Lithium wall conditioning in NSTX has resulted in reduced divertor recycling, improved energy confinement, and reduced frequency of edge-localized modes (ELMs), up to the point of complete ELM suppression. NSTX tiles were removed from the vessel following the 2008 campaign and subsequently analyzed using X-ray photoelectron spectroscopy as well as nuclear reaction ion beam analysis. In this paper we relate surface chemistry to deuterium retention/recycling, develop methods for cleaning of passivated NSTX tiles, and explore a method to effectively extract bound deuterium from lithiated graphite. Li–O–D and Li–C–D complexes characteristic of deuterium retention that form during NSTX operations are revealed by sputter cleaning and heating. Heating to ∼850 °C desorbed all deuterium complexes observed in the O 1s and C 1s photoelectron energy ranges. Tile locations within approximately ±2.5 cm of the lower vertical/horizontal divertor corner appear to have unused LiO bonds that are not saturated with deuterium, whereas locations immediately outboard of this region indicate high deuterium recycling. X-ray photo electron spectra of a specific NSTX tile with wide ranging lithium coverage indicate that a minimum lithium dose, 100–500 nm equivalent thickness, is required for effective deuterium retention. This threshold is suspected to be highly sensitive to surface morphology. The present analysis may explain why plasma discharges in NSTX continue to benefit from lithium coating thickness beyond the divertor deuterium ion implantation depth, which is nominally <10 nm.

Deuterium fueling experiments were conducted with the NSTX Liquid Lithium Divertor (LLD). Lithium evaporation recoated the LLD surface to approximate flowing liquid Li to sustain D retention. In the first experiment with the diverted outer strike point on the LLD, the difference between the applied D gas input and the plasma D content reached very high values without disrupting the plasma, as would normally occur in the absence of Li pumping, and there was also little change in plasma D content. In the second experiment, constant fueling was applied, as the LLD temperature was varied to change the surface from solid to liquid. The D retention was relatively constant, and about the same as that for solid Li coatings on graphite, or twice that achieved without Li PFC coatings. Contamination of the LLD surface was also possible due to compound formation and erosion and redeposition from carbon PFCs.

Although tungsten is considered the best candidate as a plasma facing component (PFC) in the divertor region in the International Thermonuclear Experimental Reactor (ITER), severe morphology changes such as cavities, blisters, bubbles and nanostructure formation are expected. Increasing defect sinks in the tungsten microstructure is one of the possible solutions to mitigate the irradiation damage. In this work, helium irradiation at low energy (50 and 200 eV) and temperatures of 600 (threshold of vacancy migration) and 950 degrees C were performed on multimodal and ultraflne grained tungsten prepared by spark plasma sintering and severe plastic deformation (SPD), respectively. The multimodal samples consisted of small grains (300-700 nm size) juxtaposed to larger grains (1-3 gm size). Detachment of the small grains was observed in the multimodal grained tungsten irradiated at 600 degrees C and a fluence of 1 x 10(22) m(-2) due to grain boundary grooving. On the same sample but at 950 degrees C, detachment and nanostructuring of the small grains were observed together with recrystallization of the large grains. Irradiation of the SPD samples at 200 eV and 950 degrees C to a fluence of about 2 x 10(22) m(-2), resulted in nanostructuring of the ultrafine grained shear bands in the microstructure. (C) 2012 Elsevier B.V. All rights reserved.

NSTX 2010 experiments were conducted using a molybdenum Liquid Lithium Divertor (LLD) surface installed on the outer part of the lower divertor. This tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. The LLD molybdenum front face has a 45% porosity to provide sufficient wetting to spread 37 g of lithium, and to retain it in the presence of magnetic forces. Lithium Evaporators were used to deposit lithium on the LLD surface. At the beginning of discharges, the LLD lithium surface ranged from solid to liquefied depending on the amount of applied and plasma heating. Noteworthy improvements in plasma performance were obtained similar to those obtained previously with lithiated graphite, e.g., ELM-free, quiescent edge, H-modes. During these experiments with the plasma outer strike point on the LLD, the rate of deuterium retention in the LLD, as indicated by the fueling needed to achieve and maintain stable plasma conditions, was the about the same as that for solid lithium coatings on the graphite prior to the installation of the LLD, i.e., about two times that of no-lithium conditions. The role of lithium impurities in this result is discussed. Following the 2010 experimental campaign, inspection of the LLD found mechanical damage to the plate supports, and other hardware resulting from forces following plasma current disruptions. The LLD was removed, upgraded, and reinstalled. A row of molybdenum tiles was installed inboard of the LLD for 2011 experiments with both inner and outer strike points on lithiated molybdenum to allow investigation of lithium plasma facing issues encountered in the first testing of the LLD. (C) 2011 Elsevier B.V. All rights reserved.

