T.C. Luce

Principal physics developments evaluated in the ITER design review

As part of the ITER Design Review and in response to the issues identified by the Science and Technology Advisory Committee, the ITER physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, material choice and heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation.

Low voltage Ohmic and electron cyclotron heating assisted startup in DIII-D

There is considerable interest in the development of low voltage startup scenarios for large tokamaks since it is proposed that in ITER the electric field which will be applied for ionization and plasma current ramp-up will be limited to values of E ≤ 0.3 V/m. Studies of low voltage startup have been carried out in DIII-D with and without electron cyclotron preionization and preheating. Successful Ohmic startup has been achieved with E ~ 0.25 V/m by paying careful attention to error fields and prefill pressure, while electron cyclotron heating (ECH) assisted startup with E ~ 0.15 V/m has been demonstrated. ECH assisted startup gives improved reliability at such low electric fields and permits operation over an extended range of prefill pressures and error magnetic fields. Using ECH, startup at E = 0.3 V/m with |B⊥| > 50 G over most of the vessel cross-section has been demonstrated. Such an error field represents an increase by more than a factor of two over the highest value for which Ohmic startup was achieved at the same electric field. During low voltage Ohmic startup with extreme values of prefill pressure and/or error magnetic fields, excessive breakdown delays are observed. The experimental data agree well with theoretical predictions based on the Townsend avalanche theory. ECH assisted startup is always prompt. The primary effect of ECH during the plasma current ramp-up is a decrease of the resistive component of the loop voltage Vrcs. A significant reduction (~30%) in Vres is achieved for low ECH powers (PRF ~ 300-400 kW), but a further large increase in PRF results in only a modest additional decrease in Vres. ECH was not applied over the whole ramp-up phase in these experiments and produced a reduction in volt-second consumption up to the current flat-top (Ip ~ 1 MA) of 10%. These experiments confirm that the low electric fields specified in the ITER design are acceptable and demonstrate the substantial benefits which accrue from the use of ECH assisted startup.

Quiescent double barrier high-confinement mode plasmas in the DIII-D tokamak

High-confinement (H-mode) operation is the choice for next-step tokamak devices based either on conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the beta limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] this year have demonstrated a new operating regime, the quiescent H-mode regime, which solves these problems. We have achieved quiescent H-mode operation that is ELM-free and yet has good density and impurity control. In addition, we have demonstrated that an internal transport barrier can be produced and maintained inside the H-mode edge barrier for long periods of time (>3.5 s or >25 en...

Applications of high-power millimeter waves in fusion energy research

High-power millimeter wave sources are a key enabling technology in fusion energy research. The present state of the art of application of these sources to the areas of heating, current generation, and scattering for diagnostic purposes in fusion plasmas is reviewed. The extrapolation of these applications to future devices and the requirements which they place on sources and transmission lines are also discussed.

Non-dimensional scaling of turbulence characteristics and turbulent diffusivity

Plasma turbulence characteristics, including radial correlation lengths, decorrelation times, amplitude profile and flow velocity, have been measured during a ρ* scan on DIII-D while all other transport relevant dimensionless quantities (e.g., β, ν*, κ, q, Te/Ti) are held nearly constant. The turbulence is measured by examining the correlation properties of the local long wavelength (k⊥ρi ≤ 1) density fluctuations, measured with beam emission spectroscopy. The radial correlation length of the turbulence Lc,r is shown to scale with the local ion gyroradius, Lc,r ≈ 5ρi, while the decorrelation times scale with the local acoustic velocity as τc~a/cs. The turbulent diffusivity parameter, D~(Lc,r2/τc), scales in a roughly gyro-Bohm-like fashion, as predicted by the gyrokinetic equations governing turbulent transport. The experimental one fluid power balance heat diffusivity scaling and that from GLF23 modelling compare reasonably well.

