Session JI2: Edge and Divertor Physics

Edge transport and turbulence reduction, and formation of ultra-wide pedestals with lithium coated PFCs in NSTX

The coating of plasma facing components (PFCs) with lithium improves energy confinement [1] and eliminates ELMs in the National Spherical Torus Experiment (NSTX), the latter due to a relaxation of the density and pressure profiles that reduces the drive for peeling-ballooning modes [2]. Here we show that both a reduction in recycling (due to lithium pumping) and cross-field transport is needed to reproduce the measured profile changes. Furthermore we document a concomitant density fluctuation reduction measured in the steep gradient region. The experimental transport coefficients are obtained via data-constrained modeling using the SOLPS code [3], which couples a 2D fluid treatment of the edge plasma transport to a Monte Carlo neutrals calculation. First, a reduction in the PFC recycling coefficient from R$\sim $0.98 to R$\sim $0.90 is required to match the drop in D$\alpha$ emission with lithium coatings. Furthermore, a $\sim $75{\%} drop of the D and $\chi $e from 0.8 $< \psi $N $<$ 0.93 are needed to match the profile relaxation with lithium coatings; indeed, the region of low transport in the H-mode simply extends to the innermost domain of the simulation. Transport is similar with and without lithium coatings outside of $\psi$N $\sim$ 0.93, with D/$\chi$e $\sim$ 0.2/1.0 m2/s. Turbulence measurements using an edge reflectometry system [4] show a decrease in broadband density fluctuations with lithium coatings, primarily at frequencies $<$10 kHz. These transport changes allow the realization of very wide pedestals, with a $\sim $100{\%} width increase relative to the reference discharges. \\[4pt] [1] H. W. Kugel et al, Phys. Plas. 15 (2008) 056118. \\[0pt] [2] R. Maingi et al, Phys. Rev. Lett. 103 (2009) 075001. \\[0pt] [3] R. Schneider et al, Contr. Plas. Phys. 46 (2006) 3. \\[0pt] [4] S Kubota et al, Bull. Am. Phys. Soc. 53 (2008) 188.   

Taming the Plasma Material Interface with the ``Snowflake'' Divertor in NSTX

Recent results from NSTX provide support to the innovative ``snowflake'' divertor (SFD) configuration [1] as a promising plasma-material interface (PMI) concept for future magnetic fusion energy devices, through the demonstration of the SFD with significant divertor peak heat flux reduction and impurity control simultaneously with good H-mode confinement. In ITER and future tokamaks, the divertor PMI must be able to exhaust steady-state heat fluxes up to 10 MW/m$^{2}$ with minimal material erosion. In spherical tokamaks, these requirements are aggravated by the inherently compact divertor geometry. The SFD uses a second-order null-point created by bringing in close proximity two first-order X-points of the standard divertor configuration. The SFD configuration was obtained in NSTX with two divertor magnetic coils controlled in real time. Experiments in NSTX conducted in 0.8 MA 4-6 MW NBI-heated discharges qualitatively confirmed the SFD properties predicted by analytic theory and 2D multi-fluid edge transport modeling with the UEDGE code [1, 2]. When compared to the standard divertor geometry, the SFD in NSTX showed an increase in plasma-wetted area by 100-200 {\%} and an increased divertor volume (with X-point connection length increased by 50-100{\%}). Partial detachment of the outer strike point region (first $\sim $2-3 mm of the scrape-off layer width 6-7 mm mapped to midplane) was evident through a significant reduction of the peak divertor heat flux, a 100{\%} increase in divertor plasma radiation, and formation of a zone with $T_{e}$ =0.8-1.2 eV, $n_{e}$ =2-6x10$^{20}$m$^{-3}$, resulting in significant recombination and volumetric momentum losses. Core carbon inventory and radiated power were reduced by up to 70{\%}, apparently as a result of reduced divertor physical and chemical sputtering in the SFD. \\[4pt] [1] D. D. Ryutov, Phys. Plasmas \textbf{14}, 64502 (2007). \\[0pt] [2] M. V. Umansky \textit{et al}., Nucl. Fusion \textbf{49}, 075005 (2009).   

