ITER͒ ͓M. Rosenbluth et al., Plasma Physics and Controlled Nuclear Fusion ͑International Atomic Energy Agency, Vienna, 1995͒, Vol. 2, p. 517͔ scenarios with low collisionality e * , are often limited by low-m/n nonideal magnetohydrodynamic ͑MHD͒ modes. The effect of saturated MHD modes is a reduction of the confinement time by 10%-20%, depending on the island size and location, and can lead to a disruption. Recent theories on neoclassical destabilization of tearing modes, including the effects of a perturbed helical bootstrap current, are successful in explaining the qualitative behavior of the resistive modes and recent results are consistent with the size of the saturated islands. Also, a strong correlation is observed between the onset of these low-m/n modes with sawteeth, edge localized modes ͑ELM͒, or fishbone events, consistent with the seed island required by the theory. We will focus on a quantitative comparison between both the conventional resistive and neoclassical theories, and the experimental results of several machines, which have all observed these low-m/n nonideal modes. This enables us to single out the key issues in projecting the long-pulse beta limits of ITER-size tokamaks and also to discuss possible plasma control methods that can increase the soft  limit, decrease the seed perturbations, and/or diminish the effects on confinement.
Control of the resistive wall mode in advanced tokamak plasmas on DIII-D
Nuclear Fusion, 2000
Resistive wall mode (RWM) instabilities are found to be a limiting factor in advanced tokamak (AT) regimes with low internal inductance. Even small amplitude modes can affect the rotation profile and the performance of these ELMing H-mode discharges. Although complete stabilization of the RWM by plasma rotation has not yet been observed, several discharges with increased beam momentum and power injection sustained good steady-state performance for record time extents. The first investigation of active feedback control of the RWM has shown promising results: the leakage of the radial magnetic flux through the resistive wall can be successfully controlled, and the duration of the high beta phase can be prolonged. The results provide a comparative test of several approaches to active feedback control, and are being used to benchmark the analysis and computational models of active control.
Measurement of resistive wall mode stability in rotating high-β DIII-D plasmas
Nuclear Fusion, 2005
Toroidal plasma rotation in the order of a few percent of the Alfvén velocity can stabilize the resistive wall mode and extend the operating regime of tokamaks from the conventional, ideal MHD no-wall limit up to the ideal MHD ideal wall limit. The stabilizing effect has been measured passively by measuring the critical plasma rotation required for stability and actively by probing the plasma with externally applied resonant magnetic fields. These measurements are compared to predictions of rotational stabilization of the sound wave damping and of the kinetic damping model using the MARS code.
Confinement physics of H-mode discharges in DIII-D
Plasma Physics and Controlled Fusion, 1989
Our 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. We 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.