CDX-U Research Articles

Energy, angle, and temperature dependencies of the sticking of D atoms on Li surfaces

Detailed experimental and computational information on the response of lithium surfaces to irradiation by slow hydrogenic particles (ions, atoms, molecules) is sparse and mainly speculative. In this work, we present a computational study of the reflection and retention of deuterium (D) atoms at crystalline and amorphous lithium surfaces at 300 and 500 K, where the D atoms have an impact energy in the range of 0.025–5 eV and incident angles of 0° (perpendicular incidence) or 85° (near-grazing incidence). Classical molecular dynamics simulations are performed with the reactive bond-order force field (ReaxFF) potentials. This study provides quantitative information on the deuterium sticking probability and recycling coefficient for lithium surfaces. Our results support the ongoing work at the Lithium Tokamak eXperiment-β fusion experiment as well as relevant experiments in the laboratory setting.

Initial Results From the Newly Upgraded LTX-β

LTX-β, the upgraded Lithium Tokamak Experiment, recently began its first campaign with a goal to study the transport properties of gradient-free electron temperature profile equilibria with increased toroidal field and neutral beam injection. The temperature-gradient-free equilibria in LTX were enabled by the lithium plasma-facing surface. To make the lithium surface conditions more consistent for LTX-β, a new, faster lithium evaporator system has been developed. The toroidal magnetic field has been nearly doubled to ≥0.3 T. Auxiliary heating and fueling with a high flux (35 A, 20 kV) neutral beam has begun. The beam provides core fueling without cold edge neutrals, to test whether flat temperature profiles can be sustained, and makes possible core ion temperature measurements via charge exchange recombination spectroscopy. Neutral beam fueling has been demonstrated with increases in line-integrated density. The coupling of the beam matches well with NUBEAM predictions for shine through. Improved performance has been observed following lithium evaporative coatings and an increase in field strength.

Plasma Facing Component Characterization and Correlation With Plasma Conditions in Lithium Tokamak Experiment-β

Lithium coatings in the Lithium Tokamak eXperiment (LTX) led to flat temperature profiles. The flat temperature profiles were observed along with a hot, low density edge, implying a broad, collisionless scrape-off layer (SOL). Additionally, in vacuo X-ray photoelectron spectroscopy (XPS) measurements established that lithium coatings evaporatively deposited onto high-Z plasma facing components (PFCs) became oxidized while retaining the ability to achieve good plasma performance long after lithium was applied to the PFCs. Longstanding theory predicted flat temperature profiles with low recycling walls, which was presumed to be due to hydrogen binding with elemental lithium to form lithium hydride. The presence of oxidized lithium, however, raised questions regarding the exact mechanism of hydrogen retention in LTX. To investigate these questions, the upgraded facility LTX- $\beta $ includes a new sample exposure probe (SEP) for more detailed in vacuo analysis of PFC samples. The SEP is equipped with a vacuum suitcase capable of transporting samples representative of the LTX- $\beta $ outer midplane PFCs to a stand-alone XPS system while maintaining pressures lower than the LTX- $\beta $ base vacuum to limit the contamination between sample exposure and analysis. The low-energy resolution XPS system used in past experiments could only enable the determination of elemental percentages on the PFC sample surfaces. Because the new XPS system has higher energy resolution, it is more direct to assign chemical compounds to the measured binding energies. This capability has been confirmed by comparing XPS data from PFC test samples with measurements using a commercial high-resolution XPS system. Quartz crystal microbalances (QCMs) were used to quantify the thickness of the deposited lithium on the LTX- $\beta $ PFCs. This article describes the application of the SEP to characterize the PFC surfaces using XPS and their relationship to plasma conditions.

