Bulletin of the American Physical Society
53rd Annual Meeting of the APS Division of Plasma Physics
Volume 56, Number 16
Monday–Friday, November 14–18, 2011; Salt Lake City, Utah
Session NO4: Research in Support of ITER |
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Chair: Charles Greenfield, General Atomics Room: Ballroom E |
Wednesday, November 16, 2011 9:30AM - 9:42AM |
NO4.00001: Controlling ITER Scenarios J.A. Snipes, D.J. Campbell, T. Casper, Y. Gribov, S.-H. Kim, A. Winter The three principal ITER operational scenarios are the 15 MA, 5.3 T, Q=10 inductive scenario, the 1000 s, 10.5-13.5 MA, 5.3 T, Q$>$5 Hybrid scenario, and the 3000 s, 7.5-10 MA, 5.3 T, Q$\sim $5 steady-state scenario. Extensive modeling of the inductive scenario indicates that the ITER baseline actuators should be capable of all of the required basic control and that the ITER diagnostic specifications should provide adequate measurements with which to carry out such control. Current ramp-up times as short as 50 s and ramp-down times as low as 60 s are within control limits. Expected plasma disturbances can also be controlled. More advanced control is required for the hybrid and steady-state scenarios. Depending on transport assumptions, some modeling indicates that the baseline actuators should also be capable of achieving the ITER performance goals in the hybrid scenario with modest confinement improvement (H98$\sim $1.2). For steady-state scenarios, it is likely that substantial upgrades to the heating and current drive systems will be required to achieve the high performance and pulse length goals. High confinement (H98$\le $1.7) is also required to achieve these performance goals, challenging stability limits and requiring simultaneous control of multiple instabilities (e.g., ELMs, NTMs, RWMs) with limited shared actuators. The ITER Plasma Control System is being developed taking into account these challenging control requirements. [Preview Abstract] |
Wednesday, November 16, 2011 9:42AM - 9:54AM |
NO4.00002: Plasma Response and Transport Associated with RMP ELM Suppression on DIII-D M.R. Wade Recent experiments on DIII-D have focused on improving the understanding of the plasma response and associated transport changes leading to ELM suppression when resonant magnetic perturbation (RMPs) are applied. ELM suppression has been obtained in ITER Similar Shape discharges when applying either $n=3$ (either with two or a single row of internal coils) or $n=2$ RMP fields. While a substantive density decrease is generally observed, ELM suppression is limited to a narrow range in $q_{95}$ that is consistent with expectations based on vacuum field modeling. However, the $q_{95}$ ELM suppression range is $\beta$ dependent, suggesting sensitivity to the ideal MHD plasma response. Modulating the RMP has revealed nearly instantaneous responses in the edge toroidal rotation, radial electric field, and turbulence levels. Experiments are planned to differentiate the vacuum and ideal responses by varying the effective magnetic spectrum over a wide range utilizing a variety of tools (e.g., $n=2$, $n=3$ single-row vs two rows, single null vs double null) and to further characterize the edge profile and turbulence changes during fine-scale $q_{95}$ scans, with and without RMP modulation. [Preview Abstract] |
Wednesday, November 16, 2011 9:54AM - 10:06AM |
NO4.00003: Measuring Error Fields in ITER Before Its First Plasma M.G. Bell, N. Pomphrey, A.H. Boozer ITER is concerned about the possible impact of magnetic field errors on plasma operation and stability. It proposes to measure field errors due to imperfections and misalignments of the coils prior to the first plasma operation so that errors, particularly those with toroidal harmonic $n=1$, can be corrected with its set of non-axisymmetric coils. ITER presents challenges for such measurements because of its scale, its superconducting coils and the use of ferromagnetic inserts inside the toroidal field coil to reduce the intrinsic $n=18$ ripple. The sets of partial flux loops in ITER which span the inner poloidal circumference of the vacuum vessel at six toroidal locations provide a way to measure the field normal to the vacuum vessel surface. If appropriate metrology is performed on these loops after installation, it should be possible to determine from a data set in which all poloidal field coils are separately energized, the field errors with $n=1$ and poloidal harmonics $m=1-3$ in the plasma region. A method is proposed to correct for the possible influence of the ferromagnetic inserts. Measurements made with Hall-effect sensors at several locations {\em outside} the toroidal field coil together with NMR sensors inside the coil may provide the simplest way to characterize its field errors. [Preview Abstract] |
Wednesday, November 16, 2011 10:06AM - 10:18AM |
