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 GI2: Plasma Wall and Impurity Physics |
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Chair: John Canik, Oak Ridge National Laboratory Room: Ballroom DE |
Tuesday, October 30, 2012 9:30AM - 10:00AM |
GI2.00001: Reduction of Net Erosion of High-Z Divertor Surface by Local Redeposition in DIII-D Invited Speaker: P.C. Stangeby Utilizing the unique capability to expose material samples to well characterized diverted plasmas, recent DIII-D measurements have confirmed theoretical expectations of the relative net and gross erosion rates of molybdenum in the divertor region. Knowledge of these erosion rates is important for predicting first wall lifetime in future fusion devices. Theory suggests that the net erosion rate will be much less than gross erosion due to prompt local deposition of eroded ions by gyro-orbit motion, the strong E-field toward the target and friction with the fast plasma flow toward the target. However, experimental evidence to date has been contradictory. The results here, which are the most definitive to date, are consistent with the basic theoretical predictions. The net and gross erosion rates were measured utilizing 1-cm and 1-mm diameter Mo samples that are mounted on the DIII-D Divertor Material Evaluation System (DiMES) system and simultaneously exposed near the attached outer strike point of an L-mode plasma for 4 s. Due to the spatial extent of the re-deposition, the larger sample gives the net erosion while the smaller sample is indicative of the gross erosion. Post-mortem ion beam analysis (RBS) of the larger sample, indicates a 2.9 nm film thickness reduction (or 0.72 nm/s net erosion rate). Similar analysis of the smaller sample yields a 1.3 nm/s gross erosion rate, consistent with spectroscopic measurements of Mo I emission. The net to gross erosion ratio of 0.56 is consistent with calculations using a modeling package including REDEP/WBS and OEDGE codes. Using as input the measured plasma density and temperature profiles from divertor Langmuir probes, these codes estimate a net to gross erosion ratio of 0.46. Details of the modeling and implications for future devices will be discussed. [Preview Abstract] |
Tuesday, October 30, 2012 10:00AM - 10:30AM |
GI2.00002: First Operation with the JET ITER-Like Wall Invited Speaker: Rudolf Neu To consolidate ITER design choices and prepare for its operation, JET has implemented ITER's plasma facing materials, namely Be at the main wall and W in the divertor. In addition, protection systems, diagnostics and the vertical stability control were upgraded and the heating capability of the neutral beams was increased to over 30~MW. First results confirm the expected benefits and the limitations of all metal plasma facing components (PFCs), but also yield understanding of operational issues directly relating to ITER. H-retention is lower by at least a factor of 10 in all operational scenarios compared to that with C PFCs. The lower C content ($\sim $ factor 10) have led to much lower radiation during the plasma burn-through phase eliminating breakdown failures. Similarly, the intrinsic radiation observed during disruptions is very low, leading to high power loads and to a slow current quench. Massive gas injection using a D$_{2}$/Ar mixture restores levels of radiation and vessel forces similar to those of mitigated disruptions with the C wall. Dedicated L-H transition experiments indicate a reduced power threshold by 30{\%}, a distinct minimum density and pronounced shape dependence. The L-mode density limit was found up to 30{\%} higher than for C allowing stable detached divertor operation over a larger density range. Stable H-modes as well as the hybrid scenario could be only re-established when using gas puff levels of a few 10$^{21}$e/s. On average the confinement is lower with the new PFCs, but nevertheless, H factors around 1 (H-Mode) and 1.2 (at $\beta _{N}\sim $3, Hybrids) have been achieved with W concentrations well below the maximum acceptable level ($<$10$^{-5})$. [Preview Abstract] |
Tuesday, October 30, 2012 10:30AM - 11:00AM |
