Bulletin of the American Physical Society
54th Annual Meeting of the APS Division of Plasma Physics
Volume 57, Number 12
Monday–Friday, October 29–November 2 2012; Providence, Rhode Island
Session UO7: ITER and Magnetic Fusion Development |
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Chair: Amanda Hubbard, Massachusetts Institute of Technology Room: 556AB |
Thursday, November 1, 2012 2:00PM - 2:12PM |
UO7.00001: USBPO Disruption Task: Critical Issues and Research Needs for the ITER DMS J.C. Wesley Research needed to support the 2016 Final Design Review for the ITER Disruption Mitigation System (DMS) are identified and presented in terms of four critical-issue-driven topics: 1) thermal energy (TE) mitigation and disposition to the first wall without melting, 2) control of the current quench rate following TE mitigation, 3) avoidance of runaway electron (RE) avalanching following TE mitigation, and 4) control and benign dissipation of REs from unmitigated disruptions and/or from REs generated if the high-density pre-emptive collisional mitigation method proposed for (3) proves infeasible. Issue-driven R\&D logic, test and integration strategies and critical hardware, diagnostic and theory/modeling needs are identified and a USBPO framework for implementation proposed. Opportunities for contributions from U.S. and international facilities and the time-critical need for coordinated research activities and test of candidate mitigation methods and hardware with the most ``ITER-like'' conditions available will be discussed. [Preview Abstract] |
Thursday, November 1, 2012 2:12PM - 2:24PM |
UO7.00002: Disruption Mitigation Experiments with Two Gas Jets on Alcator C-Mod, and Implications for ITER R.S. Granetz, G.M. Olynyk, M.L. Reinke, D.G. Whyte, J.L. Terry, B.L. Lipschultz, S.K. Combs Until recently, all disruption mitigation experiments with massive gas injection have used one injection location at a time, and measurements have shown that the resulting radiated power is often toroidally asymmetric, which could lead to melting of first wall surfaces in ITER. Therefore, the proposed ITER MGI system has multiple gas jets distributed around the torus, but the effectiveness of this needs to be demonstrated on current machines. On Alcator C-Mod, a 2nd gas jet has been installed 154$^{\circ}$ around the torus from the existing gas jet. The hardware components of both gas jets are nominally identical. A toroidally-distributed set of six AXUV detectors has been added to better measure the toroidal peaking factor (TPF) of the radiated power. Experiments have been carried out to characterize the effect of two jets on the radiation TPF by varying the relative timing between the firing of the gas jets shot-to-shot. Measurements of the radiation asymmetry during the pre-thermal quench, thermal quench, and current quench phases will be presented, as well as correlations with the growth rate of n=1 MHD modes. It will also be shown that very slight differences in hardware between the two gas jet systems are important. Implications for the ITER MGI system will be discussed. [Preview Abstract] |
Thursday, November 1, 2012 2:24PM - 2:36PM |
UO7.00003: First Production of C$_{60}$ Nanoparticle Plasma Jet for Study of Disruption Mitigation for ITER I.N. Bogatu, J.R. Thompson, S.A. Galkin, J.S. Kim, S. Brockington, A. Case, S.J. Messer, F.D. Witherspoon Unique fast response and large mass-velocity delivery of nanoparticle plasma jets (NPPJs) provide a novel application for ITER disruption mitigation, runaway electrons diagnostics and deep fueling. NPPJs carry a much larger mass than usual gases. An electromagnetic plasma gun provides a very high injection velocity (many km/s). NPPJ has much higher ram pressure than any standard gas injection method and penetrates the tokamak confining magnetic field. Assimilation is enhanced due to the NP large surface-to-volume ratio. Radially expanding NPPJs help achieving toroidal uniformity of radiation power. FAR-TECH's NPPJ system was successfully tested: a coaxial plasma gun prototype ($\sim$35 cm length, 96 kJ energy) using a solid state TiH$_{2}$/C$_{60}$ pulsed power cartridge injector produced a hyper-velocity ($>$4 km/s), high-density ($>$10$^{23}$~m$^{-3})$, C$_{60}$ plasma jet in $\sim$0.5 ms, with $\sim$1$-$2 ms overall response-delivery time. We present the TiH$_{2}$/C$_{60}$ cartridge injector output characterization ($\sim$180 mg of sublimated C$_{60}$ gas) and first production results of a high momentum C$_{60}$ plasma jet ($\sim$0.6 g$\cdot$km/s). [Preview Abstract] |