NO4.00004: Effect of 3-D fields on the divertor detachment in NSTX J-W. Ahn, R. Maingi, A.G. McLean, J.M. Canik, J.D. Lore, T.K. Gray, A. Diallo, M. Jaworski, B. LeBlanc, S. Kubota, V.A. Soukhanovskii, K. Tritz, A. Loarte Divertor plasma detachment was induced by divertor gas puffing in NSTX and is accompanied by a drop of pedestal electron temperature ($T_{e})$ and density ($n_{e})$, therefore pressure ($p_{e})$, and this drop becomes stronger, particularly in $T_{e}$, with increasing gas puffing rate. The applied 3-D fields were observed to reattach the detached plasma for lower gas puffing rates, but the divertor plasma remained detached with higher puffing rates even with the 3-D field application. The pedestal $T_{e}$ profile for the detached plasma at lower puff rates shows a clear increase during the 3-D field application while the pedestal density increase is not obvious. The 3-D field does not affect the pedestal profiles of the detached plasma at higher puff rates. The Ultra Soft X-ray (USXR) data indicates that the emission intensity first changes in the edge region, both by the detachment and 3-D field application, and then propagates into the pedestal region. The change of the transport processes both in the edge and core region across the detachment and the 3-D field application will be also investigated with diagnostic measurements available. This work was supported by the US Department of Energy, contract {\#} DE-AC05-000R22725, DE-AC02-09CH11466, DE-AC52-07NA27344, and DE-AC02-09CH11466. [Preview Abstract] |
Wednesday, November 16, 2011 10:18AM - 10:30AM |
NO4.00005: Overview of the JET ITER-like Wall, First Results and Scientific Programme Guy Matthews The ITER-like Wall (ILW) is the first integrated tokamak experiment with a beryllium main chamber wall and tungsten divertor as foreseen for the activated operational phase of ITER: The ILW will study plasma-wall interaction (PWI) processes (material erosion, material mixing etc.), and the compatibility of the ITER materials with low fuel retention and high power operation. Replacement of the JET CFC first wall by solid Be limiters, and a combination of bulk W and W-coated CFC divertor tiles was performed by remote handling and completed in May 2011 in parallel with a neutral beam heating upgrade to 35MW and enhancement of diagnostic capabilities. Mitigation of the power and energy loads in the divertor to acceptable levels at high power plasma performance will require high-density plasmas and radiative cooling via impurity seeding. Experiments were carried out with the carbon wall in preparation for the ILW to operate plasmas within ILW limits and provide reference plasmas for key physics studies. Although first plasma is scheduled for mid-August, the scientific programme in support of ITER will start earlier with machine conditioning. In this paper, an overview of the ILW, first results and the outlook for the scientific programme will be presented. [Preview Abstract] |
Wednesday, November 16, 2011 10:30AM - 10:42AM |
NO4.00006: Growth of tungsten nano-tendrils in the Alcator C-Mod lower divertor G.M. Wright, D. Brunner, B. LaBombard, B. Lipschultz, J.L. Terry, D.G. Whyte The conditions for the growth of tungsten (W) nano-tendrils have been well defined in linear plasma devices (T$_{surface}>$900 K, $\Gamma _{He+}>$10$^{22}$ m$^{-2}$s$^{-1}$, 10 eV$<$E$_{He+}<$150 eV) but, until now, there has been no documented nano-tendril growth in a tokamak environment. We have exploited the high power density and all-metal wall in Alcator C-Mod to successfully grow W nano-tendrils on a Langmuir probe (ramped $\sim $10$^{o}$ into the parallel plasma flux) in the lower divertor during a single run day. Scanning electron microscopy shows fully formed nano-tendrils over the surface of the probe after an estimated 15-30 seconds of growth time. Having shown that these nano-tendrils can form in a tokamak divertor and given that the growth conditions are met in an all-W ITER divertor during the DT phase, this provides strong evidence that these types of ``fuzzy'' surfaces will be present in the ITER divertor. The effects of this extreme surface morphology on plasma-surface interactions remains unclear, but possible implications for ITER operation will be discussed. [Preview Abstract] |
Wednesday, November 16, 2011 10:42AM - 10:54AM |
NO4.00007: Shifting the CFC/W transition point on the first ITER divertor target plates: the effect on ITER plasmas R.A. Kolesnikov, R.H. Bulmer, L.L. LoDestro, T.A. Casper, R.A. Pitts In the 2007 ITER Design Review, the CFC/W transition point on the first ITER divertor target plates was lowered by 10 cm to allow some experience to be gained in the non-active phases of vertical target operation with strike points on W surfaces, in preparation for a full W divertor in the nuclear phase. In this work, we use the CORSICA code to investigate the range of possible H-mode equilibria with this lowered transition, with emphasis on the maximum current, achievable shapes, etc. We then investigate the operational space as the transition is lowered still further (both L- and H-mode), while still ensuring sufficient carbon vertical target extent to fulfill the requirements of the non-active phase program (e.g., operation at reference 15 MA currents). The primary aim of this feasibility/sensitivity study is to determine if the current transition point, which can still be modified within some range if required, is optimized with respect to gaining early operational experience on an all-metal target before the nuclear phases begin. Also, we study how such equilibria affect the range of plasma self-inductance and volt-second consumption. [Preview Abstract] |
