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.

The development of techniques for neoclassical tearing mode (NTM) suppression or avoidance is crucial for successful high beta/high confinement tokamaks. Neoclassical tearing modes are islands destabilized and maintained by a helically perturbed bootstrap current and represent a significant limit to performance at higher poloidal beta. The confinement-degrading islands can be reduced or completely suppressed by precisely replacing the "missing" bootstrap current in the island O-point or by interfering with the fundamental helical harmonic of the pressure. Implementation of such techniques is being studied in the DIII-D tokamak [J.L. Luxon, et al., Plasma Phys. and Control. Fusion Research, Vol. 1 (International Atomic Energy Agency, Vienna, 1987) p. 159] in the presence of periodic q = 1 sawtooth instabilities, a reactor relevant regime. Radially localized off-axis electron cyclotron current drive (ECCD) must be precisely located on the island. In DIII-D the plasma control system is put into a "search and suppress" mode to make either small rigid radial position shifts of the entire plasma (and thus the island) or small changes in toroidal field (and thus, ECCD location) to find and lock onto the optimum position for complete island suppression by ECCD. This is based on real-time measurements of an m n = 3 2 mode amplitude dB dt θ . The experiment represents the first use of active feedback control to provide continuous, precise positioning. An alternative to ECCD makes use of the six toroidal section "C-Coil" on DIII-D to provide a large non-resonant static m = 1, n = 3 helical field to interfere with the fundamental harmonic of an m n = 3 2 NTM. While experiments show success in inhibiting the NTM if a large enough n = 3 field is applied before the island onset, there is a considerable plasma rotation decrease due to n = 3 "ripple".

The first suppression of the important and deleterious m=2/n= 1 neoclassical tearing mode (NTM) is reported using electron cyclotron current drive (ECCD) to replace the "missing" bootstrap current in the island 0-point. Experiments on the DIII-D tokamak verify that maximum shrinkage of the m=2/n=l island occurs when the ECCD location coincides with the q=2 surface. The DIII-D plasma control system is put into "search and suppress" mode to make small changes in the toroidal field to find and lock onto the optimum position, based on real time measurements of dBddt, for complete m=2/n=l NTM suppression by ECCD. The requirements on the ECCD for complete island suppression are well modeled by the modified Rutherford equation for the DIII-D plasma conditions. ...Recently several tokamaks have demonstrated the suppression of the m = 3/n = 2 NTM using unmodulated ECCD positioned at the island location. On ASDEX Upgrade, co-ECCD was verified to be more effective at NTM stabilization than counter-ECCD or pure heating alone [7-91. These experiments used a programmed sweep of the toroidal magnetic field to ensure that the current drive layer matched the mode location at some time during the ECCD pulse. On JT-GOU, suppression of the m = 3/n = 2 mode was also achieved for 1.5 s in steady conditions using ECCD

A potential new standard in stationary tokamak performance is emerging from experiments on DIII-D. These experiments have demonstrated the ability to operate near the free boundary, n = 1 stability limit with good confinement quality under stationary conditions. The normalized fusion performance is at or above that projected for Qfus = 10 operation in the International Thermonuclear Experimental Reactor (ITER) design over a wide operating range in both edge safety factor (2.8–4.7) and plasma density (35–70% of the Greenwald density). Projections to ITER based on this data are uniformly positive and indicate that a wide range of operating options may be available on ITER, including the possibility of sustained ignition. Recent experiments have demonstrated the importance of a small m = 3, n = 2 neoclassical tearing mode in avoiding sawteeth and the effect of edge localized modes on tearing mode stability at an edge safety factor near 3. Transport studies using the GLF23 turbulence transport code suggest that E × B shear stabilization is important in reproducing the measured profiles in the simulation. Yet, even in cases in which the toroidal rotation is moderate, confinement quality is robustly better than the standard H-mode confinement scalings.

Abstract. The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, and current, is a critical element for achieving mission performance. Analysis of PF capabilities has focused on the 15 MA Q = 10 scenario with a 300-500 s flattop burn phase. The operating space available for the 15 MA ELMy H-mode plasma discharges in ITER and upgrades to the PF coils or associated systems to establish confidence that ITER mission objectives can be reached have been identified. Time dependent self-consistent free-boundary calculations were performed to examine the impact of plasma variability, discharge programming, and plasma disturbances. Based on these calculations a new reference scenario was developed based upon a large bore initial plasma, early divertor transition, low level heating in L-mode, and a late H-mode onset. Equilibrium analyses for this scenario indicate that the original PF coil limitations do not allow low l i (<0.8) operation or lower flux states, and the flattop burn durations were predicted to be less than the desired 400 s. This finding motivates the expansion of the operating space, considering several upgrade options to the PF coils. Analysis was also carried out to examine the feedback current reserve required in the CS and PF coils during a series of disturbances and a feasibility assessment of the 17 MA scenario was undertaken. Results of the studies show that the new scenario and modified PF system will allow a wide range of 15 MA 300-500 s operation and more limited but finite 17 MA operation.