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There is a widespread need to understand and improve the aging characteristics and properties of catalysts that support essential chemical reactions, including methanol-to-hydrocarbons (MTH) conver...

Future fusion reactors must be able to breed the tritium they will consume. Several breeding and tritium extraction technologies are under investigation internationally. PbLi is of particular inter...

Neutron-irradiated tungsten (W) samples were exposed to helium (He)–seeded deuterium (D) plasmas using a linear plasma device called Tritium Plasma Experiment in order to investigate the synergetic...

Permeation is investigated for the introduction of hydrogen isotopes into lead lithium (PbLi) for the Tritium Extraction eXperiment (TEX). TEX is a forced-convection PbLi loop under construction at Idaho National Laboratory that will test the vacuum permeator (VP) method of tritium extraction from PbLi. The source permeator (SP) delivers atomic hydrogen (H, D, and T) from a gas-phase reservoir into the PbLi via a permeable dense metal membrane. A modular system and a fixed SP system are presented. In the modular design, PbLi flows through the inside of a tubular membrane, and gas-phase hydrogen is introduced on the outside of the membrane. Atomic hydrogen permeates radially inward through the membrane into the PbLi. In the fixed design, PbLi flows into an expansion chamber with closed-ended tubular membranes inserted. Gas-phase hydrogen is introduced on the inside of the closed-ended membranes, and atomic hydrogen permeates radially outward into the flowing PbLi. Hydrogen transport models based on steady-state mass transport through PbLi and permeation through the metal membrane were developed to assess the operation of the SP relative to experimental variables and to allow understanding of uncertain parameter effects, such as PbLi hydrogen transport properties and the effective hydrogen permeability of the VP. This modeling effort considers iron as the SP material and vanadium as the VP material.

In reactor-relevant fusion divertor conditions, tungsten (W) will be used as an armor material due to its excellent thermal properties. It will be exposed to impurities from numerous sources, inclu...

ITER-grade tungsten (W) specimens were exposed to similar deuterium (D) plasma condition (ion flux density of 6.0 × 1021 D m − 2s−1, D ion fluence of 5.0 × 1025 D m − 2) at the surface temperature of 623 K. Thermal desorption spectroscopy was used to measure total D retention in one W specimen after D plasma exposure. Glow-discharge optical emission spectroscopy was used to measure D depth profiling from the other W specimen exposed to the similar condition, and deep D trapping up to 25 μm was observed. When the normalized D depth profile was used with a reaction-diffusion code TMAP7 to model experimental D desorption behavior, an excellent agreement to experimental results was achieved. The modeling results suggested that the predominate mechanism of the deep D trapping might be D trapping in intrinsic intergranular cracks in ITER-grade W. Keyword: Tritium retention, Neutron-irradiation, Plasma facing-components

We investigated deuterium (D) retention in three W samples irradiated with MeV neutrons at high damage level of 0.39 ~ 0.74 displacements per atom (dpa) at high temperatures, 894 K, 1074 K and 1379 K. The W specimens were exposed to high-flux (~1 × 1022 m−2 s−1) and high-fluence (~5 × 1025 m−2) D plasma at 873 K in the Tritium Plasma Experiment. Broad desorption peaks extended from 900 K to 1200 K were observed for the neutron-irradiated W by thermal desorption spectroscopy (TDS). The retention in neutron-irradiated specimens was much larger than for an un-irradiated specimen. The highest D retention was obtained for a specimen irradiated at 894 K. With increasing neutron irradiation temperature, the retention was reduced about by half at 1074 K and further increase of the temperature (1379 K) resulted in comparable retention. In addition, one-dimensional diffusion calculations (D desorption in TDS and D depth distribution in plasma exposure) were performed to derive retention parameters (the detrapping energy, the depth occupied by D atoms and D/W ratio) from experimental D retention properties of neutron-irradiated W. By TDS simulation calculation, simple dependences of the peak temperature, height and width of TDS peaks on the retention parameters were obtained with total retention in the orders of 1019 ~ 1022 m−2. The calculation of the depth distribution of trapped D atoms made a relationship between the D/W ratio and the depth occupied by D atoms after plasma exposure at relevant conditions. By comparing the relationship (the D/W and the depth) with that obtained from the experimental results, we estimate each retention parameters for the specimens irradiated by high-dose neutrons at the high temperatures. And, we discussed the neutron-irradiation temperature dependence of D retentions.