GI2.00003: Comparison of Tungsten Fuzz Growth in Alcator C-Mod and Linear Plasma Devices Invited Speaker: G.M. Wright The growth of tungsten (W) nano-tendrils or ``fuzz'' is a well-known process in linear plasma devices (LPD) requiring a clean tungsten surface, elevated surface temperatures (T$_{surf}$ = 900-2000 K), and a flux of low-energy helium (E$_{He }>$ 20 eV). In a dedicated experiment on Alcator C-Mod, W fuzz was grown, for the first time, in a tokamak environment. The W fuzz was grown on a Langmuir probe in the lower divertor. During L-mode helium plasma discharges this W Langmuir probe received q$_{surf} \sim $ 30 MW/m$^{2}$ and its surface temperature increased from 400 K to 2300 K. Over 14 sequential discharges, the W probe had an integrated exposure time of $\sim $15 s at temperatures between 900-2300 K. Focused ion beam cross-sectioning showed a 600 $\pm $ 150 nm thick tungsten fuzz layer was grown on the probe surface. The W fuzz showed no signs of uni-polar arcing and there was no nano-tendril melting or damage despite the high surface heat fluxes. Three full current (900 kA) unmitigated mid-plane plasma disruptions occurred during the discharge sequence including one on the final discharge of the experiment, but the W fuzz was undamaged. The W probe fuzz layer thickness is, within uncertainties, in agreement with an empirical fuzz growth rate formula from the PISCES LPD. However the validity of the PISCES formula in the high temperature regime experienced by the W Langmuir probe is uncertain. For a more relavent comparison, an experiment on the Pilot-PSI high-power LPD has exposed a tungsten target to surface temperatures, He flux densities, and exposure times similar to what was experienced by the W Langmuir probe in Alcator C-Mod. The W target in Pilot-PSI grew W fuzz with a morphology and layer thickness nearly identical to the Alcator C-Mod W fuzz demonstrating the growth process and mechanism is nearly identical in tokamaks and LPD. This helps validate LPD research on W fuzz for predictions or calculations of growth in tokamaks. [Preview Abstract] |
Tuesday, October 30, 2012 11:00AM - 11:30AM |
GI2.00004: Particle Control and Confinement in the Lithium Tokamak eXperiment (LTX) with Lithium-Coated Walls Invited Speaker: Dick Majeski The Lithium Tokamak eXperiment (LTX) is a low aspect ratio tokamak with R=0.4 m, a=0.26 m, and kappa=1.5. The toroidal field is 2.1 kG, plasma current less than 100 kA, and discharge duration less than 50 msec. LTX is fitted with a 1 cm thick heated liner. The plasma-facing surface of the liner is clad with stainless steel, conformal to the last closed flux surface. The liner can be heated to 300 - 400 C, and coated with lithium. With a high-Z steel wall, discharges are strongly affected by wall conditioning. In LTX, the only wall conditioning technique used is lithium wall coating. Discharges without lithium coatings are limited to plasma currents of 15 kA, and discharge durations to 5 msec. With lithium coatings discharge currents and discharge durations increase 5-fold. Peak electron temperatures, from preliminary Thomson scattering measurements, range from 100 - 200 eV. Electron temperature profiles for lithium-wall discharges will be presented. DEGAS2-based estimates of local recycling using an extensive set of Lyman-alpha detectors will be discussed. Particle pumping has been compared for solid, room temperature and hot (300C), liquefied lithium coatings. The fueling efficiency of a number of different gas injection techniques, including supersonic gas injection and molecular cluster injection has been characterized. These techniques can produce fueling efficiencies of up to 35\%. A set of liquid lithium injectors to fill the two lower shell segments with up to 50 g of liquid lithium has been installed, as well as an electron beam stirring system to ensure that the plasma contacts a clean lithium surface. In a collaboration with ORNL, the Doppler shifted emission of Li ions has been used to estimate the ion temperature and rotation profiles of LTX discharges. Analysis is in progress and preliminary results will be reported. [Preview Abstract] |
Tuesday, October 30, 2012 11:30AM - 12:00PM |
