Bulletin of the American Physical Society
50th Annual Meeting of the Division of Plasma Physics
Volume 53, Number 14
Monday–Friday, November 17–21, 2008; Dallas, Texas
Session GO3: Research in Support of ITER |
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Chair: Charles Greenfield, General Atomics Room: Reunion A |
Tuesday, November 18, 2008 9:45AM - 9:57AM |
GO3.00001: NSTX ELM and RWM control experiments and modeling for ITER R.J. Hawryluk, S.A. Sabbagh, R. Maingi, J. Bialek, J.M. Canik, S.P. Gerhardt, J.E. Menard, J.-K. Park In light of the importance of mitigating edge localized modes (ELMs) in ITER, NSTX has recently studied the effects of resonant magnetic perturbations (RMPs) on ELM stability. Ideal Perturbed Equilibrium Code (IPEC) simulations indicate that the empirically determined ergodization criterion (Chirikov parameter $>$ 1 across pedestal) is readily achieved using NSTX external mid-plane RMP coils. However, NSTX experiments using a range of applied toroidal mode numbers indicate ELMs are modified but not stabilized in ELMy discharges and can be destabilized in long-lived ELM-free discharges. Importantly, RMP ELM-pacing can reduce impurity accumulation in ELM-free H-modes. No strong changes in the edge T$_{e}$ or n$_{e}$ profiles are observed during RMP, but the rotation is observed to decrease. These results and IPEC simulations are consistent with the need for both mid-plane and off-midplane RMP coils in ITER to minimize edge rotation damping. The proposed ITER RMP coil set is also predicted to be effective for RWM control, and VALEN simulations indicate that ITER Scenario 4 can be stabilized up to $\beta_{N}$ of 3.7 (well above no-wall limit of 2.5) with modest power and current requirements. [Preview Abstract] |
Tuesday, November 18, 2008 9:57AM - 10:09AM |
GO3.00002: Comparison of ELM Control Using One vs Two Rows of RMP Coils in DIII-D M.E. Fenstermacher, T.E. Evans, T.H. Osborne, M.J. Schaffer, J.S. deGrassie, R.J. Groebner, A.W. Leonard, P.B. Snyder, R.A. Moyer Large Type-I edge localized modes (ELMs), in plasmas with an ITER similar shape at the ITER pedestal collisionality, $\nu_e^*\sim 0.1$ and low edge safety factor ($q_{95}\approx 3.6$), were suppressed by $n=3$ resonant magnetic perturbations (RMPs) using either a single toroidal row or two poloidally separated rows of internal coils (I-coils). ELM suppression with a single row of internal coils was achieved at approximately the same $q_{95}$ surface-averaged perturbation field as with two rows of coils, but required higher current per coil. Maintaining complete suppression of ELMs using $n=3$ RMPs from a single toroidal row of internal coils was less robust to variations in input neutral beam injection torque than previous ELM suppression cases using both rows of internal coils. With either configuration of RMP coils, maximum ELM size is correlated with the width of the edge region having good overlap of the magnetic islands from vacuum field calculations. [Preview Abstract] |
Tuesday, November 18, 2008 10:09AM - 10:21AM |
GO3.00003: Isotope dependence of H-mode threshold and confinement in ASDEX Upgrade Matthias Reich, Rainer Fischer, Nathaniel Hicks, Bernd Kurzan, Thomas Puetterich, Francois Ryter, Elisabeth Wolfrum In view of the ITER low activation phase (H and He), it is important to predict the confinement properties and especially the L-to-H power threshold (P$_{thresh})$ for all working gases. Due to extensive studies in deuterium plasmas and fairly good documentation of hydrogen plasmas (in particular for ASDEX Upgrade), it is known that P$_{thresh}$ (H) $\sim $ 2 P$_{thresh}$ (D). Dedicated experiments to improve upon the existing database, especially upon the scarce helium data, have recently been conducted at ASDEX Upgrade. Now the database provides good coverage for all three gases (H, He and D). The data extend over a range from 1.5 T to 3 T in toroidal magnetic field and 3 10$^{19}$ to 8 10$^{19}$ m$^{-3}$ in density. An almost linear dependence of threshold on the magnetic field is consistently found across all three gases. P$_{thresh}$ and its density dependence in helium are very similar to that in deuterium. In both gases P$_{thresh}$ exhibits a minimum at about 4 10$^{19}$ m$^{-3}$. The new results from hydrogen beams into helium plasmas (as is the currently favored ITER first plasma scenario) and ECRH in helium and deuterium are compared to the previous results from deuterium and hydrogen plasmas. The influence of rotation on the power threshold is estimated from L-to-H transitions obtained with ECRH and NBI. [Preview Abstract] |
Tuesday, November 18, 2008 10:21AM - 10:33AM |
