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 JI1: ITER Physics |
Hide Abstracts |
Chair: Chuck Greenfield, General Atomics Room: Acadia |
Tuesday, October 28, 2014 2:00PM - 2:30PM |
JI1.00001: Narrow limiter SOL power channels and their impact on ITER first wall shaping Invited Speaker: Martin Kocan Until recently, it was generally accepted that the profile of parallel heat flux density in the SOL of limited tokamak plasmas can be well approximated by a single exponential with decay length $\lambda_{\mathrm{q}}$. This popular belief was emphatically shown to be erroneous in 2012, when IR measurements on the inner column of JET limiter discharges revealed the presence of a narrow heat flux channel adjacent to the last closed flux surface, resembling a feature seen elsewhere two decades ago, but never seriously pursued by the edge physics community. This near-SOL decay occurs with $\lambda_{\mathrm{q}}$ few mm, much shorter than the main SOL $\lambda_{\mathrm{q}}$, and can raise the heat flux at the limiter apex a factor 1-4 above the value expected from a single, broad exponential. The JET observations were of great practical consequence, demonstrating that the logarithmically-shaped ITER inner wall (IW), foreseen as a limiter surface for plasma start-up, would be unsuited to handling the power loads produced by such a narrow feature. Alerted by this JET data, the ITER Organization (IO) initiated a multi-machine effort to examine this new physics, with the C-Mod, DIII-D, COMPASS and TCV tokamaks all finding the narrow heat flux channel in dedicated experiments. This talk will describe how these new data are helping to unravel the physics of the narrow feature and how they have provided a strong enough basis for the IO to modify the IW toroidal shape profile. The new IW shape is optimized for a double-exponential profile with $\lambda_{\mathrm{q}} = $ 4 and 50 mm, both derived from multi-machine databases for the near and main-SOL features. It has the interesting property of mitigating the impact of the narrow feature, whilst paying no penalty if the latter is not eventually found in ITER. If it were, and without the modification, IWL limiter operation up to several MA, as required by the ITER Heat Load Specifications, would not be possible. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization. [Preview Abstract] |
Tuesday, October 28, 2014 2:30PM - 3:00PM |
JI1.00002: Impurity transport experiments in Alcator C-Mod to address high priority R and D for ITER* Invited Speaker: Matthew Reinke The decision to start ITER operation in the non-active phase (H/He plasmas) with a W divertor has brought increased attention to physics issues related to high Z impurity transport and control. Lack of impurity control would lead to the radiative collapse of plasmas by W accumulation and increased disruptivity, which is detrimental to ITER operation. Prediction of plasma contamination by high Z impurities in ITER and its control requires understanding of W production and transport in the SOL through the edge transport barrier (in H-mode) and into the core plasma in conditions as similar as possible to those in ITER. Experiments in Alcator C-Mod have been carried out to understand these physics processes and their control in electron heated (ICRH$+$LHCD) L-mode and H-mode plasmas without significant core particle or momentum sources as in ITER. Core transport of impurities injected by laser blow off in L-modes shows that, in the absence of sawteeth, W develops stationary peaked profiles (for 100s of ms vs. $\tau_{\mathrm{E}} =$ 14 ms) while lower Z impurities (e.g. Ar) do not. Sawteeth are observed to be very effective in flushing out W on timescales of the sawtooth period ($=$10 ms) in these L-mode plasmas. Injection of Ca shows that the impurity diffusivity in the core plasma decreases by a factor of 10 in H-mode with respect to L-mode and that an inwards pinch is necessary at the H-mode edge to match observed increases in impurity confinement time in H-mode, up a factor of 5-50 compared to L-mode. Initial results of impurity transport in He L-mode plasmas show similar results to D plasmas. The results of these and other ongoing C-Mod experiments (including impurity transport and impurity peaking control by heating sources (magnitude, fast particle population, radial location, etc.) in H-mode plasmas with ITER-like peaked density profiles and T$_{\mathrm{e}} =$ T$_{\mathrm{i}}$ ) will be described and consequences for ITER impurity transport control will be drawn. Supported by the US DOE under DoE Contract No. DE-FC02-99ER54512 [Preview Abstract] |