Thursday, November 1, 2012 2:36PM - 2:48PM |
UO7.00004: NTM Suppression and Avoidance at DIII-D Using Real-Time Mirror Steering E. Kolemen, R.A. Ellis, R.J. La Haye, J. Lohr, S. Noraky, B.G. Penaflor, A.S. Welander The Electron Cyclotron Current Drive (ECCD) real-time steerable mirrors at DIII-D were developed and successfully operated to avoid and suppress neoclassical tearing modes (NTM). The NTM avoidance/suppression control logic moves the deposition location of the ECCD with six sets of real-time steerable mirrors in order to align it with the NTM location. The steerable mirrors enable changing the deposition location at approximately 2 m/s with accuracy of a few millimeters while keeping the plasma radial position and the toroidal field constant during NTM avoidance/suppression. The real-time system enables simultaneous avoidance/suppression of multiple magnetic islands (such as $m/n=2/1$ and 3/2 islands). [Preview Abstract] |
Thursday, November 1, 2012 2:48PM - 3:00PM |
UO7.00005: Validation of quasilinear models for fast ion relaxation due to Alfv\'en Eigenmodes for burning plasmas Nikolai Gorelenkov We offer and validate reduced quasi-linear models to describe the relaxed fast ion (FI) profiles expected in the presence of Alfv\'enic modes in tokamaks and apply it to projected ITER plasma conditions. The focus of this presentation is on the 1.5D model, which is being applied and validated against recent DIII-D experiments. These experiments were well diagnosed for a number of validating goals. With the parametric dependencies embedded in the presented analytic model and with the quantitative normalization coming from NOVA-K modeling the 1.5D diffusion is successfully validated against that data. The agreement is achieved for the absolute values and for the time behavior of the fast ion losses as the AE activity approaches the threshold conditions. Given its validation the model can be applied to future burning plasma experiment, such as ITER, with or without the NOVA-K-like numerical growth rate normalization. Moreover, more complete, 2D, approach to the quasi-linear diffusion is being developed, which covers a variety of cases in term of the instability excitation when the unstable modes exhibit themselves as isolated modes, partially isolated modes and completely overlapped modes. [Preview Abstract] |
Thursday, November 1, 2012 3:00PM - 3:12PM |
UO7.00006: Monostatic density profile reflectometry measurements on DIII-D and relevance to ITER low-field-side reflectometer W.A. Peebles, C. Wannberg, X. Nguyen, N.A. Crocker, L. Zeng, T.L. Rhodes, E.J. Doyle, G. Hanson, T. Bigelow, J. Wilgen Typically, density profile reflectometry systems employ bi-static antenna configurations to avoid the deleterious effects caused by spurious reflections. However, there are potential advantages, particularly on ITER, if a monostatic antenna configuration could be employed. Such a configuration would allow either a reduction in the number of waveguides necessary to satisfy measurement requirements or an expansion in capability assuming a fixed number of antennas. To address the feasibility of this configuration, a monostatic reflectometer operating from 33 to 75~GHz in both O-and X-mode has been designed, fabricated, installed and tested on DIII-D. Preliminary results appear positive with measured profiles in good agreement with Thomson scattering. More extensive tests are underway, including investigation of the ability to accommodate plasma height variations using a fixed antenna. System design and preliminary results will be presented. [Preview Abstract] |
Thursday, November 1, 2012 3:12PM - 3:24PM |