The objective of the materials analysis particle probe (MAPP) in NSTX is to enable prompt and direct analysis of plasma-facing components exposed to plasma discharges. MAPP allows multiple samples to be introduced to the level of the plasma-facing surface without breaking vacuum and analyzed using X-ray photoelectron spectroscopy (XPS), ion-scattering and direct recoil spectroscopy, and thermal desorption spectroscopy (TDS) immediately following the plasma discharge. MAPP is designed to operate as a diagnostic within the ∼12 min NSTX minimum between-shot time window to reveal fundamental plasma-surface interactions. Initial calibration demonstrates MAPP's XPS and TDS capabilities.

Lithium conditioning of plasma-facing surfaces has been implemented in National Spherical Torus Experiment (NSTX) leading to improvements in plasma performance such as reduced D recycling and a reduction in edge localized modes. Analysis of postmortem tiles and offline experiments along with atomistic modeling has identified interactions between Li-O-D and Li-C-D as chemical channels for deuterium retention in ATJ graphite. However, previous surface chemistry analysis of NSTX tiles were conducted postmortem (i.e., after a completed annual campaign), and it was not possible to correlate the performance of particular discharges with the state of the material surface at the time. Materials Analysis Particle Probe (MAPP) is the first in-vacuo surface analysis diagnostic directly integrated into a tokamak and capable of chemical surface analysis of plasma facing samples retrieved from the vessel in between discharges. It uses X-ray photoelectron spectroscopy, direct recoil spectroscopy, low energy ion surface spectroscopy, and thermal desorption spectroscopy to investigate the chemical functionalities between D and lithiated graphite at both the near surface (5-10 nm) and top surface layer (0.3-0.6 nm), respectively. MAPP will correlate plasma facing component surface chemistry with plasma performance and lead the way to improved understanding of plasma-surface interactions and their effect on global plasma performance. Remote operation and data acquisition, integrated into NSTX diagnostic and interlocks, make MAPP an advanced PMI diagnostic with stringent engineering constraints.

Lithium wall conditioning has been found to enhance plasma performance for graphite walled fusion devices such as TFTR, CDX-U, T-11M, TJ-II and NSTX. Among observed plasma enhancements is a reduction in edge density and reduced deuterium recycling. The mechanism by which lithiated graphite retains deuterium is largely unknown. Under controlled laboratory conditions, X-ray photoelectron spectroscopy (XPS) is used to observe the chemical changes that occur on ATJ graphite after lithium deposition. The chemical state of lithiated graphite is found to change upon deuterium irradiation indicating the formation Li-O-D, manifest at 532.9 ± 0.6 eV. Lithium-deuterium interactions are also manifest in the C 1s photoelectron energy range and show Li-C-D interactions at 291.2 ± 0.6 eV. Post-mortem NSTX tiles that have been exposed to air upon extraction are cleaned and examined, revealing the chemical archaeology that formed during NSTX operations. XPS spectra show strong correlation (± 0.3 eV) in Li-O-D and Li-O peaks from post-mortem and control experiments, thus validating offline experiments. We report findings that show that deuterium is found to interact with lithium after lithium has already reacted with carbon and oxygen.

Lithium wall conditioning in TFTR, CDX-U, T-11M, TJ-II and NSTX is found to yield enhanced plasma performance manifest, in part, through improved deuterium particle control. X-ray photoelectron spectroscopy (XPS) experiments examine the affect of D irradiation on lithiated graphite and show that the surface chemistry of lithiated graphite after D ion bombardment (500eV/amu) is fundamentally different from that of non-Li conditioned graphite. Instead of simple LiD bonding seen in pure liquid Li, graphite introduces additional complexities. XPS spectra show that Li–O–D (533.0±0.6eV) and Li–C–D (291.4±0.6eV) bonds, for a nominal Li dose of 2μm, become “saturated” with D at fluences between 3.8 and 5.2×1017cm−2. Atomistic modeling indicate that Li–O–D–C interactions may be a result of multibody effects as opposed to molecular bonding.