Wednesday, November 16, 2011 10:54AM - 11:06AM |
NO4.00008: Study of deuterium removal from co-deposited beryllium layers by flash heating Jonathan Yu, Russ Doerner, Eric Hollmann, Richard Pitts, Karl Umstadter Tritium removal from plasma facing components in ITER will be achieved by standard bakeout (assuming most deposition in the divertor), but a degree of main chamber de-tritiation might be possible with flash heating during rapid plasma shutdown. This main chamber de-tritiation possibility has been investigated in the laboratory, and results suggest that a rapid shutdown radiation flash will not yield significant fuel outgassing from main chamber Be deposits. The linear PISCES-B facility is used to create $\sim $100 nm thick Be co-deposit layers on a collection plate located outside a deuterium (D) plasma using Be sputtered from a plasma-exposed target. The co-deposit is then illuminated with a 1064 nm laser with ITER-relevant disruption photon flash energy densities and pulse length, and the remaining D content is measured using thermal desorption spectroscopy. Less than 25{\%} of the trapped D escapes from a Be co-deposit even when the flashed sample temperature exceeds 1000 K. [Preview Abstract] |
Wednesday, November 16, 2011 11:06AM - 11:18AM |
NO4.00009: ITER-like Discharge Development in Alcator C-Mod C.E. Kessel, S. Wolfe, I. Hutchinson, A.E. Hubbard, J.W. Hughes, Y. Lin, S. Wukitch Demonstrating discharges on Alcator C-Mod with ITER characteristics is important to study plasma behavior during various phases and validate modeling used to project to ITER. Concentration has been on the rampup and rampdown phases. The flattop phase must meet, as close as possible, a number of parameters simultaneously; q95, elongation, n/nGr, beta-N, and H98. Experiments were performed to meet these parameters, lowering the toroidal field to 2.7 T and using 2$^{nd}$ harmonic hydrogen minority heating. The lower field allowed more reliable access to these parameters. These discharges meet the ITER parameters closely, with the n/nGr value reaching 0.72 approaching the ITER value of 0.85, and were sustained for 0.5 to 1 s. EDA H-modes were obtained, showing the quasi-coherent mode at about 100 kHz, with some intermittent ELMy behavior. In addition, MHD modes are observed in the 10-25 kHz range with toroidal mode numbers n=2,3, which appear to be correlated with increasing betaN. Work supported by DE-AC02-09CH11466 and DE-FC02-99ER54512. [Preview Abstract] |
Wednesday, November 16, 2011 11:18AM - 11:30AM |
NO4.00010: Impact of different heating and current drive mixes on steady-state scenarios for ITER M. Murakami, J.M. Park, L.L. Lao, T.C. Luce, R. Prater, H.E. St. John, P.T. Bonoli Impact of a range of different sets of heating and current drive mixes on the ITER steady-state scenarios are examined exploiting an iterative steady state solution procedure using a new fast transport solver FASTRAN utilizing the ONETWO and EFIT codes. There is trade off between Q and f{\_}NI,$_{ }$as in the I{\_}P scan (8 -- 10 MA): optimization of 8MA scenarios lead to f{\_}NI =100{\%} and Q \underline {$<$} 4.5, while optimization of 9-MA scenarios lead to f{\_}NI =95{\%} and Q \underline {$<$} 5.3 using day-1 baseline H{\&}CD capability, These values are close, but still somewhat short in simultaneously achieving the Q = 5 and f{\_}NI = 100{\%}. Upgrades of ECCD (with TORAY/CQL3D for parallel momentum conservation effects) considered include the Upper Steering Mirror (USM) and Equatorial Launcher Top Steering Mirror (EL-TSM) systems for current profile control and (2) doubling the total EC power to 40 MW. Effects of different density, density peaking, q{\_}min and transport models will also be discussed. [Preview Abstract] |
Wednesday, November 16, 2011 11:30AM - 11:42AM |
NO4.00011: I-mode for ITER? D.G. Whyte, E. Marmar, A. Hubbard, J. Hughes, A. Dominguez, M. Greenwald I-mode is a recently explored confinement regime that features a temperature pedestal and H-mode energy confinement, yet with L-mode particle confinement and no density pedestal nor large ELMs. Experiments on Alcator C-Mod and ASDEX-Upgrade show this leads to a stationary collisionless pedestal that inherently does not require ELMs for core impurity and particle control, possibly making I-mode an attractive operating regime for ITER where ELM heat pulses are expected to surpass material limits. We speculate as to how I-mode could be obtained, maintained and exploited for the ITER burning plasma physics mission. Issues examined include I-mode topology and power threshold requirements, pedestal formation, density control, avoiding H-mode, and the response of I-mode to alpha self-heating. Key uncertainties requiring further investigation are identified. [Preview Abstract] |