Oxidation kinetics of pure tungsten (W) and tungsten-rhenium (W-Re) alloys with Re contents of 1%, 5% and 15% has been examined in the oxygen gas atmosphere at 873 K. The oxidation kinetics of the W-Re alloys were well characterized by a parabolic rate law. During oxidation of the W-Re alloys, the Re oxide sublimated near the top surface of the oxide layer. The sublimation of the Re oxide may play roles as a sintering agent and/or a stress relief agent in the W oxide layer to be more resistant to oxidation than pure W. There will be an effective value of Re content in W oxide layers for suppression of oxidation depending on oxidation temperatures or atmospheres. The Re oxide in the gas phase could deposit at cooler remote area on materials surface. Furthermore, the Re oxide deposits can be easily moved or transported to different area by dissolving in a water content or a moisture of flowing gas atmosphere.

Tungsten (W) samples were damaged by neutron and 6.4 MeV Fe-ion irradiation above 1000 K simulating the divertor operation temperature. Deuterium (D) retention properties were examined by decorating the damaged W with D and subsequent thermal desorption spectroscopy (TDS) measurements. Vacancy clusters were the major D trapping site in the W irradiated with Fe-ion at 873 K, although D retention by vacancy clusters decreased in the W irradiated with Fe-ion at 1173 K due to dynamic annealing. The D de-trapping activation energy from vacancy clusters was found to be 1.85 eV. D retention in neutron damage W was larger than that damaged by Fe-ion due to the uniform distribution of irradiation defects. The D desorption behaviors from neutron damaged W was simulated well by assuming the D de-trapping activation energy to be 1.52 eV.

Deuterium (D) retention behaviors for the 3 keV Helium (He+) implanted damaged-Tungsten (W) and 10 keV Carbon (C+) - 3 keV He+ simultaneous implanted damaged-W were evaluated by thermal desorption spectroscopy (TDS) to understand the synergetic effect of defect formation and C/He existence on D retention behavior for W with various damage level. For the He+ implantation, the retention of D trapped by dislocation loops was controlled by 3 keV He+ fluence. The D retention in the deeper region was reduced by He+ implantation with higher He+ fluence due to the formation of He bubbles and dense defects at the surface region which would reduce the effective D diffusion coefficient. In addition, in the case of the simultaneous C+ - He+ implantation, the reduction of D retention trapped in the deeper region was also found by the higher C+ - He+ fluence. It can be said that D retention behavior was controlled by the formation of He induced defects and accumulation of He near the surface even if the damages were introduced in the deeper region.

The Idaho National Laboratory (INL) Fusion Safety Program (FSP) is the Office of Fusion Energy Sciences’ (FES) lead laboratory for Magnetic Fusion Energy (MFE) Safety. Our mission is to assist the US and international fusion communities in developing the inherent safety and environmental potential of fusion power by: 1) Developing fusion licensing data and analysis tools, 2) Participating in national and international collaborations and design studies, and 3) Assisting the US and international fusion community in licensing activities and guidance in operational safety. To achieve our mission, the FSP maintains core competencies in the several areas, two of which are: fusion safety code development and tritium retention and permeation in fusion plasma facing component (PFC) and blanket materials. This article details recent accomplishments of our program in these areas and future directions for fusion safety research and development at INL.

Six single crystal tungsten specimens were neutron irradiated to a dose of 0.1 displacements per atom (dpa) at three different irradiation temperatures (633 K, 963 K, and 1073 K) at the High Flux Isotope Reactor in Oak Ridge National Laboratory under the US-Japan PHENIX project. A pair of neutron-irradiated tungsten specimens was exposed to deuterium (D) plasma to D ion fluence of 5.0 × 1025 m−2 at three different exposure temperatures (673 K, 873 K, and 973 K) at the Tritium Plasma Experiment in Idaho National Laboratory. A combination of thermal desorption spectroscopy, nuclear reaction analysis, and rate-diffusion modeling code (Tritium Migration Analysis Program, TMAP) were used to understand D behavior in neutron-irradiated tungsten. A broad D desorption spectrum from the plasma-exposure temperature up to 1173 K was observed. Total D retention up to 1.9 × 1021 m−2 and near-surface D concentrations up to 1.7 × 10−3 D/W were experimentally measured from the 0.1 dpa neutron-irradiated single crystal tungsten. Trap density up to 2.0 × 10−3 Trap/W and detrapping energy ranging from 1.80 to 2.60 eV were obtained from the TMAP modeling.