Tuesday, October 28, 2014 3:00PM - 3:30PM |
JI1.00003: The Quiescent H-mode Regime for High Performance ELM-Stable Operation in Future Burning Plasmas Invited Speaker: A.M. Garofalo Recent experiments on DIII-D have increased confidence in the ability to achieve high confinement, ELM-stable operation on ITER through implementation of the quiescent H-mode (QH-mode) regime. By tailoring the plasma shape to improve the edge stability, the QH-mode operating space has been extended to densities exceeding 70\% of the Greenwald limit, overcoming the long-standing low-density limit of QH-mode operation. In addition, the simultaneous achievement of QH-mode at ITER relevant values for beta, confinement, and safety factor sustained for many energy confinement times in an ITER similar shape has been demonstrated for the first time. QH-mode provides excellent energy confinement, even at near zero plasma rotation, while operating without ELMs and with strong impurity transport via the benign edge harmonic oscillation (EHO). Peeling-ballooning theory of the plasma edge explains the EHO as a saturated kink-peeling mode, and predicts that ITER will operate in the edge regime where QH-mode can exist. In the theory, the density range over which the plasma encounters the kink-peeling boundary widens as the plasma cross-section shaping is increased, thus increasing the QH-mode density threshold. The DIII-D results are in excellent agreement with these predictions, and non-linear MHD analysis of reconstructed QH-mode equilibria shows unstable low n kink-peeling modes growing to a saturated level, consistent with the theoretical picture of the EHO. Furthermore, high density operation in the QH-mode regime has opened a path to a new, previously predicted region of parameter space dubbed ``Super H-mode,'' characterized by very high pedestals that can be more than a factor of two above the peeling-ballooning stability limit for similar ELMing H-mode discharges at the same density. [Preview Abstract] |
Tuesday, October 28, 2014 3:30PM - 4:00PM |
JI1.00004: The High-$\beta_N$ Hybrid Scenario for ITER and FNSF Steady-State Mission Invited Speaker: Francesca Turco New experiments on DIII-D have demonstrated the steady-state potential of the hybrid scenario, with 1 MA of plasma current driven fully noninductively and $\beta_N$ up to 3.7 sustained for $\sim$3 s ($\sim$1 current diffusion time, $\tau_R$, in DIII-D), providing the basis for an attractive option for steady-state operation in ITER and FNSF. Excellent confinement is achieved ($H_{98y2}\sim 1.6$) without performance limiting tearing modes. The usual Advanced Tokamak (AT) approach relies on a large fraction of off-axis current drive and careful current drive alignment to reach $q_{min}>2$ and high bootstrap current ($>$70\%). In contrast, the hybrid regime overcomes the need for off-axis current drive efficiency, taking advantage of the poloidal magnetic flux pumping, believed to be the result of a saturated 3/2 tearing mode, to produce a self-organized current density profile. This allows for efficient current drive close to the axis, without deleterious sawtooth instabilities. In these new experiments, the edge surface loop voltage is driven down to zero for $>$1 $\tau_R$ when the poloidal $\beta$ is increased above 1.9 by utilizing 3.15 MW of electron cyclotron current drive at a plasma current of 1.0 MA and density of $\sim$4$\times$10$^{19}\,$m$^{-3}$. Stationary operation of hybrid plasmas with all on-axis current drive is sustained at pressures slightly above the ideal no-wall limit, while the calculated ideal with-wall MHD limit is $\beta_N\sim\,$4-4.5. For the first time, off-axis NBI power has been used to broaden the pressure and current profiles in this scenario, seeking to take advantage of higher predicted kink stability limits and lower values of tearing stability index $\Delta^\prime$, as calculated by the DCON and PEST3 codes. Preliminary results based on measured profiles predict ideal limits at $\beta_N >4.5$. With collisionality and edge safety factor values comparable to those envisioned for ITER and FNSF, the high-$\beta_N$ hybrid represents an attractive high performance option for the steady-state missions of these devices, [Preview Abstract] |
Tuesday, October 28, 2014 4:00PM - 4:30PM |
