Session BI1: Current Drive, Energetic Particles, and Steady State

Lower Hybrid Current Drive Experiments in Alcator C-Mod

Lower Hybrid Current Drive (LHCD) experiments have been implemented in Alcator C-Mod. The long-term objective is to use LHCD to control j(r) and to supplement the bootstrap current in high beta-poloidal and enhanced confinement regimes. Thus far up to 800 kW of RF power at 4.6 GHz have been coupled to C-Mod plasmas through a waveguide grill arranged in 4 rows, each with 24 waveguides. Electronic control of the amplitude and phase in each waveguide allows dynamic variation of the n-parallel spectrum with 1 ms response time over the range 1.5 $<$4. This feature enables measurement of the reflection coefficient as a function of n-par in one discharge. By varying the gap between the separatrix and the grill, its dependence on density at the grill is also obtained. Good agreement between the measurements and modeling based on the Brambilla coupling code (M. Brambilla 1976 Nuclear Fusion \textbf{16, }47-54) is obtained by assuming a small ($\sim $1 mm) vacuum gap between the grill and the plasma in the code. Hard X-Ray imaging of the accelerated fast-electron bremsstrahlung reveals a relatively broad emission profile with energies in excess of 100 keV. Non-thermal synchrotron emission is also observed. Modeling of the hard X-ray and ECE emission using both the CQL3D and DKE codes in synthetic diagnostic mode is in progress and comparison with the experimental spectra and spatial profiles will be presented. LHCD experiments have thus far operated in the ITER relevant line-average density range from 5e19m-3 to 8e19 m-3 and toroidal field 5.4 $<$ B $<$ 6.5 T. At the 800 kW power level the maximum decrease in loop voltage due to LHCD is about 75{\%} of the ohmic value at a plasma current of 700 kA. Analysis is underway to quantify the LH driven current fraction. A decrease in internal inductance of about 0.1 has been measured, suggesting a broadening of the current profile. Sawtooth stabilization and other core MHD activity have also been observed as well as substantial electron heating.   

Solenoid-free Plasma Start-up in NSTX using Transient CHI

For the first time, 160 kA of closed flux current is produced in the National Spherical Torus Experiment (NSTX), without using a solenoid. Transient Coaxial Helicity Injection (CHI) is used to generate the self-contained equilibrium; an important step in the production of a starting equilibrium for solenoid free operation. Until now, almost all tokamaks and spherical torus experiments have relied on a central solenoid to produce the plasma current needed to confine the plasma. An alternate method for plasma startup is essential for developing a fusion reactor based on the spherical torus concept and could also reduce the cost of a tokamak reactor. In this method, an external power supply rapidly drives current between coaxial electrodes, in the presence of toroidal and poloidal magnetic fields, which causes the poloidal flux connecting the lower divertor plates to rapidly expand into the chamber. When the injected current is rapidly decreased, magnetic reconnection occurs near the electrodes, with the toroidal plasma current forming closed flux surfaces, which is assessed using Thomson scattering electron temperature measurements and by equilibrium reconstructions. The method has previously been demonstrated in the HIT-II device [R. Raman \textit{et al.}, Phys Rev. Lett. 90 075005 (2003)]. The significance of these new NSTX results, which represent a proof-of-principle demonstration of the concept, are (a) demonstration of the process in a vessel volume thirty times larger than HIT-II on a size scale closer to a reactor, (b) a remarkable multiplication factor of 60 between the injected current and the achieved toroidal current, compared to six in previous experiments, and (c) for the first time, fast time-scale visible imaging of the entire process that shows discharge formation, disconnection from the injector and the reconnection of magnetic field lines to form closed flux.   

Improved confinement MST RFP plasmas with hot ions and high density

In MST plasmas with improved confinement, we have achieved for the first time a large ion temperature, reaching at least 1 keV. This is achieved with the capture of ion heat generated during magnetic reconnection. In separate discharges, we have quadrupled the plasma density by injecting deuterium pellets. This has also resulted in an increased ion temperature and in the largest beta yet observed in the RFP during improved confinement. Fluctuation reduction via auxiliary current drive in MST has previously resulted in a ten-fold improvement in the total energy confinement time and a doubling of beta. However, this improvement was due solely to a large increase in the electron temperature, T$_{e}$, exceeding 1 keV. Essentially no change in the ion temperature, T$_{i}$, was observed. We can now reliably trigger magnetic reconnection (large fluctuation) events that generate many MW of global ion heating power, driving T$_{i}$ to well over 1 keV and, according to the neutron flux, creating a population of fast ions. Reducing fluctuations soon after such events allows sustainment of the large T$_{i}$, with a several-fold reduction in thermal ion energy transport. Improved confinement plasmas were previously limited to relatively low density in order to prevent the destabilization of edge-resonant tearing modes. The four-fold density increase (up to 4 x 10$^{19}$ m$^{-3})$ with pellet injection occurs without triggering these modes, due, we believe, to the central deposition of particles without the edge cooling that accompanies gas puffing. With pellet injection, there is still a substantial increase in T$_{e}$, and this is largely matched by an increase in T$_{i}$, yielding a record total beta of 26{\%}. Work supported by USDOE and NSF.   