To evaluate Helium (He) effect on hydrogen isotope retention in tungsten (W), He+ was introduced into W bulk by 201 – 1000 keV He+, or W surface by 3 keV He+ for 6 MeV Fe ion damaged W at room temperature. The deuterium (D) retention behavior was evaluated by thermal desorption spectroscopy (TDS). In addition, the amount of tritium (T) at surface and bulk were separately evaluated by beta-ray induced X-ray spectroscopy (BIXS). The experimental results indicated that the formation of He-void complexes reduced the D trapping in vacancies and voids which have higher trapping energy by the bulk He retention. The BIXS measurement also supported the He enhanced the D reduction in the W bulk region. On the other hand, the He ion irradiation near the surface region enhanced D trapping by dislocation loops or surface, indicating the existence of He near surface interfered the D diffusion toward the bulk. It was concluded that the He existence in bulk or surface will significantly change the D trapping and diffusion behavior in damaged W. Keywords: He in damaged W, Hydrogen isotope retention behavior, Damaged W, Fusion

Glow discharge optical emission spectroscopy (GD-OES) was used to investigate the deep depth profile of deuterium in tungsten exposed to deuterium ions at various fluences. Deuterium was found to continue to decrease in intensity at depths 10–20 µm, even at the lowest fluences, well beyond the probing depths of conventional deuterium measurement techniques. The detection sensitivity was sufficient to allow for definite separation in spectra for all fluences used in the present experiments. The integrated deuterium signal from GD-OES measurement was compared with total deuterium content for samples implanted with deuterium with fluences ranging from 1.0 × 1020 D/m2 to 1.0 × 1022 D/m2. The result showed a clear trend, where the intensity scales with implanted deuterium fluence to the 0.25 power. These results motivate further experiments and calibration of the system for future absolute retention measurements using GD-OES. Keywords: Deuterium, Tungsten, Diffusion, Glow discharge optical emission spectroscopy

Fusion and fission reactions are profoundly dependent on hydrogen for sustained reactions. Fusion is fueled by the isotopes of hydrogen, and predominantly hydrogen-based moderators slow fission neutrons to propagate chain reactions. Intentional tritium production in fusion and concomitant tritium production in fission reactors introduce challenges. The technology to efficiently extract, harvest, and quantify tritium in and from advanced fission molten salt coolants and fusion molten tritium breeder materials will require more pronounced research and development. Measuring and quantifying hydrogen is necessary in all areas of nuclear materials. Although many characterization techniques cannot directly detect hydrogen, numerous techniques provide the necessary information to understand the behavior of hydrogen in nuclear materials.

To investigate deuterium retention and the onset of blistering, deuterium was implanted in cold rolled tungsten foils at fluences ranging from 3 ×1020 to 3 ×1022D/m2. Ion energies were 300eV and 2000eV in order to be below and above the tungsten theoretical damage energy threshold. While energy dependent phenomena were observed, blistering occurs regardless of ion energy. Both plastically and elastically deformed blisters were found, as manifest in before and after micrographs. The fraction of plastically deformed blisters did not saturate at the fluences used in these studies. However, the size of the largest blister that relaxed during TDS does saturate at ∼7µm. A simple conceptual model is presented, which proposes that the deuterium released from elastically deformed blisters appears at ∼600K in the thermal desorption spectra, which is consistent with large vacancy clusters. Keywords: Tungsten, Deuterium, Retention, Blistering, Thermal desorption spectroscopy