Design and calibration of a retarding field energy analyzer for the LTX-β scrape off layer and modeling of electrostatic potential in a collisionless SOL

Abstract The Lithium Tokamak eXperiment (LTX) is a spherical tokamak device designed to study lithium plasma facing components (PFCs). The lithium coated wall of LTX has been demonstrated to produce a plasma edge with high electron temperature (200 eV or greater). Plasma density in the outer scrape-off layer (SOL) is also found to be very low, around 2 × 1017m-3, as a result of the low recycling lithium boundary. The high temperature, low collisionality region of the plasma extends into the SOL. The recent upgrade to LTX-β includes installation of a neutral beam, which will provide further heating and fueling of the core plasma. Core and edge diagnostics will also be expanded. As part of this expansion, a Retarding Field Energy Analyzer (RFEA) has been developed for the SOL of LTX-β. Measurements of the ion temperature, ion energy distribution, and the local space potential will be performed in the SOL plasma using this RFEA. Upgraded high field side (HFS) and low field side (LFS) Langmuir probes will replace existing triple probes so that higher electron temperatures can be more reliably measured. The HFS probes are also positioned to give radial and vertical gradient measurements. The design of the RFEA will be presented, along with calibration data. Since a high temperature, low collisional edge is expected for LTX-β, with a high mirror ratio near the LCFS (around 4), the majority of particles in the SOL will be mirror-trapped. Trapped particle effects will therefore become significant in the physics of the SOL plasma, and warrant further theoretical investigations. Here we present a theoretical study of the ambipolar potential formed in the collisionless SOL via differential loss of the electrons and ions, known as the Pastukhov potential in the literature. Numerical results will also be presented.

The charge exchange recombination spectroscopy diagnostic on the upgraded Lithium Tokamak eXperiment (LTX-β).

The Lithium Tokamak eXperiment has undergone an upgrade to LTX-β, a major part of which is the addition of neutral beam injection (NBI). NBI has allowed for a new charge exchange recombination spectroscopy (CHERS) system to be installed in order to measure impurity concentrations, ion temperature, and toroidal velocity. Previously on LTX measuring these parameters relied on passive spectroscopy and inversion techniques and had large uncertainty. The CHERS system has 52 total views, split into four groups of 13, half facing toward the beam and half symmetrically facing away from the beam, so the background non-beam related emission can be simultaneously subtracted. Both sets of views sample a major radius of 27-59 cm, with resolution through the beam of 1.5-2.5 cm. LTX-β is expected to have its magnetic axis near 35 cm, with minor radii of 18-23 cm. Three separate spectrometers will be used for the diagnostic, giving the system great flexibility to simultaneously measure emission from multiple impurity lines. The viewing optics are f/1.8, allowing all of the spectrometers to be fully illuminated. Design and calibration of the system as well as the advantages of various configurations of the spectrometers will be highlighted.

Magnetic perturbation diagnostics in the high-temperature lithiated environment of LTX-β.

Magnetic perturbation measurements will be invaluable for characterizing Lithium Tokamak Experiment Beta (LTX-β) plasmas due to the time-evolving 3D nature of the magnetic fields generated by eddy currents in the vessel and copper shell segments, as well as enhanced MHD instability drive due to newly introduced neutral beam heating. The LTX-β upgrade includes two new arrays of Mirnov coils: a shell eddy sensor array of two-axis coils distributed over the back surface of one shell segment and a toroidal array of poloidal field coils at the low-field side midplane gap. Evaporative lithium wall-coating and the high temperatures required for liquid lithium wall operation both complicate the implementation of in-vessel diagnostics. While the shell array is protected from lithium exposure, the shell segment to which it is mounted will at times exceed 300 °C. The toroidal array, however, will experience direct line-of-sight exposure to the lithium evaporator as well as close proximity to the hot shell and may also be subject to poorly confined beam-driven fast ions. We describe how the two new Mirnov coil arrays meet these environmental challenges and enhance the LTX-β diagnostic suite.

Observation of Flat Electron Temperature Profiles in the Lithium Tokamak Experiment.

It has been predicted for over a decade that low-recycling plasma-facing components in fusion devices would allow high edge temperatures and flat or nearly flat temperature profiles. In recent experiments with lithium wall coatings in the Lithium Tokamak Experiment (LTX), a hot edge (>200 eV) and flat electron temperature profiles have been measured following the termination of external fueling. Reduced recycling was demonstrated by retention of ∼60% of the injected hydrogen in the walls following the discharge. Electron energy confinement followed typical Ohmic confinement scaling during fueling, but did not decrease with density after fueling terminated, ultimately exceeding the scaling by ∼200%. Achievement of the low-recycling, hot edge regime has been an important goal of LTX and lithium plasma-facing component research in general, as it has potentially significant implications for the operation, design, and cost of fusion devices.