UO7.00007: Laser cleaning of ITER's diagnostic mirrors C.H. Skinner, C.A. Gentile, R. Doerner Practical methods to clean ITER's diagnostic mirrors and restore reflectivity will be critical to ITER's plasma operations. We report on laser cleaning of single crystal molybdenum mirrors coated with either carbon or beryllium films 150 - 420 nm thick. A 1.06 $\mu$m Nd laser system provided 220 ns pulses at 8 kHz with typical power densities of 1-2 J/cm$^{2}$. The laser beam was fiber optically coupled to a scanner suitable for tokamak applications. The efficacy of mirror cleaning was assessed with a new technique that combines microscopic imaging and reflectivity measurements [1]. The method is suitable for hazardous materials such as beryllium as the mirrors remain sealed in a vacuum chamber. Excellent restoration of reflectivity for the carbon coated Mo mirrors was observed after laser scanning under vacuum conditions. For the beryllium coated mirrors restoration of reflectivity has so far been incomplete and modeling indicates that a shorter duration laser pulse is needed. No damage of the molybdenum mirror substrates was observed.\\[4pt][1] C.H. Skinner et al., Rev. Sci. Instrum. at press. [Preview Abstract] |
Thursday, November 1, 2012 3:24PM - 3:36PM |
UO7.00008: Demonstrating the Physics Basis for the ITER 15 MA Inductive Discharge on Alcator C-Mod C.E. Kessel, S.M. Wolfe, I.H. Hutchinson, J.W. Hughes, Y. Lin, Y. Ma, D.R. Mikkelsen, F. Poli, M.L. Reinke, S.J. Wukitch Rampup discharges in C-Mod, matching ITE's current diffusion times show ICRF heating can save V-s but results in only weak effects on the current profile, despite strong modifications of the central electron temperature. Simulation of these discharges with TSC, and TORIC for ICRF, using multiple transport models, do not reproduce the temperature profile evolution, or the experimental internal self-inductance li, by sufficiently large amounts to be unacceptable for projections to ITER operation. For the flattop phase experiments EDA H-modes approach the ITER parameter targets of q95=3, H98=1, n/nGr=0.85, betaN=1.7, and k=1.8, and sustain them similar to a normalized ITER flattop time. The discharges show a degradation of energy confinement at higher densities, but increasing H98 with increasing net power to the plasma. For these discharges intrinsic impurities (B, Mo) provided radiated power fractions of 25-37\%. Experiments show the plasma can remain in H-mode in rampdown with ICRF injection, the density will decrease with Ip while in the H-mode, and the back transition occurs when the net power reaches about half the L-H transition power. C-Mod indicates that faster rampdowns are preferable. \textit{Work supported by US Dept of Energy under DE-AC02-CH091146}6 \textit{and DE-FC02-99ER54512.} [Preview Abstract] |
Thursday, November 1, 2012 3:36PM - 3:48PM |
UO7.00009: Stability boundaries and development of the ITER baseline scenario G.L. Jackson, T.C. Luce, R.J. Buttery, A.W. Hyatt, J.R. Ferron, R.J. La Haye, P.A. Politzer, W.M. Solomon, F. Turco, E.J. Doyle Plasmas stable to $m/n=2/1$ tearing modes (TMs), in the ITER baseline scenario (IBS) with ITER similar $I_p/aB_T$, have been demonstrated in DIII-D, evolving to stationary conditions. Previous studies showed the possibility that long pulse IBS plasmas might be susceptible to TMs. However within a defined stability boundary, most of these longer duration discharges have achieved stationary conditions ($\Delta\tau_{duration}\leq 7.5$~s and $\leq 11\tau_R$) with high Co-beam torque and without the need for ECCD. To mitigate 2/1 TMs at reduced torque, broad ECCD deposition was found to be most effective when positioned near the $q=3/2$ flux surface, although a subset of low torque pulses were also obtained without ECCD. The DIII D internal coils (I coils) were used to achieve ELM suppression in IBS plasmas with ECCD at $q_{95}=3.15$ for durations up to 1 s with only the upper row of I coils, providing a broad $n=3$ spectrum. Conditions stable to $n=1$ tearing modes in IBS discharges and the effect of $Z_{eff}$, density, and other parameters are discussed. [Preview Abstract] |
Thursday, November 1, 2012 3:48PM - 4:00PM |
