ADVANCED TOKAMAK PROFILE EVOLUTION IN DIII-D. PDF Download

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Languages : en
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Using off-axis electron cyclotron current drive (ECCD), self-consistent integrated advanced tokamak operation has been demonstrated on DIII-D, combining high[beta] (>3%) at high q(q[sub min]> 2.0) with good energy confinement (H[sub 89][approx] 2.5) and high noninductive current fraction (f[sub BS][approx] 55%, f[sub NI][approx] 90%). Modification of the current profile by ECCD led to internal transport barrier formation even in the presence of type I edge localized modes. Improvements were observed in all transport channels, and increased peaking of profiles led to higher bootstrap current in the core. Separate experiments have shown the ability to maintain a nearly steady-state current profile for up to 1 s with q[sub min]> 1.5. Modeling indicates that this favorable current profile can be maintained indefinitely at a higher[beta][sub N] using tools available to the near-term DIII-D program. Modeling and simulation have become essential tools for the experimental program in interpreting the data and developing detail plans for new experiments.

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Languages : en
Pages : 5

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In DIII-D, as in a number of tokamaks, high performance is obtained with various optimized magnetic shear configurations that exhibit internal transport barriers. Negative central shear (NCS) discharges are created transiently during the current ramp-up by auxiliary heating and current drive from neutral beam injection. Both q{sub min} and the radius at which it occurs, [rho]{sub qmin}, decrease with time as the Ohmic current diffuses inward. The q-profiles calculated using EFIT with external magnetic and Motional Stark Effect (MSE) measurements as constraints are comparable to those calculated with the Corsica code, a time-dependent, 2D equilibrium and 1D transport modeling code. Corsica is used to predict the temporal evolution of the current density from a combination of measured profiles, transport models and neoclassical resistivity. Using these predictive capabilities, the authors are exploring methods for increasing the duration and [rho]{sub qmin} of the NCS configuration by local control of the current density profile with simulations of the possible control available from the electron cyclotron heating and current drive system currently being upgraded on DIII-D. Their intention is not to do a detailed investigation of transport models but rather to provide a reasonable model of heat conductivity to be able to simulate effects of electron cyclotron heating (ECH) and current drive (ECCD) on confinement in NCS configurations. The authors adjust free parameters (c, c1 and c2) in the model to obtain a reasonable representation of the temporal evolution of electron and ion temperature profiles consistent with those measured in selected DIII-D shots. In all cases, they use the measured density profiles rather than self-consistently solve for particle sources and particle transport at this time.

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Languages : en
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Key elements of a sustained advanced tokamak discharge in DIII-D are a large fraction of the total current from bootstrap current (f[sub BS]) and parameters that optimize the capability to use electron cyclotron current drive (ECCD) at[rho][approx] 0.5 to maintain the desired current profile[1-4]. Increased f[sub BS] results from increasing both the normalized beta ([beta][sub N]) and the minimum value of the safety factor (q[sub min]). Off-axis ECCD is, for the available gyrotron power, optimized at high[beta][sub N], high electron temperature (T[sub e]) and low electron density (n[sub e]). As previously reported[2-4], these required elements have been separately demonstrated: density control at high[beta][sub N] with n[sub e][le] 5 x 10[sup 19] m[sup -3] using divertor-region pumping, stability at high[beta], and off-axis ECCD at the theoretically predicted efficiency. This report summarizes recent work on optimizing and integrating these results through evaluation of the dependence of the beta limit on q[sub min] and q[sub 95], exploration of discharges with relatively high q[sub min], testing of feedback control of T[sub e] for control of the q profile evolution, and modification of the current profile time evolution when ECCD is applied.

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Languages : en
Pages : 4

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The effect of co and counter fast wave current drive (FWCD) on the plasma current profile has been measured for neutral beam heated plasmas with reversed magnetic shear on the DIII-D tokamak. Although the response of the loop voltage profile was consistent with the application of co and counter FWCD, little difference was observed between the current profiles for the opposite directions of FWCD. The evolution of the current profile was successfully modeled using the ONETWO transport code. The simulation showed that the small difference between the current profiles for co and counter FWCD was mainly due to an offsetting change in the o@c current proffie. In addition, the time scale for the loop voltage to reach equilibrium (i.e., flatten) was found to be much longer than the FWCD pulse, which limited the ability of the current profile to fully respond to co or counter FWCD.