Wednesday, November 16, 2011 11:42AM - 11:54AM |
NO4.00012: Shape and Current Profile Effects on Runaway Electron Confinement V.A. Izzo, A.N. James, D.A. Humphreys, R.S. Granetz, D.G. Whyte, G.M. Olynyk The potential for several MA of current carried by multi-MeV runaway electrons (REs) during ITER disruptions has motivated a variety of experiments in present-day tokamaks studying RE generation, confinement, and control. In both DIII-D and Alcator C Mod, different RE behavior is seen in limited vs. diverted plasmas, suggesting better RE confinement for limited shapes. NIMROD simulations of rapid shutdowns in both devices support this finding, and show reduced stochasticity in limited plasma shapes. Integration of RE drift-orbits also shows differences in RE strike-points that are consistent with experimental observations. In DIII-D a wide variation in RE confinement results for diverted discharges may also point to current density profile effects on RE confinement. Several DIII-D diverted discharges are modeled with NIMROD. Confined RE fractions found in NIMROD are mostly consistent with observed RE currents in DIII-D, although other effects, such as seed generation and avalanching may contribute to the experimental variation. [Preview Abstract] |
Wednesday, November 16, 2011 11:54AM - 12:06PM |
NO4.00013: Operating ITER Robustly Without Disruptions D.A. Humphreys, N.W. Eidietis, A.W. Hyatt, J.A. Leuer, T.C. Luce, E.J. Strait, M.L. Walker, A.S. Welander, J.C. Wesley, L. LoDestro, L.D. Pearlstein Disruptivity in ITER must be minimized to limit downtime and maximize use of the limited number of discharges. Minimizing disruptivity requires sufficient control capability, including robustness to disturbances and disruption avoidance through prediction of controllability limits. Robust control implies a balance of passively stable nominal scenarios, robust operation near or beyond open loop stability limits, and responses to off-normal events to avoid disruptive termination. Such a solution is possible because disruptions result from deterministic loss of controllability due to many proximal causes (e.g.\ loss of hardware resources, human error, or uncontrollable disturbances), most of which can be addressed with good physics models and known control methods. We illustrate the required approach with DIII-D experiments to assess ITER controllability and pre-qualify ITER scenarios, and with design and analysis ensuring sufficiently robust vertical control for ITER. [Preview Abstract] |
Wednesday, November 16, 2011 12:06PM - 12:18PM |
NO4.00014: Inter-ELM power decay length for ASDEX Upgrade, JET and ITER Thomas Eich, Bernhard Sieglin, Andrea Scarabosio, Wojtek Fundamenski, Robert Goldston, Albrecht Herrmann Experimental measurements of the SOL power decay length ($\lambda_q$) estimated from analysis of fully attached divertor heat load profiles from two tokamaks, JET and ASDEX Upgrade, are presented. Data was measured by means of infrared thermography. An empirical scaling reveals parametric dependency $\lambda_q/{\rm mm} = 0.73 \cdot B_{T}^{-0.78}q_{cyl}^{1.2}P_{SOL}^{0.1}R_{geo}^{0}$. A comparison of these measurements to a heuristic particle drift-based model(R.J.GOLDSTON, www.pppl.gov/pub\_report//2011/PPPL-4604.pdf) shows satisfactory agreement in both absolute magnitude and scaling. Extrapolation to ITER gives $\lambda_q\simeq$1 mm decay length and $\lambda_{int}\simeq$2.4 mm integral width. [Preview Abstract] |
Wednesday, November 16, 2011 12:18PM - 12:30PM |
NO4.00015: Plans for ECE diagnostic components for ITER M.E. Austin, H.K.B. Pandya, R.F. Ellis, R. Feder, A.E. Hubbard, W.L. Rowan, P.E. Phillips, S. Danani, S. Kumar Designs have been developed for the front end optics, port plug components, transmission line system and instrumentation for the ITER ECE diagnostics. Recent work has focused on the transmission line. The ITER ECE diagnostic will require a broadband low-loss transmission system to propagate radiation from the port plug to the diagnostic hall. A prime candidate for this is a corrugated waveguide line; however there are concerns of losses at high frequencies due to mode conversion and Bragg scattering. To better understand these loss mechanisms a study of the DIII-D corrugated waveguide system has been undertaken. By measuring with the DIII-D Michelson interferometer the losses and comparing them to theoretical calculations it will be possible extrapolate to the expected performance of an ITER system. Also, a comparison with other options for transmission systems, such as dielectric waveguide, will be shown. [Preview Abstract] |
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