Bulletin of the American Physical Society
56th Annual Meeting of the APS Division of Plasma Physics
Volume 59, Number 15
Monday–Friday, October 27–31, 2014; New Orleans, Louisiana
Session UO3: Research in Support of ITER |
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Chair: Francesca Poli, Princeton Plasma Physics Laboratory Room: Salon D |
Thursday, October 30, 2014 2:00PM - 2:12PM |
UO3.00001: Time-resolved kinetic modelling of ELM-induced tungsten influx in ITER Steven Lisgo, James Harrison, Martin Kocan, Richard Pitts, Steffen Potzel, Detlev Reiter, Peter Stangeby High performance operation in ITER (Q $\sim$ 10) will require tungsten (W) core concentrations below $\sim$ 10$^{-5}$. The steady-state influx of W from the strike-points will be nominal since only detached plasmas can satisfy the engineering heat-flux limit of 10 MW m$^{-2}$, but high energy particles reaching the target plates during Edge-Localized Modes (ELMs) will exceed the W sputtering threshold. Given the very low W concentration limit for the core, operational planning requires that the production and transport of W in the boundary plasma be assessed for controlled ELMs and infrequent natural Type-I ELMs, and in the absence and presence of resonant magnetic perturbations (RMPs). ELM simulations with the SOLPS plasma fluid code were recently performed, where prompt redeposition was found to reduce the W influx by more than an order of magnitude [D. Coster et al, 40th EPS, 2014]. The present study employs the OSM-EIRENE-DIVIMP code package, which utilizes an empirical fluid model to describe the bulk plasma evolution and W is treated kinetically. Model benchmarks against experimental data are presented. RMPs will be addressed in future work. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization. [Preview Abstract] |
Thursday, October 30, 2014 2:12PM - 2:24PM |
UO3.00002: Revisited ELM divertor heat load scaling to ITER with JET and ASDEX Upgrade data Thomas Eich, Bernhard Sieglin, Albrecht Herrmann, Guy Matthews, Marc Beurskens, Alberto Loarte The divertor heat load due to type-I ELMs is a major concern for ITER. Both the suppression of ELM losses and the scaling of ELM induced heat loads is hence a major research topic at the various operating tokamaks. The ELM energy density was calculated by assuming a relative ELM loss in ITER of few percent and a moderate broadening of the wetted area. The absolute numbers entering the scaling are the stored energy and the inter-ELM wetted area. Though it was shown both by DIII-D and JET that the ELM wetted area increases with the ELM loss energy, a scaling providing a quantitative estimate remained elusive. Recent attempts revisiting data from JET operation with carbon PFCs and with ITER-like-wall show an approximately linear dependence of ELM energy density with the pedestal top electron pressure. An attempt to scale the ELM energy density to ITER with pedestal pressure has the advantage that it directly utilizes measurements from Thomson-Scattering and infra-red thermography. However, it requires well diagnosed target heat load data for a wide scan in pedestal pressure. Currently this attempt is applied to data from ASDEX Upgrade to establish a multi-machine scaling. [Preview Abstract] |
Thursday, October 30, 2014 2:24PM - 2:36PM |
UO3.00003: Power Deposition on the DIII-D Inner Wall Limiter P.C. Stangeby, C.K. Tsui, J.D. Elder, C.J. Lasnier, A.G. McLean, A.W. Leonard, J.A. Boedo, D.L. Rudakov, M. Kocan, R.A. Pitts Power deposition on the inner wall limiter (IWL) of DIII-D was measured by infrared (IR) thermography and calculated from plasma profiles measured by an inner column Swing-Probe for 6 ohmic discharges. In some cases clear evidence was found for a narrow feature with $\lambda_{short}\,\sim$ ion poloidal gyro-radius $\sim$ a few mm, and of strength $q_{||0\underline{~}short}/q_{||0\underline{~}long} \sim 0.5$ $\pm$ a factor of 5, where $q_{||}$ is the parallel power flux density. The objective of the experiment was to check the assumptions made in defining the shape of the ITER IWL, in particular to see if the radial gradient of $q_{||}$ increases near the last closed flux surface on \mbox{DIII-D} in agreement with observations in other tokamaks [JET, COMPASS, TCV, C-Mod]. On the basis of the results from the IWL experiments done on the 5 tokamaks, ITER decided in April 2014 to re-design the limiter shape to accommodate a narrow power feature. [Preview Abstract] |