A frequency-modulated continuous-wave reflectometer for the Lithium Tokamak Experiment.

The frequency-modulated continuous-wave reflectometer on LTX (Lithium Tokamak Experiment) and the data analysis methods used for determining electron density profiles are described. The diagnostic uses a frequency range of 13.1-33.5 GHz, for covering a density range of 0.21-1.4×1013 cm-3 (in O-mode polarization) with a time resolution down to 8 μs. The design of the diagnostic incorporates the concept of an "optimized" source frequency sweep, which minimizes the large variation in the intermediate frequency signal due to a long dispersive transmission line. The quality of the raw data is dictated by the tuning characteristics of the microwave sources, as well as the group delay ripple in the transmission lines, which can generate higher-order nonlinearities in the frequency sweep. Both effects are evaluated for our diagnostic and best practices are presented for minimizing "artifacts" generated in the signals. The quality of the reconstructed profiles is also improved using two additional data analysis methods. First, the reflectometer data are processed as a radar image, where clutter due to echoes from the wall and backscattering from density fluctuations can be easily identified and removed. Second, a weighed least-squares lamination algorithm POLAN (POLynomial ANalysis) is used to reconstruct the electron density profile. Examples of density profiles in LTX are presented, along with comparisons to measurements from the Thomson scattering and the λ = 1 mm interferometer diagnostics.

Equilibrium reconstruction with 3D eddy currents in the Lithium Tokamak eXperiment

Axisymmetric free-boundary equilibrium reconstructions of tokamak plasmas in the Lithium Tokamak eXperiment (LTX) are performed using the PSI-Tri equilibrium code. Reconstructions in LTX are complicated by the presence of long-lived non-axisymmetric eddy currents generated by a vacuum vessel and first wall structures. To account for this effect, reconstructions are performed with additional toroidal current sources in these conducting regions. The eddy current sources are fixed in their poloidal distributions, but their magnitude is adjusted as part of the full reconstruction. Eddy distributions are computed by toroidally averaging currents, generated by coupling to vacuum field coils, from a simplified 3D filament model of important conducting structures. The full 3D eddy current fields are also used to enable the inclusion of local magnetic field measurements, which have strong 3D eddy current pick-up, as reconstruction constraints. Using this method, equilibrium reconstruction yields good agreement with all available diagnostic signals. An accompanying field perturbation produced by 3D eddy currents on the plasma surface with a primarily n = 2, m = 1 character is also predicted for these equilibria.

Compatibility of lithium plasma-facing surfaces with high edge temperatures in the Lithium Tokamak Experiment

High edge electron temperatures (200 eV or greater) have been measured at the wall-limited plasma boundary in the Lithium Tokamak Experiment (LTX). Flat electron temperature profiles are a long-predicted consequence of low recycling boundary conditions. Plasma density in the outer scrape-off layer is very low, 2–3 × 1017 m−3, consistent with a low recycling metallic lithium boundary. Despite the high edge temperature, the core impurity content is low. Zeff is estimated to be ∼1.2, with a very modest contribution (<0.1) from lithium. Experiments are transient. Gas puffing is used to increase the plasma density. After gas injection stops, the discharge density is allowed to drop, and the edge is pumped by the low recycling lithium wall. An upgrade to LTX–LTX-β, which includes a 35A, 20 kV neutral beam injector (on loan to LTX from Tri-Alpha Energy) to provide core fueling to maintain constant density, as well as auxiliary heating, is underway. LTX-β is briefly described.

Responsivity calibration of the LoWEUS spectrometer.

We performed an in situ calibration of the relative responsivity function of the Long-Wavelength Extreme Ultraviolet Spectrometer (LoWEUS), while operating on the Lithium Tokamak Experiment (LTX) at Princeton Plasma Physics Laboratory. The calibration was accomplished by measuring oxygen lines, which are typically present in LTX plasmas. The measured spectral line intensities of each oxygen charge state were then compared to the calculated emission strengths given in the CHIANTI atomic database. Normalizing the strongest line in each charge state to the CHIANTI predictions, we obtained the differences between the measured and predicted values for the relative strengths of the other lines of a given charge state. We find that a 3rd degree polynomial function provides a good fit to the data points. Our measurements show that the responsivity between about 120 and 300 Å varies by factor of ∼30.