Quiet periods, zonal flows, and blob formation in the edge turbulence of NSTX

This talk will present recent observations of ``quiet periods'' in the edge turbulence of NSTX which are correlated with 3 kHz reversals in the direction of local poloidal flow. The turbulence measurements were made using the gas puff imaging (GPI) diagnostic, which can make 2D movies of the edge turbulence at $\leq$ 400,000 frames/sec. During quiet periods the strong edge turbulence in L-mode plasmas becomes transiently similar to the quiescent edge seen in H-mode plasmas. These quiet periods recur at about 3 kHz, which is near the frequency calculated for GAMs by the NLET code [1] and for zonal flows by the SOLT code [2]. The poloidal flows near the separatrix are correlated with blob formation and transport into the SOL, similarly to the models described in [2-4]. The local turbulence flow shear was also modulated by the quiet periods, but was not unusually large preceding the L-H transition. These results will be compared with previous experiments on edge zonal flows [5] and blob formation [6]. In general, they support the idea that edge zonal flows or GAMs could be regulating blob formation and SOL transport. \\[4pt] [1] K. Hallatschek and A. Zeiler, Phys. Plasmas 7, 2554 (2000)\\[0pt] [2] D.A. Russell, J.R. Myra, D.A. D'Ippolito, Phys. Plasmas 16, 122304 (2009)\\[0pt] [3] S.I. Krasheninnikov and A. I. Smolyakov, Phys. Plasmas 14,102503 (2007)\\[0pt] [4] P.H. Diamond, S.I. Itoh, K. Itoh, T.S. Hahm, Plas. Phys. Cont. Fus. 47, R35 (2005)\\[0pt] [5] A. Fujisawa, Nucl. Fusion 49 013001 (2009)\\[0pt] [6] I. Furno et al, Phys. Plasmas 15, 055903 (2008)   

Scaling of the power exhaust channel in Alcator C-Mod

Physics-based transport models that can accurately simulate the heat-flux power widths in tokamaks are lacking at the present time. Yet this parameter is fundamentally important for ITER and more critically important for DEMO. In order to improve our understanding, Alcator C-Mod has recently installed an extensive array of divertor heat flux diagnostics to explore power exhaust channels over its wide range of accessible conditions, including reactor-level parallel heat fluxes ($>$ 0.5 GW/m2) with high neutral opacity in an ITER-like vertical divertor plate geometry. This research is part of a coordinated science program with DIII-D and NSTX [1]. C-Mod's divertor heat flux ``footprints'' are found to exhibit a two zone structure: a narrow channel at the strike point location and a tail feature that extends into the far scrape-off layer. The balance in power carried by these two features can change depending on core radiation levels. In EDA H-mode discharges, the narrow feature is typically 2-3 mm wide mapped to the outer midplane; integral heat flux widths ($\lambda _{p})$ are 3.5 to 5 mm. These data stand in contrast to the empirical scalings used to estimate $\lambda _{p}$ for ITER [2], which predict 0.5 mm for C-Mod. $\lambda _{p}$ are found to systematically decrease with increasing plasma stored energy, which is in turn linked to the height of the edge pedestal and the strength of the quasi-coherent mode that accompanies EDA H-modes. These correlations yield an inverse relationship between $\lambda _{p}$ and plasma current for discharges that attain the same normalized pressure gradient in the pedestal. Thus pedestal physics appears to be strongly coupled to the width of the power exhaust channel on adjacent open field lines. \\[4pt] [1] DoE Joint Facilities Research Target for FY2010 http://www.science.doe.gov/ofes/performancetargets.shtml. \\[0pt] [2] Kirnev, et al., Plasma Phys. Control. Fusion 49 (2007) 689-701.   