Thursday, October 30, 2014 2:36PM - 2:48PM |
UO3.00004: X-Divertors on ITER - with no hardware changes Prashant Valanju, Brent Covele, Mike Kotschenreuther, Swadesh Mahajan, Charles Kessel Using CORSICA, we have discovered that X-Divertor (XD) equilibria are possible on ITER -- without any extra PF coils inside the TF coils, and with no changes to ITER's poloidal field (PF) coil set, divertor cassette, strike points, or first wall. Starting from the Standard Divertor (SD), a sequence of XD configurations (with increasing flux expansions at the divertor plate) can be made by reprogramming ITER PF coil currents while keeping them all under their design limits (Lackner and Zohm have shown this to be impossible for Snowflakes). The strike point is held fixed, so no changes in the divertor or pumping hardware will be needed. The main plasma shape is kept very close to the SD case, so no hardware changes to the main chamber will be needed. Time-dependent ITER-XD operational scenarios are being checked using TSC. This opens the possibility that many XDs could be tested and used to assist in high-power operation on ITER. Because of the toroidally segmented ITER divertor plates, strongly detached operation may be critical for making use of the largest XD flux expansion possible. The flux flaring in XDs is expected to increase the stability of detachment, so that H-mode confinement is not affected. Detachment stability is being examined with SOLPS. [Preview Abstract] |
Thursday, October 30, 2014 2:48PM - 3:00PM |
UO3.00005: High resolution scrape-off layer profile measurements in limited and diverted plasmas in C-Mod -- investigation of heat flux channel width physics B. LaBombard, J.L. Terry, D. Brunner, E. Edlund, T. Golfinopoulos, J.W. Hughes, C. Theiler, J. Walk, S. Wolfe, D. Whyte Narrow scrape-off layer (SOL) heat flux channel widths ($\lambda _{\mathrm{q}})$ are seen on many tokamaks, both in inner-wall limited (IWL) and diverted discharges. These observations have important consequences for ITER and reactors, impacting the design of inner limiter tiles for heat load during startup and pushing the limits of dissipative divertor operation and control. A dominant $\sim$ 1/I$_{\mathrm{p}}$ scaling for $\lambda_{\mathrm{q}}$ is seen in a wide range of cases (IWL, H-mode and L-mode diverted [at low density]), suggestive of a poloidal ion gyroradius effect. It is troubling that $\lambda_{\mathrm{q}}$ does not appear to scale with major radius -- a challenge for reactors. The latter observation contrasts with H-mode pedestal widths that increase with machine size, implying that the physics that sets the local gradient scale lengths in pedestal and SOL may be different. We have recently implemented a scanning ``Mirror Langmuir Probe'' diagnostic on C-Mod with the idea of exploring this critical interface with very high resolution. Narrow $\lambda_{\mathrm{q}}$ ``features'' in IWL discharges have been mapped out in detail, exhibiting a $\sim$ 1/I$_{\mathrm{p}}$ scaling with some evidence of a break-in-slope feature at the LCFS. We will report on these findings and on L and H-mode experiments in progress, in which divertor conditions are varied (low recycling, high-recycling, detached). [Preview Abstract] |
Thursday, October 30, 2014 3:00PM - 3:12PM |
UO3.00006: Perspectives on the Final Design Review process from the US ITER DRGA team T.M. Biewer, C.C. Klepper, W. DeVan, V. Graves, C. Marcus, P. Andrew, D.W. Johnson Among the ITER procurements awarded to the US ITER Domestic Agency, and subsequently to the ORNL Fusion \& Materials for Nuclear Systems Division, is the design and fabrication of the Diagnostic Residual Gas Analyzer (DRGA) system. The DRGA system reached the Final Design Review (FDR) in July 2014, and is the first US-credited diagnostic system to achieve this milestone. The design effort has focused on the vacuum and mechanical interface of the DRGA gas sampling tube with the ITER vacuum vessel and cyrostat. In addition to technical issues needed to negotiate the mechanical interface, a significant number of procedural issues at US ITER and the ITER IO were encountered to navigate the DRGA project to this milestone. The process has been beneficial to both the DRGA project, and in-turn to US ITER, by illuminating the procedures in practice. This presentation will highlight some of the issues encountered and relay perspectives for designing hardware for ITER. [Preview Abstract] |