Deuterium retention behavior in tungsten damaged by fast neutrons at high temperatures (0.43 dpa at 918 K and 0.74 dpa at 1079 K) and 6.4 MeV Fe2+ (0.3 dpa at R.T.) were investigated to evaluate the tritium retention property of fusion reactor divertors. A deuterium gas absorption method was carried out to avoid additional damage that may be induced by plasma exposure, then, deuterium retention and desorption behaviors were investigated quantitatively by means of thermal desorption spectroscopy and the following simulation code. The deuterium desorption spectra for tungsten samples were analyzed by the numerical code which includes the elementary steps of hydrogen isotope migration processes including diffusion, trapping, detrapping, and surface recombination. The evaluated deuterium detrapping energy from the irradiation defects in neutron irradiated tungsten sample was larger than that in 6.4 MeV Fe2+ irradiated tungsten. It was suggested that the dominant deuterium trapping site in the neutron irradiated tungsten would be voids which was formed by the accumulation of vacancies during neutron irradiation under high temperature and long duration.

Understanding and managing D retention in plasma facing components is essential for tritium safety in fusion reactors. Neutron irradiated and virgin low carbon arc cast (LCAC) Mo, as well as Mo foil samples with and without He pre-irradiation, were used to investigate D retention. D and He retention were investigated simultaneously in thermal desorption spectroscopy using a high resolution residual gas analyzer. Results show a significant increase in D retention with He pre-irradiation. Vacancies and vacancy clusters are found to retain D in LCAC samples, but neutron irradiated Mo retains more D in vacancy clusters.

The Tritium Plasma Experiment (TPE) is a unique high-flux linear plasma device that can handle beryllium, tritium, and neutron-irradiated plasma facing materials, and is the only existing device dedicated to evaluate in-vessel tritium inventory in the nuclear environment for fusion safety. The electrical upgrade were recently carried out to enhance operational safety and to improve plasma performance. New DC power supplies and a new control center enable remote plasma operations from outside of the contamination area for tritium, minimizing the possible exposure risk with tritium and beryllium and eliminating heat stress issue. In November 2015, the TPE successfully achieved first deuterium plasma via remote operation after a significant three-year upgrade. Simple linear scaling estimate showed that the TPE is expected to achieve Γimax of >1.0 × 1023 m−2 s−1 and qheat of >1 MW m−2 with new power supplies. This upgrade not only improves operational safety of the worker, but also enhances plasma performance to better simulate extreme plasma-material conditions expected in ITER, FNSF, and DEMO for improving in-vessel tritium inventory assessment in fusion nuclear environment.

Dense vanadium-based membranes offer high permeability and perfect selectivity to hydrogen isotopes, maintain favorable neutronic properties, and are compatible with liquid metals such as PbLi. These properties make vanadium membranes a promising fusion fuel cycle technology for processes such as tritium extraction from PbLi and exhaust processing. Surface contamination has a deleterious effect on the gas-phase hydrogen permeation through vanadium, and the reported permeabilities range from 10-14 to 10-7 mol m-1 s-1 Pa-0.5. Thin dense films of palladium applied to clean vanadium surfaces enable a consistently high hydrogen permeability. In this study, uncoated vanadium resulted in deuterium permeabilities ranging from 2.8 × 10-11 to 6.4 × 10-9 mol m-1 s-1 Pa-0.5 at 300 °C–700 °C, respectively. Post-test analysis revealed a VOx surface layer and VCx subsurface layer formed on the feed side, while the as-received surface oxide dissolved leaving a submonolayer oxide on the permeate surface. The Pd-coated V resulted in a maximum deuterium permeability of 2.1 × 10-7 mol m-1 s-1 Pa-0.5 at 375 °C upon activation of the Pd surface by oxidation and reduction. The deuterium permeation declined upon heating to 500 °C due to intermetallic diffusion between the Pd and V. The Mo2C-coated V resulted in deuterium permeabilities ranging from 2.7 × 10-10 to 1.8 × 10-9 at 500 °C–700 °C, respectively, and a post-test analysis found the carbon in the Mo2C layer had dissolved into the V near the interface.

