Divertor characterization experiments

Recent DIII-D experiments with enhanced Scrape-off Layer (SOL) diagnostics permit detailed characterization of the SOL and divertor plasma under various operating conditions. We observe two distinct plasma modes: attached and detached divertor plasmas. Detached plasmas are characterized by plate temperatures of only 1 to 2 eV. Simulation of detached plasmas using the UEDGE code indicate that volume recombination and charge exchange play an important role in achieving detachment. When the power delivered to the plate is reduced by enhanced radiation to the point that recycled neutrals can no longer be efficiently ionized, the plate temperature drops from around 10 eV to 1-2 eV. The low temperature region extends further off the plate as the power continues to be reduced, and charge exchange processes remove momentum, reducing the plasma flow. Volume recombination becomes important when the plasma flow is reduced sufficiently to permit recombination to compete with flow to the plate

Gyrokinetic analysis and simulation of pedestals to identify the culprits for energy losses using 'fingerprints'

Fusion performance in tokamaks hinges critically on the efficacy of the edge transport barrier (ETB) in suppressing energy losses. The new concept of 'fingerprints' is introduced to identify the instabilities that cause transport losses in the ETBs of many of today's experiments, from among widely posited candidates. Analysis of the gyrokinetic-Maxwell equations and gyrokinetic simulations of experiments reveals that each mode type produces characteristic ratios of transport in the various channels: density, heat, and impurities. This, together with experimental observations of transport in some channel or of the relative size of the driving sources of channels, can identify or determine the dominant modes causing energy transport. In multiple H-mode cases with edge-localized modes that are examined, these fingerprints indicate that magnetohydrodynamic (MHD)-like modes are apparently not the dominant agent of energy transport; rather, this role is played by micro-tearing modes (MTMs) and electron temperature gradient (ETG) modes, and in addition, possibly by ion temperature gradient/ trapped electron modes (ITG/TEM) on JET (Joint European 'Torus). MHD-like modes may dominate the electron particle losses. Fluctuation frequency can also be an important means of identification, and is often closely related to the transport fingerprint. The analytical arguments unify and explain previously disparate experimental observations on multiple devices, including DIII-D, JET, and ASDEX-U. Detailed simulations of two DIII-D ETBs also demonstrate and corroborate this.Peer reviewe

SUSTAINED STABILIZATION OF THE RESISTIVE WALL MODE BY PLASMA ROTATION IN THE DIII-D TOKAMAK

OAK-B135 A path to sustained stable operation, at plasma pressure up to twice the ideal magnetohydrodynamic (MHD) n = 1 free-boundary pressure limit, has been discovered in the DIII-D tokamak. Tuning the correction of the intrinsic magnetic field asymmetries so as to minimize plasma rotation decay during the high beta phase and increasing the angular momentum injection, have allowed maintaining the plasma rotation above that needed for stabilization of the resistive wall mode (RWM). A new method to determine the improved magnetic field correction uses feedback to sense and minimize the resonant plasma response to the non-axisymmetric field. At twice the free-boundary pressure limit, a disruption precursor is observed, which is consistent with having reached the ''ideal wall'' pressure limit predicted by stability calculations

CONTROL OF MHD STABILITY IN DIII-D ADVANCED TOKAMAK DISCHARGES

OAK-B135 Advanced tokamak research in DIII-D seeks to optimize the tokamak approach for fusion energy production, leading to a compact, steady state power source. High power density implies operation at high toroidal beta, {beta}{sub T}=<p>2{micro}{sub 0}/B{sub T}{sup 2}, since fusion power density increases roughly as the square of the plasma pressure. Steady-state operation with low recirculating power for current drive implies operation at high poloidal beta, {beta}{sub P} = <p>2{micro}{sub 0}/<B{sub P}>{sup 2}, in order to maximize the fraction of self-generated bootstrap current. Together, these lead to a requirement of operation at high normalized beta, {beta}{sub N} = {beta}{sub T}(aB/I), since {beta}{sub P}{beta}{sub T} {approx} 25[(1+{kappa}{sup 2})/2] ({beta}{sub N}/100){sup 2}. Plasmas with high normalized beta are likely to operate near one or more stability limits, so control of MHD stability in such plasmas is crucial

Progress Toward Fully Noninductive, High Beta Discharges in Diii-D

OAK-B135 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 {beta} above the n = 1 no-wall limit, enabled by active feed-back control. The ideal wall {beta} limit is optimized by modifying the plasma shape, current and pressure profile. Present DIII-D AT experiments operate with f{sub BS} {approx} 50%-60%, with a long-term goal of {approx} 90%. Additional current is provided by neutral beam and electron cyclotron current drive, the latter being localized well away from the magnetic axis ({rho} {approx} 0.4-0.5). Guided by integrated modeling, recent experiments have produced discharges with {beta} {approx} 3%, {beta}{sub N} {approx} 3, f{sub BS} {approx} 55% and noninductive fraction f{sub NI} {approx} 90%. Additional control is anticipated using fast wave current drive to control the central current density

Current ramps in tokamaks: from present experiments to ITER scenarios

In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for L-mode (namely, the Bohm-gyroBohm model and scaling based models with a specific fixed radial shape and energy confinement time factors of H(96-L) = 0.6 or H(IPB98) = 0.4) has been validated on a multi-machine experimental dataset for predicting the l(i) dynamics within +/- 0.15 accuracy during current ramp-up and ramp-down phases. Simulations using the Coppi-Tang or GLF23 models (applied up to the LCFS) overestimate or underestimate the internal inductance beyond this accuracy (more than +/- 0.2 discrepancy in some cases). The most accurate heat transport models are then applied to projections to ITER current ramp-up, focusing on the baseline inductive scenario (main heating plateau current of I(p) = 15 MA). These projections include a sensitivity study to various assumptions of the simulation. While the heat transport model is at the heart of such simulations (because of the intrinsic dependence of the plasma resistivity on electron temperature, among other parameters), more comprehensive simulations are required to test all operational aspects of the current ramp-up and ramp-down phases of ITER scenarios. Recent examples of such simulations, involving coupled core transport codes, free-boundary equilibrium solvers and a poloidal field (PF) systems controller are also described, focusing on ITER current ramp-down.</p