UO7.00010: H-mode threshold physics studies on Alcator C-Mod in support of ITER Y. Ma, J.W. Hughes, B. LaBombard, A.E. Hubbard, E.S. Marmar, D.C. McDonald Comprehensive studies on H-mode threshold physics have been conducted on Alcator C-Mod tokamak, covering many ITER-relevant conditions, $e.g.$ similar magnetic field and density range, metallic wall, and divertor configuration. C-Mod experiments confirm that the density dependence of H-mode threshold power ($P_{th})$ is U-shaped without clear dependence on plasma current, and the local minimum of $P_{th}$ in density ($n_{min})$ decreases as $B_{T}$ is reduced [1]. An effect of divertor geometry on $P_{th}$ was identified, with a dramatic ($\sim $50{\%}) reduction in $P_{th}$ seen in ``slot'' divertor operation accompanied by longer SOL connection length [2]. Experimental results were also compared with a new physics-based model for $P_{th }$[3], showing reasonable agreement of density, $B_{T}$, and divertor geometry dependences with model predictions. A significant implication of this model is that $n_{min}$ occurs as the SOL transitions from sheath-limited to conduction-limited regime, which also seems to agree with experiments. Supported by USDoE award DE-FC02-99ER54512. \\[4pt] [1] Y.Ma, \textit{et al} Nucl. Fusion \textbf{52} (2012) 023010.\\[0pt] [2] Y.Ma, \textit{et al} PPCF \textbf{54} (2012) 082002.\\[0pt] [3] W.Fundamenski, \textit{et al} Nucl. Fusion \textbf{52} (2012) 062003. [Preview Abstract] |
Thursday, November 1, 2012 4:00PM - 4:12PM |
UO7.00011: Advances in optimization and uncertainty quantification of ITER scenarios Alexei Pankin, Scott Kruger, John Cary, Arnold Kritz, Tariq Rafiq Determination of plasma conditions that lead to improved confinement is extremely important for efficient ITER operation. Results for optimization of ITER performance is frequently obtained through heuristic predictive modeling using reduced transport models. Limited subsets of input parameters, typically associated with a mixture of heating mechanisms, are generally considered. Prediction uncertainty associated with the extrapolation of reduced models to new parameter space is usually not addressed. Recent improvements in computational capabilities and numerical algorithms as well as the development of new techniques for sensitivity analysis, uncertainty quantification, and optimization can be used to bring the robustness of ITER scenario modeling to a new level. The development of an approach that utilizes these techniques for optimization and uncertainty quantification of ITER performance is considered. In this study the DAKOTA toolkit for uncertainty quantification and optimization is used to optimize the plasma energy confinement time of ITER discharges by controlling the H-mode pedestal parameters. The effect of uncertainty in pedestal width predictions on the fusion power production is quantified and reported. [Preview Abstract] |
Thursday, November 1, 2012 4:12PM - 4:24PM |
UO7.00012: Testing TGLF in ITER-demonstration plasmas and predictions for ITER Robert Budny, Xingqiu Yuan, Gary Staebler An important goal for ITER and other future Tokamaks is achieving burning plasma conditions. To aid in achieving this goal, predictions of plasma performance are needed to facilitate design of scenarios, heating and current drive systems, and diagnostics. PTRANSP [1] is being used to make integrated, self-consistent, and time-dependant predictions for ITER. Plasma profiles have been predicted using the GLF23 [2] model. An improved model, TGLF [3] has been installed in PTRANSP. This presentation discusses the implementation and testing using ITER-demonstration plasmas such as from JET and DIII-D experiments. Examples include H-mode plasmas at high current. The TGLF simulations are compared with those from GLF23. PTRANSP time-dependant predictions for ITER are given.\\[4pt] [1] R.V. Budny, Nuclear Fusion {\bf 52} (2012) 013001.\\[0pt] [2] R. Waltz, {\it et al.,} Phys. Plasmas {\bf 4} (1997) 2482.\\[0pt] [3] J. Kinsey, {\it et al.,} Nuclear Fusion {\bf 51} (2011) 083001. [Preview Abstract] |
Thursday, November 1, 2012 4:24PM - 4:36PM |