GO3.00004: H-mode Power Threshold, Pedestal and ELM Characteristics and Transport in Hydrogen Plasmas in DIII-D P. Gohil Comparing the physics of hydrogen plasmas with deuterium plasmas is important for the validation of key issues in plasma physics, such as ion mass dependences. This is vitally important for the first operational phase of ITER, which will use hydrogen plasmas. The issues for hydrogen plasmas include: (a) determining the H-mode power threshold and validating H-mode threshold scaling relations; (b) investigating the pedestal width and ELM characteristics; (c) examining plasma transport and turbulence behavior. Preliminary results with hydrogen plasmas and hydrogen neutral beam injection indicate that the H-mode power threshold is significantly higher than that for deuterium but decreases with decreasing applied beam torque (as seen in deuterium). Because of the power threshold dependence on the input torque, the threshold power for hydrogen plasmas with counter-injected beams is similar to the threshold power for deuterium plasmas with co-injected beams. Results from experiments to investigate issues (a-c) in hydrogen plasmas will be presented. [Preview Abstract] |
Tuesday, November 18, 2008 10:33AM - 10:45AM |
GO3.00005: ITER relevant current ramping experiments in Alcator C-Mod Adrianus Sips, Steve Wolfe, Ian Hutchinson, Charles Kessel, Yu Lin, Earl Marmar, Joseph Snipes, Steven Wukitch ITER requires routine operation at 15 MA within the operational constraints of the device. The original proposed poloidal coil-set is specified only for rather low plasma inductance (l$_{i}$=0.7-1.0). In C-Mod, the current rise and current decay phase of the ITER discharge scenario have been studied, trying to keep l$_{i}$ low. The experiments used early X-point formation during the current ramp-up, and remained diverted during the current ramp-down. Ohmic ramp up discharges achieved li as low as 0.9 at the start of the flat top, ICRF heated discharges only reduced l$_{i}$ by $\sim $ 0.05, despite raising T$_{e0}$ from 2 keV (ohmic) to 4 keV (3 MW ICRH). For the ramp-down, l$_{i}$ could be kept below 1.2 during the first half of the current decay, using a slow I$_{p}$ ramp-down rate still consuming flux from the transformer. These dedicated experiments are supported by interpretation of the results with TSC/TRANSP, and provide input for validating the transport models used in extrapolating the results to ITER. Supported by USDoE awards DE-FC02-99ER54512 and DE-AC02-76CH03073. [Preview Abstract] |
Tuesday, November 18, 2008 10:45AM - 10:57AM |
GO3.00006: ITER Vertical Stability Guidance from Multi-machine Experiments D.A. Humphreys, N. Eidietis, G.L. Jackson, J.A. Leuer, M.L. Walker, A.S. Welander, T.A. Casper, L.L. LoDestro, W.H. Meyer, L.D. Pearlstein, M. Ferrara, I.H. Hutchinson, S.M. Wolfe, D.A. Gates, E. Kolemen, J. Lister, F. Sartori, W. Treutterer Sufficiently robust vertical stability control is critical to ITER, in which the consequences of a vertical displacement event (VDE) disruption can be very severe. Experimental results from many devices have provided guidance to determine the necessary level of robustness, and theoretical analysis has quantified the tradeoffs inherent in various design choices. The maximum controllable displacement normalized by minor radius is shown to be a useful metric for performance, and must be greater than 4\% for robustness to VDEs in operating machines. Analysis of controllability limits, axisymmetric control performance, noise environments, and disturbances in operating devices including Alcator C-Mod, DIII-D, NSTX, TCV, and JET will be presented. [Preview Abstract] |
Tuesday, November 18, 2008 10:57AM - 11:09AM |
GO3.00007: Predictions of H-mode performance in ITER Robert Budny Time-dependent integrated predictions of performance metrics such as the fusion power P$_{DT}$, Q$_{DT}\equiv$ P$_{DT}/$P$_{ext}$, and alpha profiles are presented. The PTRANSP [1] code is used, along with GLF23 to predict plasma profiles, NUBEAM for NNBI and alpha heating, TORIC for ICRH, and TORAY for ECRH. Effects of sawteeth mixing, beam steering, beam shine-through, radiation loss, ash accumulation, and toroidal rotation are included. A total heating of P$_{ext}$=73MW is assumed to achieve H-mode during the density and current ramp-up phase. Various mixes of NNBI, ICRH, and ECRH heating schemes are compared. After steady state conditions are achieved, P$_{ext}$ is stepped down to lower values to explore high $Q_{DT}$. Physics and computation uncertainties lead to ranges in predictions for P$_{DT}$ and Q$_{DT}$. Physics uncertainties include the L$\rightarrow$H and H$\rightarrow$L threshold powers, pedestal height, impurity and ash transport, and recycling. There are considerably more uncertainties predicting the peak value for Q$_{DT}$ than for P$_{DT}$. \\[0pt] [1] R.V. Budny, R. Andre, G. Bateman, F. Halpern, C.E. Kessel, A. Kritz, and D. McCune, Nuclear Fusion {\bf 48} (2008) 075005. [Preview Abstract] |