Lithium has been utilized to enhance the plasma performance for a variety of fusion devices such as TFTR, CDX-U and NSTX. Lithium in both the solid and liquid states has been studied extensively for its role in hydrogen retention and reduction in sputtering yield. A liquid lithium diverter (LLD) was recently installed in the National Spherical Torus Experiment (NSTX) fusion reactor to investigate lithium applications for plasma-facing surfaces (PFS). Representative samples of LLD material were exposed to lithium depositing and simulated plasma conditions offline at Purdue University to study changes in surface chemical functionalities of Mo, O, Li and D. X-ray photoelectron spectroscopy (XPS) conducted on samples revealed two distinct peak functionalities of lithiated porous molybdenum exposed to deuterium irradiation. The two-peak chemical functionality noticed in porous molybdenum deviates from similar studies conducted on lithiated graphite; such deviation in data is correlated to the complex surface morphology of the porous surface and the correct “wetting” of lithium on the sample surface. The proper lithium “wetting” on the sample surface is essential for maximum deuterium retention and corresponding LLD pumping of deuterium.

Lithium deposition on Grade ATJ graphite substrates and metallic substrates under low-energy D2+ irradiation are compared. Transient and steady-state release rate of ejected species are measured for non-lithiated and lithiated ATJ graphite surfaces. Irradiation fluxes of order 1015 cm−2 s−1 exposed samples while ejected species are monitored with a line-of-sight quadrupole mass spectrometer. For lithiated ATJ graphite the dominant D emission channels are D2O and HDO and indicate the importance of lithium, water breakdown at the surface and oxide formation on desorption dynamics. Exponential decay in the transient release rate of ejected species is found for lithiated ATJ graphite, indicating that near surface super-saturation of D atoms at the vacuum interface in the presence of lithium atoms. In situ X-ray photoelectron spectroscopy (XPS) surface analysis corroborates this result. Lithium-coatings on Mo substrates demonstrate high sputtering rates; however depleted Li/Mo surfaces are quickly recovered when surfaces are heated to temperatures near 500 K.

Lithium conditioning of plasma-facing surfaces (PFS) has been implemented in NSTX leading to improvements in plasma performance such as reduced D recycling and a reduction in edge localized modes (ELMS). Analysis of post-mortem tiles and offline experiments has identified interactions between Li-O-D and Li-C-D as chemical channels for deuterium retention in AT J graphite. MAPP is the first in-vacuo surface analysis diagnostic directly integrated into a tokamak and capable of shot-to-shot chemical surface analysis of plasma material interactions (PM I). X-ray photoelectron spectroscopy (XPS) and low energy ion surface spectroscopy (LEISS) can show the chemical functionalities between D and Iithiated graphite at both the near surface (5-10 nm) and top surface layer (0.3-0.6 nm) for XPS and LEISS respectively. MAPP will correlate plasma facing component (PFC) surface chemistry with plasma performance to lead the way to improved understanding of plasma-surface interactions and their effect on global plasma performance. Remote operation and data acquisition, integrated into NSTX diagnostic and interlocks, make MAPP an advanced PMI diagnostic with stringent engineering constraints.

Lithium evaporation onto National Spherical Torus Experiment (NSTX) plasma-facing components (PFCs) has resulted in improved energy confinement, and reductions in the number and amplitude of edge-localized modes (ELMs) up to the point of complete ELM suppression. The associated PFC surface chemistry has been investigated with a novel plasma–material interface probe connected to an in-vacuo surface analysis station. Analysis has demonstrated that the binding of D atoms to the polycrystalline graphite material of PFCs is fundamentally changed by lithium—in particular, deuterium atoms become weakly bonded near lithium atoms themselves bound to either oxygen or the carbon from the underlying material. Surface dust inside NSTX has been detected in real time using a highly sensitive electrostatic dust detector. In a separate experiment, electrostatic removal of dust via three concentric spiral-shaped electrodes covered by a dielectric and driven by a high-voltage three-phase waveform was evaluated for its potential application to fusion reactors.

Preparation of fine grained, hard and ductile pure tungsten for future fusion reactor applications was tested using the bottom-up approach via powder consolidation by spark plasma sintering (SPS) at different temperature (1300–1800 °C) and pressure (90–266 MPa) conditions. Pure tungsten powders with an average particle size of about 1 μm were sintered to high density (about 94%) with almost no grain growth at a temperature below 1400 °C and an applied pressure up to 266 MPa. These samples had a multi-modal grain size distribution (resembling the size distribution of the initial powder) and a very high Vickers hardness (up to 530 kg/mm2). Above 1500 °C fast grain growth occurred and resulted in a drop in hardness. XRD on the surface of bulk samples showed a small amount of tungsten oxides; however, XPS and EDS indicated that these oxides were only surface contaminants and suggested a high purity for the bulk samples. The results demonstrate that SPS can lead to ultrafine and nanocrystalline tungsten if used to consolidate pure nano tungsten powders.