JI1.00005: Improved confinement in ELM-suppressed high-density H-modes at the ITER field via modification of the plasma boundary with Lower Hybrid RF Invited Speaker: J.L. Terry Injecting Lower Hybrid (LH) power into Alcator C-Mod's high-density H-mode plasmas has enhanced global confinement by increasing pedestal temperature gradients, modifying edge rotation, and decreasing edge and SOL turbulence. These new experiments indicate that edge LHRF can be used as a tool to increase confinement via \textit{direct modification} of boundary quantities. Ray-tracing modeling and accessibility calculations for the LH waves indicate that the LH waves do \textit{not penetrate to regions inside the top of the pedestal }and are not driving current in these plasmas; instead the LH power modifies the boundary conditions. When moderate amounts of LH power (P$_{\mathrm{LH}}$/P$_{\mathrm{tot}}=$20{\%}) are applied to high-density EDA H-modes (n$_{\mathrm{eo}}=$3.5x10$^{\mathrm{20}}$ m$^{\mathrm{-3}})$, we observe the following effects: edge/SOL fluctuation power \textit{decreases by roughly an order of magnitude}; pedestal temperature gradients are increased; global energy confinement time and H-factor increase by 30-40{\%} (H$_{\mathrm{98}}$ from 0.7 to 1.0); co-current core and pedestal rotation velocities increase; power to the (outer) divertor target increases promptly with an increment that is roughly 1/2 of the injected LH power, qualitatively consistent with the inaccessibility of the LH waves; and the central frequency of the edge-localized Quasi-Coherent Mode down-shifts and becomes much more coherent. These H-mode confinement improvements brought about by the edge LHRF are the result of changes in the pedestal (e.g. changes in rotation/shear and increased pedestal temperature gradients), with no substantial change in peaking of core density or temperature profiles. There is not perfect correlation with edge turbulence suppression, indicating that the turbulence decrease may be a necessary, but not sufficient, condition for the pedestal and confinement improvements. [Preview Abstract] |
Tuesday, October 28, 2014 4:30PM - 5:00PM |
JI1.00006: Impact of the plasma response in three-dimensional edge plasma transport modelling for RMP ELM control scenarios at ITER Invited Speaker: Oliver Schmitz The constrains used in magneto-hydrodynamic (MHD) modeling of the plasma response to external resonant magnetic perturbation (RMP) fields have a profound impact on the three-dimensional (3-D) shape of the plasma boundary induced by RMP fields. In this contribution, the consequences of the plasma response on the actual 3D boundary structure and transport during RMP application at ITER are investigated. The 3D fluid plasma and kinetic neutral transport code EMC3-Eirene is used for edge transport modeling. Plasma response modeling is conducted with the M3D-C1 code using a single fluid, non-linear and a two fluid, linear MHD constrain. These approaches are compared to results with an ideal MHD like plasma response. A 3D plasma boundary is formed for all cases consisting of magnetic finger structures at the X-point intersecting the divertor surface in a helical footprint pattern. The width of the helical footprint pattern is largely reduced compared to vacuum magnetic fields when using the ideal MHD like screening model. This yields increasing peak heat fluxes in contrast to a beneficial heat flux spreading seen with vacuum fields. The particle pump out as well as loss of thermal energy is reduced by a factor of two compared to vacuum fields. In contrast, the impact of the plasma response obtained from both MHD constrains in M3D-C1 is nearly negligible at the plasma boundary and only a small modification of the magnetic footprint topology is detected. Accordingly, heat and particle fluxes on the target plates as well as the edge transport characteristics are comparable to the vacuum solution. This span of modeling results with different plasma response models highlights the importance of thoroughly validating both, plasma response and 3D edge transport models for a robust extrapolation towards ITER. [Preview Abstract] |
Follow Us |
Engage
Become an APS Member |
My APS
Renew Membership |
Information for |
About APSThe American Physical Society (APS) is a non-profit membership organization working to advance the knowledge of physics. |
© 2024 American Physical Society
| All rights reserved | Terms of Use
| Contact Us
Headquarters
1 Physics Ellipse, College Park, MD 20740-3844
(301) 209-3200
Editorial Office
100 Motor Pkwy, Suite 110, Hauppauge, NY 11788
(631) 591-4000
Office of Public Affairs
529 14th St NW, Suite 1050, Washington, D.C. 20045-2001
(202) 662-8700