Thursday, October 30, 2014 3:12PM - 3:24PM |
UO3.00007: Control of Plasma Stored Energy for Burn Control Using DIII-D In-Vessel Coils R.J. Hawryluk, B.A. Grierson, E. Kolemen, R. Nazikian, W.M. Solomon, N.W. Eidietis, A.W. Hyatt, C. Paz-Soldan, S. Wolfe A new approach has been experimentally demonstrated to control the stored energy by applying a non-axisymmetric magnetic field using the DIII-D in-vessel coils to modify the energy confinement time. In relatively low collisionality DIII-D discharges, the application of non-axisymmetric magnetic fields results in a decrease in confinement time and density pumpout. Control of the stored energy was demonstrated by the application of non-axisymmetric fields while using gas puffing to compensate the density pumpout in the pedestal. Since the fusion power in a power plant operating at high $Q_{DT}$ can be related to the plasma stored energy and hence, is a strong function of the energy confinement time, the application of non-axisymmetric fields may be an effective technique to adjust the fusion power in a power plant. [Preview Abstract] |
Thursday, October 30, 2014 3:24PM - 3:36PM |
UO3.00008: Extending the Physics Basis of ITER Baseline Scenario Stability to Zero Input Torque C. Paz-Soldan, T.C. Luce, A.M. Garofalo, G.L. Jackson, R.J. La Haye, J.M. Hanson, K.E.J. Olofsson, F. Turco, B.A. Grierson, W.M. Solomon DIII-D operation at ITER baseline scenario parameters (safety factor$\sim$3, normalized $\beta\sim 2$, low input torque) is challenging due to the destabilization of m/n=2/1 or 3/2 tearing modes that rapidly lead to a loss of H-mode confinement and potential disruption. Despite proximity to stability limits, stationary operation at ITER-equivalent levels of input torque has been achieved with improved correction of DIII-D intrinsic error fields used to remove magnetic braking torques in combination with steady gas flows and pulsed 3D fields used to pace edge-localized mode (ELM) activity. Operation with zero input torque remains elusive. In this regime, ELMs are more difficult to control and toroidal rotation more difficult to sustain. Additionally, the confinement H-factor is found to decrease significantly from large to zero torque, regardless of heating mix. These results indicate that ITER baseline scenario extrapolations derived from discharges with large input torque are likely to be optimistic. [Preview Abstract] |
Thursday, October 30, 2014 3:36PM - 3:48PM |
UO3.00009: Steady-state ELM-suppressed H-modes from KSTAR to ITER and beyond Yongkyoon In, J.G. Kwak Long-pulse, steady-state high-performance plasma is not only an important mission in KSTAR, but also directly relevant to ITER. While demonstrating the pulse-length of more than 20 sec H-mode flat-top in 2013, KSTAR has been exploring various means to achieve and sustain steady-state, ELM-suppressed/mitigated H-modes using versatile in-vessel control coils (IVCC), ECCD/ECH, and/or SMBI. In particular, taking advantage of the versatile 3-rows of IVCC, KSTAR accomplished both n$=$1 and n$=$2 RMP-driven, ELM-suppressed regimes that lasted up to 4 sec so far (limited by the discharge pulse length, not by any physics constraints, and will be extended up to 10 sec in 2014.) We also found the use of n$=$2 RMP has prevented a locked-mode from being disruptive (at least within the RMP phase). To cope with run-away electrons and/or off-normal events, a soft landing algorithm has been developed and confirmed capable of ramping down the plasma current safely. The enhanced understanding and demonstration of steady-state, high-performance plasmas in KSTAR will elevate the level of confidence about the success of ITER and beyond. [Preview Abstract] |
Thursday, October 30, 2014 3:48PM - 4:00PM |