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Languages : en
Pages : 10

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One of the main goals for the DIII-D research program is to establish an advanced tokamak plasma with high bootstrap current fraction that can be sustained in-principle steady-state. Substantial progress has been made in several areas during the last year. The resistive wall mode stabilization has been done with spinning plasmas in which the plasma pressure has been extended well above the no-wall beta limit. The 3/2 neoclassical tearing mode has been stabilized by the injection of ECH into the magnetic islands, which drives current to substitute the missing bootstrap current. In these experiments either the plasma was moved or the toroidal field was changed to overlap the ECCD resonance with the location of the NTMs. Effective disruption mitigation has been obtained by massive noble gas injection into shots where disruptions were deliberately triggered. The massive gas puff causes a fast and clean current quench with essentially all the plasma energy radiated fairly uniformly to the vessel walls. The run-away electrons that are normally seen accompanying disruptions are suppressed by the large density of electrons still bound on the impurity nuclei. Major elements required to establish integrated, long-pulse, advanced tokamak operations have been achieved in DIII-D: [beta]{sub T} = 4.2%, [beta]{sub p} = 2, f{sub BS} = 65%, and [beta]{sub N}H9 = 10 for 600 ms (H"4[tau]{sub E}). The next challenge is to integrate the different elements, which will be the goal for the next five years when additional control will be available. Twelve resistive wall mode coils are scheduled to be installed in DIII-D during the summer of 2003. The future plans include upgrading the tokamak pulse length capability and increasing the ECH power, to control the current profile evolution.

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Languages : en
Pages : 19

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Languages : en
Pages : 7

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Recent DIII-D advanced tokamak experiments with negative central shear (NCS) have resulted in operation at high normalized [beta], [beta]{sub N}=[beta]/(I/aB), to 4.2, confinement enhancement factors to H=4 (H=[tau]{sub E}/[tau]ITER-89P), and record neutron rates for DIII-D to 2.4X1016 neutrons/sec. These data were obtained during high triangularity, single and double null diverted operation with peaked (L-mode) and broad (H-mode) pressure profiles. We are modeling the spatial and temporal current profile evolution for these discharges using Corsica, a predictive 1-1/2 D equilibrium and transport code. Current profile evolution is self-consistently determined by including current diffusion resulting from current drive due to early neutral beam injection during the ohmic current ramp-up phase of the discharge and the bootstrap current drive associated with pressure profile evolution.

Author: L. Pearlstein
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Languages : en
Pages : 6

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An intermediate regime for tokamak operation has been obtained in DIII-D and in other tokamaks in which the inductive flux consumption is reduced and a broad current profile with the safety factor just above or near the sawtoothing limit is obtained and maintained. The DIII-D tokamak was operated in this regime near the no-wall b limit. High stability and good confinement was achieved at a desired level of q{sub 95} {approx} 3 to 4 for durations as long as 35{tau}{sub E}, three times the current-diffusion time. This regime offers the promise of achieving higher fusion gain and yield and/or longer burn duration for ITER.

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Languages : en
Pages : 12

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A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response.

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Languages : en
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Recent experiments on DIII-D have demonstrated the ability to sustain plasma conditions that integrate and sustain the key ingredients of Advanced Tokamak (AT) operation: high[beta] with q[sub min]” 1, good energy confinement, and high current drive efficiency. Utilizing off-axis ([rho]= 0.4) electron cyclotron current drive (ECCD) to modify the current density profile in a plasma operating near the no-wall ideal stability limit with q[sub min]> 2.0, plasmas with[beta]= 2.9% and 90% of the plasma current driven non-inductively have been sustained for nearly 2 s (limited only by the duration of the ECCD pulse). Separate experiments have demonstrated the ability to sustain a steady current density profile using ECCD for periods as long as 1 s with[beta]= 3.3% and> 90% of the current driven non-inductively.