Tuesday, November 18, 2008 11:09AM - 11:21AM |
GO3.00008: Demonstration of ITER Operational Scenarios on DIII-D P.A. Politzer The DIII-D program has recently begun an effort to provide experimental evaluation of the primary ITER operational scenarios, enabling direct cross-comparisons on a single tokamak. This work incorporates leading features of the ITER scenarios and anticipated operating characteristics. The plasma shape and aspect ratio in DIII-D match the ITER design (size reduced by a factor of 3.7), as does the value of I/aB. Key aspects of the ITER baseline ELMy H-mode (15~MA in ITER), advanced inductive (13~MA), hybrid (11~MA), and steady-state (9~MA) scenario plasmas have been replicated, providing a unified basis for transport and stability modeling and performance extrapolation. In all scenarios performance equals or closely approaches that required to realize the physics and technology goals of ITER. Baseline plasmas with normalized beta of 1.8-2.0 were studied (limited by tearing modes); for the other scenarios, the normalized beta was in the range 2.7-3.0. Confinement with $H_{98y2} > 1$ was seen in all cases. Significant differences from ITER assumptions include low internal inductance and peaked density profiles. [Preview Abstract] |
Tuesday, November 18, 2008 11:21AM - 11:33AM |
GO3.00009: Initial Operation of the ITER-Like Ion Cyclotron Antenna in JET R.H. Goulding, F.W. Baity, J. Caughman, D.A. Rasmussen, F. Durodi\'{e}, S. Huygen, E. Lerche, J. Ongena, D. Van Eester, M. Vrancken, T. Blackman, P. Jacquet, M. Nightingale, A. Argouarch The JET ITER-Like Ion Cyclotron Antenna (ILA) has been installed in JET and commissioning on plasma is underway. The antenna is a two toroidal by four poloidal strap array configured in four pairs fed in a ``conjugate-tee'' arrangement utilizing internal pre-matching capacitors, strongly limiting the VSWR rise caused by increases in loading due to ELMs. The use of several poloidal straps to minimize the voltage close to the plasma and an ``ELM resilient'' feed circuit are design features shared with the ITER ion cyclotron antenna, and operating experience gained with the JET antenna will be of great importance for maximizing ITER antenna performance. To date the antenna has been successfully matched under feedback control into varying plasma loads, and has been operated in an H-mode plasma at a maximum capacitor voltage of 38 kV. We will review antenna commissioning results in the areas of power handling, plasma loading, matching behavior, and performance of arc protection systems. [Preview Abstract] |
Tuesday, November 18, 2008 11:33AM - 11:45AM |
GO3.00010: Assessment on TF ripple Kouji Shinohara The TF ripple can reduce fast ion confinement and reduce plasma performance. Additionally, recent experiment results from JT-60U and JET imply that ripple can affect bulk plasma confinement. Here, the characteristics of TF ripple and its effect on fast ion confinement is reported. The TF ripple amplitude in ITER is $\sim $1.2{\%} in case of TF coil alone at the separatrix. This ripple will be compensated by ferritic inserts (FIs), placed between the vacuum vessel shells. In the current design, the ripple amplitude is 0.4{\%} at the separatrix at full field. It should be noted that, in this case, the ripple is overcompensated at half field and that the ripple amplitude is -0.4{\%}. A test blanket module (TBM), some of whose components are ferromagnetic, is another source of TF ripple. TBMs will be installed at three mid-plane ports. The TF ripple induced by TBMs is localized in the poloidal direction as well as in the toroidal direction. The ripple is $\sim $1{\%} at the mid-plane. Fast ion confinement was evaluated. The fast ion loss was small $<$2.5{\%} in the Scenario 2 and 4. However, at half field, the effectiveness of FI is reduced and the TBM enhances the loss; the loss is larger than half of that in case of TFC alone. [Preview Abstract] |
Tuesday, November 18, 2008 11:45AM - 11:57AM |
