DEVELOPMENT IN THE DIII-D TOKAMAK OF HYBRID OPERATION SCENARIOS FOR BURNING PLASMA EXPERIMENTS

OAK-B135 The basic parameters of proposed burning plasma experiments such as ITER and FIRE have been chosen based on analysis of multi-machine databases of confinement, stability, and divertor operation. given these specifications, it is of interest to run discharges in present-day machines such as DIII-D to verify the design basis and evaluate the margin available to achieve the mission goals. it is especially important to operate discharges which are stationary with respect to the current relaxation time scale ({tau}{sub R}) since it is well-known that higher performance can be achieved transiently. Attention has been focused on validating the baseline scenario for diverted machines--ELMing H-mode discharges with q{sub 95} = 3 with sawteeth. However, there is also interest in the ITER program to assess the feasibility of operating the tokamak in a mode to maximize the neutron fluence for the purpose of testing the design of various components critical to the nuclear fuel cycle and energy conversion systems in a fusion power plant. It was originally envisioned that these discharges would be intermediate between an inductive burn (baseline) scenario and a fully noninductive (steady state) scenario; therefore, this type of discharge has become known as a hybrid scenario. In the course of investigating these hybrid scenarios in DIII-D, two key results have been obtained. First, stationary discharges with q{sub 95} > 4 have been obtained which project to Q{sub fus} {approx} 10 in ITER. The projected duration of these discharges in ITER when using the full inductive flux capability is > 4000 s. (The significant engineering issues of site heat capacity, activation, and tritium consumption are beyond the scope of this work). Second, utilizing the same plasma initiation techniques as developed for the hybrid scenario, discharges at q{sub 95} = 3.2 project to near ignition in ITER, even with reduced parameters. This indicates the ITER design has significant performance margin and possesses the physics capability to carry out an extensive nuclear testing program. These same q{sub 95} = 3.2 discharges project to Q{sub fus} > 5 in FIRE, even with pessimistic confinement scalings

ECRH-assisted plasma start-up with toroidally inclined launch: multi-machine comparison and perspectives for ITER

Electron cyclotron resonance heating (ECRH)-assisted plasma breakdown is foreseen with full and half magnetic field in ITER. As reported earlier, the corresponding O1- and X2-schemes have been successfully used to assist pre-ionization and breakdown in present-day devices. This contribution reports on common experiments studying the effect of toroidal inclination of the ECR beam, which is >= 20 degrees in ITER. All devices could demonstrate successful breakdown assistance for this case also, although in some experiments the necessary power was almost a factor of 2 higher compared with perpendicular launch. Differences between the devices with regard to the required power and vertical field are discussed and analysed. In contrast to most of these experiments, ITER will build up loop voltage prior to the formation of the field null due to the strong shielding by the vessel. Possible consequences of this difference are discussed.X112423sciescopu

Operation at high performance in optimized shear plasmas in JET

Heating during the early part of the current rise phase gives a low or negative magnetic shear (= 0741-3335/40/6/020/img27(dq/dr)) in the centre of JET plasmas. Under these conditions the confinement improves with high additional heating power heating during the current ramp-up phase of the discharge. The reduction in the transport manifests itself as a peaking of the profiles with a large gradient region near 0741-3335/40/6/020/img28 = 0.55. The best discharges have no transport barrier at the edge of the plasma (L-mode). This allows central power deposition by the neutral beams in JET. A control of the plasma pressure, using feedback of the additional heating power in real-time, minimizes the impact of magnetohydrodynamic instabilities. As a result, these discharges achieve the highest D-D neutron rates in JET; 0741-3335/40/6/020/img29, with 0741-3335/40/6/020/img30, 0741-3335/40/6/020/img31 and 0741-3335/40/6/020/img32

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 diagnostic measurements. The duration of this state is limited by pressure profile evolution, leading to magnetohydrodynamic (MHD) instabilities after about 1 s or half of a current relaxation time (τCR). Stationary conditions are maintained in similar discharges (∼90% noninductive), limited only by the 2 s duration (1 τCR) of the present ECCD systems. By discussing parametric scans in a global parameter and profile databases, the need for low density and high beta are identified to achieve full noninductive operation and good current drive alignment. These experiments achieve the necessary fusion performance and bootstrap fraction to extrapolate to the fusion gain, Q=5 steady-state scenario in the International Thermonuclear Experimental Reactor (ITER) [R. Aymar, Fusion Energy Conference on Controlled Fusion and Plasma Physics, Sorrento, Italy (International Atomic Energy Agency, Vienna, 1987), paper IAEA-CN-77/OV-1]. The modeling tools that have been successfully employed to both plan and interpret the experiment are used to plan future DIII-D experiments with higher power and longer pulse ECCD and fast wave and co- and counterneutral beam injection in a pumped double-null configuration. The models predict our ability to control the current and pressure profiles to reach full noninductivity with increased beta, bootstrap fraction, and duration. The same modeling tools are applied to ITER, predicting favorable prospects for the success of the ITER steady-state scenario. © 2006 American Institute of Physics

Progress towards high-performance steady-state operation on DIII-D

Advanced Tokamak research in DIII-D seeks to develop a scientific basis for steady-state high-performance tokamak operation. Fully noninductive (fNI ≈ 100%) in-principle steady-state discharges have been maintained for several confinement times. These plasmas have weak negative central shear with qmin ≈ 1.5-2, βN ≈ 3.5, and large, well-aligned bootstrap current. The loop voltage is near zero across the entire profile. The remaining current is provided by neutral beam current drive (NBCD) and electron cyclotron current drive (ECCD). Similar plasmas are stationary with fNI ≈ 90-95% and duration up to 2 s, limited only by hardware. In other experiments, βN ≈ 4 is maintained for 2 s with internal transport barriers, exceeding previously achieved performance under similar conditions. This is allowed by broadened profiles and active magnetohydrodynamic instability control. Modifications now underway on DIII-D are expected to allow extension of these results to higher performance and longer duration. A new pumped divertor will allow density control in high triangularity double-null divertor configurations, facilitating access to similar in-principle steady-state regimes with βN > 4. Additional current drive capabilities, both off-axis ECCD and on-axis fast wave current drive (FWCD), will increase the magnitude, duration, and flexibility of externally driven current. © 2006 Elsevier B.V. All rights reserved