High confinement/high radiated power H-mode experiments in Alcator C-Mod and consequences for ITER Q$_{DT}$=10 operation

Understanding the mechanisms that influence plasma confinement and drive its degradation for high density/radiation conditions in H-mode is of critical importance to ITER, since the expected edge power flow for Q{\_}$_{DT}$ =10 operation is very close to that required to maintain H-mode confinement and the level of radiated power required for acceptable divertor power fluxes is $\sim $ 80{\%}. Experiments have been carried out in Alcator C-Mod to study the role of edge power flux and plasma radiation in determining the quality of H-mode plasma confinement by varying injected ICRF power up to $\sim $ 5MW, and by injecting extrinsic impurities of various Z (Ar, Ne, N) in both EDA and ELMy H-modes. Analysis of the experimental results shows that plasma energy confinement (and edge pedestal pressure) are primarily determined by the absolute level of edge power flux P$_{edge}$ = P$_{in}$-P$_{rad,core}$, independently of the way this edge power flow is achieved and thus not directly correlated with core nor total radiated power fractions. In this respect, high Z impurity seeding (Ar) is found to cause lower plasma confinement for a given level of total plasma radiation because of the larger core plasma radiation and reduced edge power flow in these experiments. With lower Z impurity seeding, EDA H-modes with high plasma confinement (H$_{98}$ = 1) have been maintained at high density with edge power flows only marginally exceeding (20-30{\%}) the required power for H-mode access, as expected in ITER Q$_{DT}$ =10 plasmas, and with a considerable reduction of the divertor power flux (by a factor of 2 or more) relative to attached plasma conditions, also as required in ITER. Detailed analysis of the experiments in C-Mod and consequences for ITER plasma performance will be discussed. Work supported by US DoE Agreement DE-FC02-99ER54512.   

Modification of divertor heat and particle flux profiles with 3-D fields in NSTX

Externally imposed non-axisymmetric magnetic perturbations produce multiple local peaks and valleys in the divertor heat and particle flux profiles [1] in NBI-heated plasmas in the National Spherical Torus Experiment (NSTX) with $B_{t}$ = 0.4T, $I_{p}$ = 800kA, \textit{$\beta $}$_{t} \quad \sim $10{\%}. The addition of 3-D fields causes pronounced lobes to form near the separatrix X-point, which leads to the ``strike point splitting'' [2, 3] and flux striations observed in experiments. ITER may rely on 3--D resonant magnetic perturbation (RMP) fields for ELM suppression, and non-axisymmetric heat and particle deposition and an increase of peak values could pose additional engineering constraints. In NSTX, the radial location and spacing of the divertor striations produced by 3-D fields are reproduced well using vacuum field tracing of the superposition of vacuum 3-D fields and 2-D equilibrium fields [1]. The applied n=3 fields can also trigger ELMs [4]. The ELM heat flux profiles (measured with a new fast IR camera [5]) appear to be phase locked to the n=3 field structure, as also reported in DIII-D experiments [3]. The inclusion of the response of the plasma inside the separatrix (calculated with IPEC [6]) as the base equilibrium for field line tracing did not alter the computed structure of striations significantly compared to the vacuum modeling. This suggests that vacuum field line tracing alone may predict the effect of 3-D fields on divertor profiles even in rapidly rotating, high-\textit{$\beta$} plasmas. This work was supported in part by US DOE, DE-AC05-00OR22725 and DE-AC02-09CH11466.\\[4pt] [1] J-W. Ahn, \textit{et al}, Nucl. Fusion \textbf{50} (2010) 045010\\[0pt] [2] T.E. Evans, \textit{et al}, \textit{J. Phy.: Conf. Series} \textbf{7} (2005) 174\\[0pt] [3] M.W. Jakubowski, \textit{et al}, Nucl. Fusion \textbf{49} (2009) 095013, and references therein\\[0pt] [4] J.M. Canik, \textit{et al}, Phys. Rev. Letts. \textbf{104} (2010) 045001\\[0pt] [5] J-W. Ahn, \textit{et al}, Rev. Sci. Intrum. \textbf{81} (2010) 023501\\[0pt] [6] J.-K. Park, \textit{et al}, Phys. Plasmas \textbf{14 }(2007) 052110