GI2.00005: Modifications of impurity transport and divertor sources with lithium wall conditioning in NSTX Invited Speaker: Filippo Scotti In the National Spherical Torus Experiment (NSTX), a large spherical tokamak with graphite plasma facing components (PFCs), lithium evaporative coatings are routinely used for wall conditioning. In lithium-conditioned H-mode discharges, core carbon accumulation is generally observed with concentrations $\leq 10\%$, measured by charge exchange recombination spectroscopy. Lithium ions do not accumulate significantly and have core concentrations $\leq 0.1\%$. In this work, modifications in carbon and lithium sources and transport due to lithium coatings are analyzed. The change in ELM behavior (from ELMy to ELM-free) together with modifications in carbon neoclassical transport, due to changes in main ion temperature and density profiles, analyzed with NCLASS, NEO and MIST transport codes, can lead to an increased carbon confinement. The high lithium diffusivity due to the presence of a strong impurity (carbon) prevents core accumulation. Spectroscopic impurity influxes (Li I-II, C II-III measured by filtered cameras and divertor spectrometers via the S/XB method) and 2D multi-fluid code analysis (UEDGE) are employed to understand impurity sources and SOL parallel and perpendicular transport. Only a moderate reduction in the measured carbon sputtering yield and divertor carbon influxes is observed with lithium coatings on graphite PFCs. The expected reduction due to the coverage of graphite surfaces with lithium, in fact, can be counteracted by the degradation of the coatings and the change in divertor plasma parameters, due to the transition of the SOL to a sheath-limited regime. Prompt re-deposition of sputtered lithium atoms, meanwhile, strongly reduces the net lithium erosion and the divertor lithium source as evident from measured ionized lithium influxes which are $< 10\%$ of neutral lithium influxes. [Preview Abstract] |
Tuesday, October 30, 2012 12:00PM - 12:30PM |
GI2.00006: Poloidal variation of high-Z impurity density in Alcator C-Mod ICRF-heated plasmas Invited Speaker: Matthew Reinke The poloidal variation of molybdenum density is measured in the core of ICRF-heated Alcator C-Mod plasmas and found to exhibit strong in/out asymmetries. Existing neoclassical parallel impurity transport theory is extended to include the effects of fast-ions and is shown to agree quantitatively with C-Mod measurements. The flux-surface variation of molybdenum is well described by n$_{z}(\theta )$/$\langle $n$_{z}\rangle $=1+n$_{z,c}$ cos($\theta )$+n$_{z,s}$sin($\theta )$, where -0.2 $<$ n$_{z,c}$/$\langle $n$_{z}\rangle \quad <$ 0.3 and -0.1 $<$ n$_{z,s}$/$\langle $n$_{z}\rangle \quad <$ 0.1 are observed over a wide range of Ohmic, L/I-mode and EDA H-mode plasmas for r/a $<$ 0.9. The in/out asymmetry, n$_{z,c}$/$\langle $n$_{z}\rangle $, is determined by a combination of centrifugal force due to toroidal rotation, leading to low-field side (LFS) accumulation, and poloidal electric fields sustained by magnetic trapping of cyclotron heated minority ions, leading to high field side (HFS) accumulation. While LFS accumulation due to centrifugal effects has been seen on other tokamaks, this represents the first observation of the effect driven entirely by intrinsic rotation. Scans of the D(H) resonance layer are shown to modify the in/out asymmetry by altering the fast-ion temperature anisotropy, T$_{-}$/T$_{\vert \vert }$, and changing the ICRF power density, P$_{RF}$/n$_{e}$, either by ramping down the input power or increasing the density is found to reduce HFS accumulation. Observations of up/down asymmetries n$_{z,s}$/$\langle $n$_{z}\rangle $, of molybdenum density are found to disagree with existing theories in the trace limit, n$_{z}$Z$^{2}$/n$_{i} \quad \ll $ 1, in the collisionless main-ion regime. The link between n$_{z}(\theta )$ and poloidal rotation, v$_{\theta }$, is emphasized, as both are assumed to be determined by neoclassical parallel impurity transport, and a more rigorous test of theory which includes matching asymmetries and v$_{\theta }$ is discussed. The use of the poloidal variation in n$_{z}$ as a diagnostic for E$_{\theta }$ and T$_{-}$/T$_{\vert \vert }$ as well as the impact of n$_{z,c}$/$\langle $n$_{z}\rangle $ on radial transport are also discussed. [Preview Abstract] |
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