Radial Redistribution of Neutral Beam Ions Induced by Multiple Alfv\'en Eigenmodes in DIII-D

For the first time the detailed radial structure of multiple Alfv\'en eigenmodes and the associated transport of 80~keV neutral beam ions in the DIII-D tokamak have been measured using an upgraded array of internal fluctuation diagnostics in combination with a new spectroscopic diagnostic for the confined fast ion population. Knowledge of the fast ion redistribution and the internal structure of Alfv\'en waves affecting fast ion transport are critical for understanding equilibrium, stability and transport physics in advanced confinement regimes in ITER. The observed mode activity in DIII-D includes toroidal Alfv\'en eigenmodes (TAEs), reversed-shear Alfv\'en eigenmodes (RSAEs) and their spatial coupling. The measured radial mode structures are in close agreement with calculations of the modes and their coupling using the NOVA code. The ideal MHD calculations take into account the effects of adiabatic compression in order to predict the ratio of temperature to density fluctuations and the results are also found to agree closely with internal measurements. Fast ion measurements indicate a significant depletion of the central beam ion population during the peak of the observed mode activity. The detailed identification of these modes allows for the accurate calculations of fast ion transport using the particle following ORBIT code. The ORBIT code analysis, using real plasma geometry and the measured structure of Alfv\'en eigenmodes, will be compared to the measured radial redistribution of neutral beam ions. These results are directly applicable to future burning plasma experiments like ITER where alpha particle driven Alfv\'en eigenmodes are expected and a predictive model of the fast ion loss and redistribution is required.   

Energetic ion-driven instabilities on JET and on MAST.

Measurements and understanding of energetic ion driven instabilities such as TAE, Alfven cascades (ACs), and chirping modes are crucial for assessing all aspects of such instabilities in burning plasmas. Ion loss and Alfven spectroscopy associated with energetic ion-driven modes were investigated in experiments on JET and on MAST with a wide range of new diagnostics. In JET with monster sawteeth, `tornado' modes (TAE inside the q=1 radius) are observed with mode numbers decreasing one-by-one before sawtooth crash. The effect of the tornado modes on fast ions is investigated with improved X-mode reflectometry, new lost ion scintillator and Faraday cups, and gamma-ray and NPA diagnostics. Modelling with the HAGIS code shows the importance of the frequency sweep of tornado modes for depleting the fast ion density inside the q=1 radius. JET reversed-shear plasmas reveal ACs, which provide information on the time evolution of the minimum safety factor, q(min). X-mode reflectometry detecting ACs provides now information on the mode localisation region, allowing determination of the q(min) evolution in space and time and facilitating the development of ITB scenarios. Alfven spectroscopy aiming at kinetic information on electron and ion pressures, as well as the use of sub-Alfvenic NBI for exciting ACs is reviewed. TAEs and chirping modes are excited by super-Alfvenic NBI in MAST discharges. In addition, there are indications of ACs in MAST discharges targeted at ITB formation. The super-Alfvenic NBI makes MAST a good test bed for testing diagnostics for Alfven instabilities. MAST is now equipped with magnetic coils with frequency range up to 5 MHz, soft X-ray camera with frequency range up to 250 kHz, and multi-channel (in energy and in lines-of-sight) NPA.   

High Beta Steady State Research with Integrated Modeling in JT-60U

Improved performance and its long sustainment were obtained by a magnetic material of ferritic steel tiles in JT-60U, which are recently installed inside the vacuum vessel to reduce the fast ion loss by the decrease of the toroidal field ripple. Temperature of the pedestal region explicitly increases by $\sim $20{\%} and the H-mode confinement clearly improved. The improvement is implied to be caused by the change of the radial electric field and the co- rotation near the plasma edge attributed to decrease the fast ion loss. Higher beta ($\beta _{N}>$3) than the no-wall beta limit was obtained by the increase of the net heating power due to the reduction of the fast ion loss and by the close conducting wall. The growth time of the RWM becomes shorter and the achieved beta becomes lower by the decrease of the co- or counter rotation. Performance of long-pulse ELMy H-mode plasma was extended by the integration of the improvement of the confinement and the stability in addition to the suppression of NTM and a real time profile control. High $\beta _{N} \quad >$2.3 simultaneously with H$_{98}\sim $1 was sustained for 23.1s ($\sim $12 times of the current diffusion time) at q$_{95}\sim $3.2, which exceed the ITER reference scenario of $\beta _{N}$ H$_{98}\sim $1.8. The electron density has been successfully controlled by the active divertor pumping in long-pulse high-density ELMy H-mode plasmas where the wall pumping is not effective and even outgas from the divertor tiles appears. In order to realize a steady-state tokamak fusion reactor, an integrated control of high performance is a critical issue. Based on the analysis of experiments of JT-60U, an integrated model of pedestal/ELM/SOL of the transport code TOPICS and the stability code MARG2D and an integrated model of divertor/neutral/impurity of the edge transport code SONIC are produced as an effective means for the evaluation and the simulation of complex burning plasmas.