Single crystalline W (tungsten) samples irradiated at 633, 963 and 1073 K by neutrons to a damage level of 0.1 dpa were exposed to a high-flux D (deuterium) plasma at 673, 873 and 973 K, respectively, in TPE (Tritium Plasma Experiment) at INL (Idaho National Laboratory). Deuterium desorption was analyzed by TDS (Thermal Desorption Spectroscopy), and D depth profiles were determined by NRA (Nuclear Reaction Analysis) at SNL (Sandia National Laboratories). HIDT (Hydrogen Isotope Diffusion and Trapping) simulation code was applied to evaluate D behavior for neutron-damaged W at higher temperature. The D retention at depths up to 3 μm for the neutron-damaged sample at 673 K was two orders of magnitude larger than that for undamaged tungsten, and its D desorption spectrum had a single broad stage at around 900 K. As the neutron irradiation/plasma exposure temperature increased, D retention was largely reduced, and the desorption temperature was shifted to higher temperatures above 1100 K. The D depth profiles by NRA also showed D migration toward bulk by higher temperature irradiation, compared to undamaged W. The HIDT simulation indicated that the major binding energy of D was changed from 1.43 eV to 2.07 eV at higher neutron irradiation and plasma exposure temperatures, suggesting that some vacancies and small vacancy clusters would aggregate to form larger voids, or depopulation of weak traps at high D plasma exposure temperatures. It can be said that more stable trapping sites played dominant roles in the D retention at higher neutron irradiation and plasma exposure temperature. The binding energy by HIDT simulation was almost consistent with the reported value by TMAP, but the consideration of not only total D retention measured by TDS but also D depth profile by NRA led to the more accurate D behavior in neutron-damaged W.

A new combined positron annihilation lifetime spectroscopy (PALS) and coincidence Doppler broadening (CDB) system has been constructed at Idaho National Laboratory (INL). The design and commissioning process are discussed with practical guidelines that can be followed for the installation of new systems. In addition, the new system was used to investigate defect structure in titanium and titanium deuteride samples where the deuterium concentration ranges from 0.10 to 2 x 10−5. The TiD samples were found to be largely defect free based on the positron lifetimes and S- W curves. Ratio curves of the CDB spectra showed elemental sensitivity to the deuterium in the samples, that increased with deuterium concentration.

Dispersion-strengthened tungsten materials represent a new class of W-based materials to be investigated for use as plasma-facing component in nuclear fusion reactors. However, the retention and permeation characteristics of these materials under low energy deuterium (D) irradiation need to be elucidated before the efficacy of these materials can be judged. Due to possible deep penetration of D in W, depth profile techniques such as glow discharge optical emission spectroscopy (GD-OES) are needed to probe D concentrations many microns beneath the material surface. In this study, the D retention behavior of W materials containing 1–10 wt% TaC, TiC, or ZrC are investigated with GD-OES. After exposure to a 5 × 1018cm−2 D ion fluence at 100 °C, D was observed beyond the D implantation depth the surface in many specimens, and the D depth profile was found to depend on the type and concentration of the added second phase. Combined with in-situ X-ray Photoelectron Spectroscopy (XPS) studies, the effects of impurity oxygen atoms on the D retention is considered, as an increasing oxygen content correlates with decreased D retention. The influence of grain size, second phase particles, and oxygen content on the retained D depth and concentration in these complex W-based materials is discussed.

The role of grain boundaries in limiting irradiation damage in nanocrystalline materials is often correlated with the grain boundary sink efficiency. Here, we demonstrate on a tungsten material system (which has very distinct vacancy and interstitial mobilities) that sink efficiency does not unequivocally describe how grain boundaries impact irradiation damage. Rather, it reflects a particular defect diffusion equation that can change if any of the bulk conditions change. Even when denuded zone formation does not occur and grain boundaries have zero sink efficiencies, grain boundaries still impact the performance of nanocrystalline materials under irradiation by acting as a saturable defect storage site. However, denuded zone formation can occur under a necessary requirement of extra defect recombination at the grain boundaries (which, for example, is not possible when vacancy migration does not occur). These insights provide answers to several outstanding questions regarding the sink efficiency of a grain boundary and assist in parametrizing the role of grain boundaries in limiting irradiation damage in nanocrystalline materials.