UO7.00013: Error Field Assessment from Driven Mode Rotation: Results from Extrap-T2R Reversed-Field-Pinch and Perspectives for ITER F.A. Volpe, L. Frassinetti, P.R. Brunsell, J.R. Drake, K.E.J. Olofsson A new ITER-relevant non-disruptive error field (EF) assessment technique not restricted to low density and thus low beta was demonstrated at the Extrap-T2R reversed field pinch. Resistive Wall Modes (RWMs) were generated and their rotation sustained by rotating magnetic perturbations. In particular, stable modes of toroidal mode number n=8 and 10 and unstable modes of n=1 were used in this experiment. Due to finite EFs, and in spite of the applied perturbations rotating uniformly and having constant amplitude, the RWMs were observed to rotate non-uniformly and be modulated in amplitude (in the case of unstable modes, the observed oscillation was superimposed to the mode growth). This behavior was used to infer the amplitude and toroidal phase of n=1, 8 and 10 EFs. The method was first tested against known, deliberately applied EFs, and then against actual intrinsic EFs. Applying equal and opposite corrections resulted in longer discharges and more uniform mode rotation, indicating good EF compensation. The results agree with a simple theoretical model. Extensions to tearing modes, to the non-uniform plasma response to rotating perturbations, and to tokamaks, including ITER, will be discussed. [Preview Abstract] |
Thursday, November 1, 2012 4:36PM - 4:48PM |
UO7.00014: Test Blanket Module Mockup Experiments in DIII-D E.J. Strait, N.H. Brooks, R.J. Buttery, R.J. La Haye, M.J. Schaffer, H. Reimerdes, J.A. Snipes, J.M. Hanson, W.W. Heidbrink, Y. Zhu, Y. In, G.J. Kramer, M. Okabayashi, J.-K. Park, W.M. Solomon, A.G. McLean, N. Oyama, T. Tala Recent DIII-D experiments have investigated the effects of localized magnetic field perturbations, using coils that approximate the magnetization of the test blanket modules (TBMs) in one ITER port. In H-mode discharges, compensation of the TBM field using an applied $n=1$ field yielded only partial recovery of the plasma rotation, and the compensation field that maximized plasma rotation differed significantly from the field that reduced the resonant magnetic response to a very low value. These results provide insight into the effects of error fields, and suggest an important role for non-resonant magnetic braking. In addition, measurements of localized heat deposition with the TBM field are being compared to orbit following calculations of fast ion loss, and a new fast ion detector has confirmed earlier observations of reduced 1 MeV triton confinement. [Preview Abstract] |
Thursday, November 1, 2012 4:48PM - 5:00PM |
UO7.00015: 3D Vacuum Magnetic Field Modeling of the ITER ELM Control Coils During Standard Operating Scenarios T.E. Evans, W. Wu, D.M. Orlov, A. Wingen, A. Loarte, T.A. Casper, O. Schmitz, G. Saibene ELM coil optimization and failure studies have been completed for 9 standard ITER operating scenarios based on vacuum island overlap width calculations. Here, the toroidal phase of the current in the upper and lower coils is scanned in 2$^\circ$ steps while the current in the center ELM coil is held constant. The minimum current needed to satisfy the DIII-D ELM suppression correlation criterion varies from 20 to 50 kAt depending on the ITER operating scenario. In general, as the coil current is increased above the minimum required to meet the DIII-D criterion the available phase angle operating space increases approximately linearly with current. The DIII-D criterion can be satisfied, in the most demanding ITER scenario, with $n=3$ perturbation fields and failures in up to 8 of the full 27 coils. In this case, the available phase angle operating space is reduced from 79\% with no failures to 27\% with 8 failures by increasing the current in the remaining ELM coils to the maximum operating current of 90 kAt. Details of these results will be discussed along with plans to extend the analysis to include the plasma response to the perturbation field. [Preview Abstract] |
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