UO3.00010: Towards Current Profile Control in ITER: Potential Approaches and Research Needs E. Schuster, J.E. Barton, W.P. Wehner Many challenging plasma control problems still need to be addressed in order for the ITER Plasma Control System (PCS) to be able to successfully achieve the ITER project goals. For instance, setting up a suitable toroidal current density profile is key for one possible advanced scenario characterized by noninductive sustainment of the plasma current and steady-state operation. The nonlinearity and high dimensionality exhibited by the plasma demand a model-based current-profile control synthesis procedure that can accommodate this complexity through embedding the known physics within the design. The development of a model capturing the dynamics of the plasma relevant for control design enables not only the design of feedback controllers for regulation or tracking but also the design of optimal feedforward controllers for a systematic model-based approach to scenario planning, the design of state estimators for a reliable real-time reconstruction of the plasma internal profiles based on limited and noisy diagnostics, and the development of a fast predictive simulation code for closed-loop performance evaluation before implementation. Progress towards control-oriented modeling of the current profile evolution and associated control design has been reported following both data-driven and first-principles-driven approaches. An overview of these two approaches will be provided, as well as a discussion on research needs associated with each one of the model applications described above. [Preview Abstract] |
Thursday, October 30, 2014 4:00PM - 4:12PM |
UO3.00011: Control Solutions for High Performance in ITER with Test Blanket Modules M.J. Lanctot, J.S. deGrassie, R.J. La Haye, C. Paz-Soldan, E.J. Strait, R.J. Buttery, J.A. Snipes, H. Reimerdes, N.C. Logan, J.-K. Park, W.M. Solomon, B. Grierson, J.M. Hanson DIII-D experiments indicate applied n=1 fields can be used in high performance plasma regimes to reduce to a tolerable level the impact of the Test Blanket Modules (TBMs) error field (EF) on energy and particle confinement. Active coils, designed to mock-up the magnetic EF from two TBMs in one ITER equatorial port, were used to mimic the magnetization from the reduced-activation ferritic martensitic steel used in present TBM designs. The optimal correction fields, identified by maximizing the plasma toroidal angular momentum, reduced the impact of the TBM EF on energy, particle, and momentum confinement at $\beta_N=2.9$ by 60\%, a factor of 2 improvement over previous results at $\beta_N=1.8$. This improved performance of n=1 control fields at high beta is consistent with the hypothesis that the strong beta dependence of TBM EF effects observed in previous campaigns is due mainly to amplification of the n=1 component of the TBM EF. Similar performance was obtained with either internal or external n=1 error field control coils. The results suggest that the impact of the TBM related EFs on high beta operation can be controlled with the external correction coils in ITER. [Preview Abstract] |
Thursday, October 30, 2014 4:12PM - 4:24PM |
UO3.00012: ICRF-induced core impurities: Source and transport studies of ICRF conventional and field aligned ICRF antennas S.J. Wukitch, B. Labombard, Y. Lin, M.L. Reinke, J. Terry Ion cyclotron range of frequency power (ICRF) is considered a good candidate to provide bulk heating for ITER and future reactors. A primary challenge to ICRF utilization is to minimize plasma-material interaction using techniques that scale to continuous operation in a thermonuclear environment. New Alcator C-Mod experiments investigate impurity contamination associated with ICRF operation determining whether it is predominantly a result of increased source, transport or some combination. Previous work showed a field aligned (FA) antenna could significantly reduce core high-Z impurity contamination and lower limiter impurity sources compared to a toroidally aligned antenna. However, measurements of the RF-enhanced plasma potentials showed little difference between antennas designs. To investigate impurity penetration/screening directly, trace nitrogen is injected at different poloidal/toroidal locations, measuring core nitrogen levels in the presence and absence of ICRF power. This provides insight into transport changes associated with the RF and antennas concepts. [Preview Abstract] |
Thursday, October 30, 2014 4:24PM - 4:36PM |