The present work is a continuation of a recent research to develop and optimize positron annihilation spectroscopy (PAS) for characterizing neutron-irradiated tungsten. Tungsten samples were exposed to neutrons in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory and damaged to 0.025 and 0.3 dpa. Subsequently, they were exposed to deuterium plasmas in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory. The implanted deuterium was desorbed through sample heating to 900 °C, and Doppler broadening (DB)-PAS was performed both before and after heating. Results show that deuterium impregnated tungsten is identified as having a smaller S-parameter. The S-parameter increases after deuterium desorption. Microstructural changes also occur during sample heating. These effects can be isolated from deuterium desorption by comparing the S-parameters from the deuterium-free back face with the deuterium-implanted front face. The application of using DB-PAS to examine deuterium retention in tungsten is examined.

The unique ability of grain boundaries to act as effective sinks for radiation damage plays a significant role in nanocrystalline materials due to their large interfacial area per unit volume. Leveraging this mechanism in the design of tungsten as a plasma-facing material provides a potential pathway for enhancing its radiation tolerance under fusion-relevant conditions. In this study, we explore the impact of defect microstructures on the mechanical behavior of helium ion implanted nanocrystalline tungsten through nanoindentation. Softening was apparent across all implantation temperatures and attributed to bubble/cavity loaded grain boundaries suppressing the activation barrier for the onset of plasticity via grain boundary mediated dislocation nucleation. An increase in fluence placed cavity induced grain boundary softening in competition with hardening from intragranular defect loop damage, thus signaling a new transition in the mechanical behavior of helium implanted nanocrystalline tungsten.

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 similar to 850 C desorbed all deuterium complexes observed in the 0 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 Li-O 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(C) 2013 Elsevier B.V. All rights reserved.

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.

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.

Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last six years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a dual lithium evaporation system which can evaporate up to ∼160 g of lithium onto the lower divertor plates between re-loadings. The unique feature of the NSTX lithium research program is that it can investigate the effects of lithium coated plasma-facing components in H-mode divertor plasmas. This lithium evaporation system has produced many intriguing and potentially important results. In 2010, the NSTX lithium program has focused on the effects of liquid lithium divertor (LLD) surfaces including the divertor heat load, deuterium pumping, impurity control, electron thermal confinement, H-mode pedestal physics, and enhanced plasma performance. To fill the LLD with lithium, 1300 g of lithium was evaporated into the NSTX vacuum vessel during the 2010 operations. The routine use of lithium in 2010 has significantly improved the plasma shot availability resulting in a record number of plasma shots in any given year. In this paper, as a follow-on paper from the 1st lithium symposium [1] , we review the recent progress toward developing fundamental understanding of the NSTX lithium experimental observations as well as the opportunities and associated R&D required for use of lithium in future magnetic fusion facilities including ITER.

Chemistry as well as sputtering and reflection dynamics of lithiated carbon material, bombarded by slow hydrogen atoms are studied. We present a realistic method for computational simulation of the dynamics of the polar Li–C–O–H material dynamics. It is based on an approximate, semi-empirical quantum mechanics of electrons and classical mechanics of nuclei. Results are validated qualitatively by comparison with experiments and with a first principle DFT computations. In particular, we explain observed details of the hydrogen bonding chemistry in lithiated carbon, showing that incoming hydrogen interacts preferably with Li-C rather than C structures.

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.

NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma-facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Zeff and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core

High (approximate to 90%) deuterium retention was observed in NSTX gas balance measurements both with- and without lithiumization of the carbon plasma-facing components. The gas retained in ohmic discharges was measured by comparing the vessel pressure rise after a discharge to that of a gas-only pulse with the pumping valves closed. For neutral beam heated discharges the gas input and gas pumped by the NB cryopanels were tracked. The discharges were followed by outgassing of deuterium that reduced the retention. The relationship between retention and surface chemistry was explored with a new plasma-material interface probe connected to an in vacuo surface science station that exposed four material samples to the plasma. XPS and TDS analysis demonstrated that binding of D atoms in graphite is fundamentally changed by lithium - in particular atoms are weakly bonded in regions near lithium atoms bound to either oxygen or the carbon matrix. This is in contrast to the strong ionic bonding that occurs between D and pure Li. (C) 2010 Elsevier B.V. All rights reserved.