QUIESCENT H-MODE PLASMAS IN THE DIII-D TOKAMAK

H-mode operation is the choice for next-step tokamak devices based either on conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge-localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the β limit and reduced core transport regions needed for advanced tokamak operation. Recent experimental results from DIII-D have demonstrated a new operating regime, the quiescent H-mode regime, which solves these problems. We have achieved quiescent H-mode operation which is ELM-free and yet has good density control. In addition, we have demonstrated that an internal transport barrier can be produced and maintained inside the H-mode edge barrier for long periods of time (>3.5 s or >25 energy confinement times τE). By forming the core barrier and then stepping up the input power, we have achieved βNH89 = 7 for up to 10 times the τE of 160 ms. The βNH89 values of 7 substantially exceed the value of 4 routinely achieved in standard ELMing \mbox{H-mode.} The key factors in creating the quiescent H-mode operation are neutral beam injection in the direction opposite to the plasma current (counter injection) plus cryopumping to reduce the density. Density control in the quiescent H-mode is possible because of the presence of an edge MHD oscillation, the edge harmonic oscillation, which enhances the edge particle transport while leaving the energy transport unaffected.

Design and simulation of extremum-seeking open-loop optimal control of current profile in the DIII-D tokamak

In a magnetic fusion reactor, the achievement of a certain type of plasma current profiles, which are compatible with magnetohydrodynamic stability at high plasma pressure, is key to enable high fusion gain and non-inductive sustainment of the plasma current for steady-state operation. The approach taken toward establishing such plasma current profiles at the DIII-D tokamak is to create the desired profile during the plasma current ramp-up and early flattop phases. The evolution in time of the current profile is related to the evolution of the poloidal flux, which is modeled in normalized cylindrical coordinates using a partial differential equation usually referred to as the magnetic diffusion equation. The control problem is formulated as an open-loop, finite-time, optimal control problem for a nonlinear distributed parameter system, and is approached using extremum seeking. Simulation results, which demonstrate the accuracy of the considered model and the efficiency of the proposed controller, are presented.

Nondimensional transport scaling in DIII‐D: Bohm versus gyro‐Bohm resolved

The scaling of cross‐field heat transport with relative gyroradius ρ* was measured in low (L) and high (H) mode tokamak plasmas using the technique of dimensionally similar discharges. The relative gyroradius scalings of the electron and ion thermal diffusivities were determined separately using a two‐fluid transport analysis. For L‐mode plasmas, the electron diffusivity scaled as χe∝χBρ1.1±0.3* (gyro‐Bohm‐like) while the ion diffusivity scaled as χi∝χBρ−0.5±0.3* (worse than Bohm‐like). The results were independent of the method of auxiliary heating (radio frequency or neutral beam). Since the electron and ion fluids had different gyroradius scalings, the effective diffusivity and global confinement time scalings were found to vary from gyro‐Bohm‐like to Bohm‐like depending upon whether the electron or ion channel dominated the heat flux. This last property can explain the previously disparate results with dimensionally similar discharges on different fusion experiments that have been published. Experimen...

Progress toward fully noninductive, high beta conditions in DIII-D

The DIII-D Advanced Tokamak (AT) program in the DIII-D tokamak [J. L. Luxon, Plasma Physics and Controlled Fusion Research, 1986, Vol. I (International Atomic Energy Agency, Vienna, 1987), p. 159] is aimed at developing a scientific basis for steady-state, high-performance operation in future devices. This requires simultaneously achieving 100% noninductive operation with high self-driven bootstrap current fraction and toroidal beta. Recent progress in this area includes demonstration of 100% noninductive conditions with toroidal beta, βT=3.6%, normalized beta, βN=3.5, and confinement factor, H89=2.4 with the plasma current driven completely by bootstrap, neutral beam current drive, and electron cyclotron current drive (ECCD). The equilibrium reconstructions indicate that the noninductive current profile is well aligned, with little inductively driven current remaining anywhere in the plasma. The current balance calculation improved with beam ion redistribution that was supported by recent fast ion diagno...

Active control for stabilization of neoclassical tearing modes

This work describes active control algorithms used by DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] to stabilize and maintain suppression of 3/2 or 2/1 neoclassical tearing modes (NTMs) by application of electron cyclotron current drive (ECCD) at the rational q surface. The DIII-D NTM control system can determine the correct q-surface/ECCD alignment and stabilize existing modes within 100–500ms of activation, or prevent mode growth with preemptive application of ECCD, in both cases enabling stable operation at normalized beta values above 3.5. Because NTMs can limit performance or cause plasma-terminating disruptions in tokamaks, their stabilization is essential to the high performance operation of ITER [R. Aymar et al., ITER Joint Central Team, ITER Home Teams, Nucl. Fusion 41, 1301 (2001)]. The DIII-D NTM control system has demonstrated many elements of an eventual ITER solution, including general algorithms for robust detection of q-surface/ECCD alignment and for real-time maintenance of alignment ...