UO3.00013: Modeling the ITER ICRF Antenna with Integrated Time Domain RF Sheath and Plasma Physics David Smithe, Daniel D'Ippolito, James Myra We present results from computer simulations of detailed 3D models of the ICRF launcher assembly, including straps, Faraday Shields, and vessel wall [1]. These simulations provide exquisite detail of the antenna near fields, and the sheaths between plasma and the metallic components of the launcher. Significant work has been done to create a sheath model [2] that allows us to estimate local values of sheath potential everywhere on the 3D structure, so that we can estimate RF rectified plasma potential [3]. Those potentials are in turn a likely source of sputtering and impurity creation, when the antennas are operating, and we discuss ongoing work to quantify these effects. Additional study of the antenna near fields also investigates slow waves which can exist in the low density scrape-off layer, and may impact power balance, and also sheath amplitudes. Movies of the 3D field and sheath oscillations will be shown. [1] ``Quantitative Modeling of ICRF Antennas with integrated Time Domain RF Sheath and Plasma Physics,'' David N. Smithe, et. al., Proceedings of the 20th Topical Conf. on RF Power in Plasmas, AIP Publishing (2013). [2] ``A radio-frequency sheath boundary condition and its effect on slow wave propagation,'' D. A. D'Ippolito and J. R. Myra, Phys. Plasmas vol. 13, p. 102508, 2006. [3] ``RF Models for Plasma-Surface Interactions: Sheath Boundary Conditions with Dielectrics,'' T. G. Jenkins and D. N. Smithe, Proceedings of the 2014 ICOPS/BEAMS Conf. [Preview Abstract] |
Thursday, October 30, 2014 4:36PM - 4:48PM |
UO3.00014: Measurements of relativistic emission from runaway electrons in Alcator C-Mod: spectrum, polarization, and spatial structure Robert Granetz, Robert Mumgaard At low densities, runaway electrons (RE's) can be generated during the flattop of Alcator C-Mod discharges with highly relativistic energies, $\gamma\gg1$, allowing careful study under steady conditions. These RE's emit light in a narrow forward-peaked cone which is detected with a number of diagnostics, including spectrometers, a video imaging camera, and polarimetry (using the MSE system), in addition to the standard hard x-ray detectors. These measurements of the relativistic emission can provide information about the RE energy distribution, pitch angle distribution, and spatial distribution. Unlike most other tokamaks, C-Mod's high magnetic field shifts the peak of the continuum emission into the visible, due to the smaller gyroradius and higher gyro-frequency, allowing for excellent spectral coverage with standard spectrometers, and thus detailed comparison to theoretical predictions of synchrotron and bremsstrahlung spectra. Additionally, camera images occasionally show highly structured formations. Profiles of the polarization fraction and polarization angle show radial structure, including a jump of $90^\circ$ outboard of the magnetic axis, in qualitative agreement with recent theoretical calculations for relativistic electrons in a tokamak field. [Preview Abstract] |
Thursday, October 30, 2014 4:48PM - 5:00PM |
UO3.00015: Influence of DIII-D Experiments on the Design of the ITER Shattered Pellet Injection System N. Commaux, D. Shiraki, L.R. Baylor, S.J. Meitner, S.K. Combs, N.W. Eidietis, E.M. Hollmann, V.A. Izzo, R.A. Moyer Shattered pellet injection (SPI) is a prime candidate for ITER disruption mitigation because of its deeper penetration and larger particle flux compared to massive gas injection (MGI). The ITER disruption mitigation system will likely use high Z impurities such as neon. An upgrade of the SPI on DIII-D (the only operating SPI in the world) enables for the first time ITER relevant injection characteristics: 400 torr.L neon pellets. The design of the SPI system is described as well as its evolution due to the results from DIII-D experiments and ITER design requirements. Recent experiments focused on differences in particle assimilation, thermal and current quench characteristics compared to MGI. Radiation asymmetries are regarded as a potential issue on ITER. Studies using MGI have showed that these effects can be significant on present devices. They are compared to new SPI results since they could influence the implementation of the SPI on ITER. [Preview Abstract] |
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