Bulletin of the American Physical Society
2005 47th Annual Meeting of the Division of Plasma Physics
Monday–Friday, October 24–28, 2005; Denver, Colorado
Session LI1b: Magnetic Plasma Technology |
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Chair: J. Reece Roth, University of Tennessee Room: Adam's Mark Hotel Plaza Ballroom ABC |
Wednesday, October 26, 2005 3:00PM - 3:30PM |
LI1b.00001: Gas Jet Disruption Mitigation Studies on Alcator C-Mod Invited Speaker: Disruptions are a major concern for C-Mod, ITER, and future tokamak reactors. Damage can come from several different effects, including high instantaneous thermal loads on material surfaces, electromagnetic loads on conducting structures due to halo and toroidal eddy currents, and highly localized damage from relativistic runaway electron impact. Reliable mitigation of these problems using techniques benign to tokamak operation are key to meeting the scientific and technological goals of these experiments. High pressure noble gas jet injection is a mitigation technique which potentially satisfies the operational requirements of fast response time and reliability, while still being benign to subsequent discharges. Previous gas jet injection experiments on DIII-D have shown good success at reducing the deleterious effects of disruptions. However, many questions remain about the effectiveness of this approach on high energy density, high pressure plasmas in high field devices such as C-Mod and ITER. Disruption mitigation experiments using an optimized gas jet injection system are being carried out on C-Mod to study the physics of gas jet penetration into high pressure plasmas, the ability of the gas jet to convert plasma energy into radiation on timescales consistent with C-Mod's fast quench times, and the reduction of halo currents with C-Mod's high current density. The dependence of impurity penetration and effectiveness on noble gas species (He, Ne, Ar), gas pressure, and plasma pressure/energy density will also be discussed. 3-D MHD modeling of the disruption physics with NIMROD, incorporating data from temperature profiles taken during the disruption quench, as well as high-speed images of the plasma cross-section in the gas nozzle region, show that edge cooling of the plasma triggers the growth of tearing modes, producing a stochastic region in the core of the plasma and rapid loss of core thermal energy. This may explain the apparent effectiveness of the gas jet despite its limited penetration. [Preview Abstract] |
Wednesday, October 26, 2005 3:30PM - 4:00PM |
LI1b.00002: Active Control for Stabilization of Neoclassical Tearing Modes Invited Speaker: We describe active control algorithms used by \hbox{DIII-D} to stabilize and maintain suppression of 3/2 or 2/1 neoclassical tearing modes (NTMs) using electron cyclotron current drive (ECCD) at the rational q-surface. The \hbox{DIII-D} NTM control system can determine the correct q-surface/ECCD alignment and stabilize existing modes within 100-200~ms of activation, or prevent mode growth with preemptive application of ECCD, in both cases enabling stable operation at normalized beta values above 3.5. Because NTMs can limit performance or cause plasma-terminating disruptions in tokamaks, their stabilization is essential to the high performance operation of ITER. The \hbox{DIII-D} NTM control system has demonstrated many elements of an eventual ITER solution, including general algorithms for robust detection of q-surface/ECCD alignment and for realtime maintenance of alignment following disappearance of the mode. This latter capability, unique to \hbox{DIII-D}, is based on realtime reconstruction of q-surface geometry by a Grad-Shafranov solver using external magnetics and internal motional Stark effect measurements. Alignment is achieved by varying either the plasma major radius (and the rational q-surface) or the toroidal field (and the deposition location). The requirement to achieve and maintain q-surface/ECCD alignment with an accuracy on the order of 1 cm is routinely met by the \hbox{DIII-D} Plasma Control System and these algorithms. We discuss the integrated plasma control design process used for developing these and other general control algorithms, which includes physics-based modeling and testing of the algorithm implementation against simulations of actuator and plasma responses. This systematic design/test method and modeling environment enabled successful mode suppression by the NTM control system upon first-time use in an experimental discharge. [Preview Abstract] |
Wednesday, October 26, 2005 4:00PM - 4:30PM |
LI1b.00003: Recent Progress on ECH Technology for ITER Invited Speaker: The Electron Cyclotron Heating and Current Drive (ECH{\&}CD) system for ITER is a critical ITER system that must be available for use on Day 1 of the ITER experimental program. The applications of the system include plasma start-up, plasma heating and suppression of Neoclassical Tearing Modes (NTMs). These applications are accomplished using 27 one megawatt continuous wave gyrotrons: 24 at a frequency of 170 GHz and 3 at a frequency of 120 GHz. There are DC power supplies for the gyrotrons, a transmission line system, one launcher at the equatorial plane and three upper port launchers. The US will play a major role in delivering parts of the ECH{\&}CD system to ITER. The present state-of-the-art includes major advances in all areas of ECH technology. In the US, a major effort is underway to supply gyrotrons of up to 1.5 MW power level at 110 GHz to General Atomics for use in heating the DIII-D tokamak. This presentation will include a brief review of the state-of-the-art, worldwide, in ECH technology. The requirements for the ITER ECH{\&}CD system will then be reviewed. ITER calls for gyrotrons capable of operating from a 50 kV power supply, after potential depression, with a minimum of 50{\%} overall efficiency. This is a very significant challenge and some approaches to meeting this goal will be presented. Recent experimental results at MIT showing improved efficiency of high frequency, 1.5 MW gyrotrons will be described. These results will be incorporated into the planned development of gyrotrons for ITER. The ITER ECH{\&}CD system will also be a challenge to the transmission lines, which must operate at high average power at up to 1000 seconds and with high efficiency. The technology challenges and efforts in the US and other ITER parties to solve these problems will be reviewed. \newline \newline *In collaboration with E. Choi, C. Marchewka, I. Mastovosky, M. A. Shapiro and R. J. Temkin. This work is supported by the Office of Fusion Energy Sciences of the U. S. Department of Energy. [Preview Abstract] |
Wednesday, October 26, 2005 4:30PM - 5:00PM |
LI1b.00004: Optimization of Compact Stellarator Configuration as Fusion Devices Invited Speaker: Optimization of the stellarator configuration requires trade-offs among a large number of physics parameters and engineering constraints. An integrated study of compact stellarator power plants, ARIES-CS, aims at examining these trade-offs and defining key R{\&}D areas. We developed configurations with A$\le$6 and excellent QA (both 2 and 3 field periods) while reducing $\alpha$ losses to $\sim $10{\%} (still higher than desirable). Stability to the linear ideal MHD modes was attained but at the expense of reduced QA (and increased $\alpha$ losses) and increased complexity of the plasma shape. Recent experimental results indicate, however, linear MHD stability limits may not be applicable to stellarators. It appears that the plasma/coil stand-off distance is not as an important as envisioned previously. By utilizing a highly efficient shield-only region in strategic areas, we reduced the minimum stand-off by $\sim $20{\%}-30{\%}. This allows a comparable reduction in the machine size. The device configuration, assembly, and maintenance procedures appear to impose severe constraints. A cost-optimization system code has been developed and is utilized to guide the optimization process. [Preview Abstract] |
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