GO3.00011: Alpha particle loss studies in ITER with test blanket modules G.J. Kramer, R.B. White, R. Nazikian Alpha particles in ITER should be confined well during the burn phase of the discharge and not lost due to the effects of magnetic field ripple. The toroidal field ripple in ITER is designed to be small, 0.2\% or less with optimized feritic inserts. However, the insertion of three test blanket modules (TBMs) increases the field ripple quite substantially with a notable increase in the loss of fusion-born alpha particles, according to simulations performed with two particle orbit following codes ORBIT and SPIRAL. we have investigated the fusion-born alpha particle losses with and without the TBMs and found that the losses increase by a factor of two overall with a marked increase in the localization of the losses when the TBMs are present. The heat load on the first wall with optimized ripple (0.2\%) and no TBMs is spread uniformly over the outer wall due to the 18-fold symmetry of the toroidal field coil set. In the simulations with the TBMs, three intense hot spots are obtained with heat loads up to 1 MW/m$^2$ in the center of the TBMs. [Preview Abstract] |
Tuesday, November 18, 2008 11:57AM - 12:09PM |
GO3.00012: Impurity Assimilation During Massive Gas Injection for Disruption Mitigation in DIII-D E.M. Hollmann, A.N. James, J.H. Yu, T.C. Jernigan, T.E. Evans, D.A. Humphreys, P.B. Parks, E.J. Strait, M.A. Van Zeeland, J.C. Wesley, W.P. West, W. Wu Efficient assimilation of injected impurities into the plasma following massive gas injection (MGI) is desirable for rapid shutdown of future tokamaks. Experiments on the DIII-D tokamak with a variety of different valves and gas species have shown that MGI impurity assimilation is a dominated by magnetohydro-dynamics: when the cold front associated with the impurities reaches the $q=2$ rational surface, transport is accelerated due to low-order tearing modes, leading rapidly to the core thermal quench (TQ). Impurity mixing efficiencies up through the TQ are of order 10\%; mixing during the subsequent current quench (CQ) is slower. Extrapolation of DIII-D results suggest that MGI shutdowns in ITER would result in tolerably low wall heat loads and vessel forces, but achieving sufficient assimilation to guarantee suppression of runaway electrons appears to be difficult. Ongoing experiments on improvements and alternatives to MGI will be discussed. [Preview Abstract] |
Tuesday, November 18, 2008 12:09PM - 12:21PM |
GO3.00013: Studies of runaway electrons during disruptions in Alcator C-Mod R.S. Granetz, D.G. Whyte, V.A. Izzo The generation of large relativistic electron populations in ITER during the disruption current quench is of concern due to the potential for damage to the first wall and vacuum vessel. The runaway avalanche process can be quenched by the quick injection of large amounts of gas (Rosenbluth criterion requires $\sim10^5$ Pa-m$^3$), which has serious implications for the ITER cryopumps and tritium handling plant. Several present-day experiments suggest that other runaway loss mechanisms exist, implying that large gas injections may not be necessary. A program to study the physics of runaways in Alcator C-Mod disruptions uses LHCD as a tool to produce a seed population of superthermal electrons prior to triggering a disruption. In experiments to date, it is clear that substantial populations of superthermal and relativistic electrons can be produced during the flattop (0.5 MA in a 1 MA discharge), but during gas jet mitigated disruptions their loss rate during the thermal quench precludes any significant avalanching during the current quench phase. Modeling of the formation of ergodic field lines by gas jet injection in C-Mod with the NIMROD code supports these experimental findings. [Preview Abstract] |
Tuesday, November 18, 2008 12:21PM - 12:33PM |
GO3.00014: Neonlike tungsten ions as probes of ion and electron temperature and bulk motion of ITER plasmas Peter Beiersdorfer, J. Clementson, M.-F. Gu, Y. Podpaly, M. Bitter, K.W. Hill, A. Safronova, U. Safronova The core ion temperature of ITER plasmas will likely be derived from the thermal Doppler broadening of x-ray lines emitted by highly charged trace elements and recorded by an array of crystal spectrometers. Although it has been suggested to seed the plasma with krypton for this purpose, we show that the emission of neonlike tungsten provides several important advantages, especially if tungsten is already a plasma constituent due to its use in the divertor region. The relevant tungsten lines have wavelengths that are readily analyzed by x-ray crystals and fall into a region where existing detectors have high quantum efficiency. Moreover, the abundance of neonlike W$^{64+}$ peaks for the core electron temperatures expected. We will present both theoretical studies and experimental results of some of the relevant tungsten x-ray emission, including crystal spectrometer measurements of neonlike tungsten confined and excited in an electron beam ion trap. [Preview Abstract] |
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