Hydrogen retention in lithium on metallic walls from “in vacuo” analysis in LTX and implications for high-Z plasma-facing components in NSTX-U

The application of lithium to plasma-facing components (PFCs) has long been used as a technique for wall conditioning in magnetic confinement devices to improve plasma performance. Determining the characteristics of PFCs at the time of exposure to the plasma, however, is difficult because they can only be analyzed after venting the vacuum vessel and removing them at the end of an operational period. The Materials Analysis and Particle Probe (MAPP) addresses this problem by enabling PFC samples to be exposed to plasmas, and then withdrawn into an analysis chamber without breaking vacuum. The MAPP system was used to introduce samples that matched the metallic PFCs of the Lithium Tokamak Experiment (LTX). Lithium that was subsequently evaporated onto the walls also covered the MAPP samples, which were then subject to LTX discharges. In vacuo extraction and analysis of the samples indicated that lithium oxide formed on the PFCs, but improved plasma performance persisted in LTX. The reduced recycling this suggests is consistent with separate surface science experiments that demonstrated deuterium retention in the presence of lithium oxide films. Since oxygen decreases the thermal stability of the deuterium in the film, the release of deuterium was observed below the lithium deuteride dissociation temperature. This may explain what occurred when lithium was applied to the surface of the NSTX Liquid Lithium Divertor (LLD). The LLD had segments with individual heaters, and the deuterium-alpha emission was clearly lower in the cooler regions. The plan for NSTX-U is to replace the graphite tiles with high-Z PFCs, and apply lithium to their surfaces with lithium evaporation. Experiments with lithium coatings on such PFCs suggest that deuterium could still be retained if lithium compounds form, but limiting their surface temperatures may be necessary.

A dual wavelength imaging system for plasma-surface interaction studies on the National Spherical Torus Experiment Upgrade.

A two-channel spectral imaging system based on a charge injection device radiation-hardened intensified camera was built for studies of plasma-surface interactions on divertor plasma facing components in the National Spherical Torus Experiment Upgrade (NSTX-U) tokamak. By means of commercially available mechanically referenced optical components, the two-wavelength setup images the light from the plasma, relayed by a fiber optic bundle, at two different wavelengths side-by-side on the same detector. Remotely controlled filter wheels are used for narrow bandpass and neutral density filters on each optical path allowing for simultaneous imaging of emission at wavelengths differing in brightness up to 3 orders of magnitude. Applications on NSTX-U will include the measurement of impurity influxes in the lower divertor strike point region and the imaging of plasma-material interaction on the head of the surface analysis probe MAPP (Material Analysis and Particle Probe). The diagnostic setup and initial results from its application on the lithium tokamak experiment are presented.

High performance discharges in the Lithium Tokamak eXperiment with liquid lithium wallsa)

The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10× compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid, exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started.

Dependence of LTX plasma performance on surface conditions as determined by in situ analysis of plasma facing components

The Materials Analysis and Particle Probe (MAPP) diagnostic has been implemented on the Lithium Tokamak Experiment (LTX) at PPPL, providing the first in situ X-ray photoelectron spectroscopy (XPS) surface characterization of tokamak plasma facing components (PFCs). MAPP samples were exposed to argon glow discharge conditioning (GDC), lithium evaporations, and hydrogen tokamak discharges inside LTX. Samples were analyzed with XPS, and alterations to surface conditions were correlated against observed LTX plasma performance changes. Argon GDC caused the accumulation of nm-scale metal oxide layers on the PFC surface, which appeared to bury surface carbon and oxygen contamination and thus improve plasma performance. Lithium evaporation led to the rapid formation of a lithium oxide (Li2O) surface; plasma performance was strongly improved for sufficiently thick evaporative coatings. Results indicate that a 5h argon GDC or a 50nm evaporative lithium coating will both significantly improve LTX plasma performance.

VUV/XUV measurements of impurity emission in plasmas with liquid lithium surfaces on LTX

The VUV/XUV spectrum has been measured on the Lithium Tokamak eXperiment (LTX) using a transmission grating imaging spectrometer (TGIS) coupled to a direct-detection x-ray charge-coupled device camera. TGIS data show significant changes in the ratios between the lithium and oxygen impurity line emission during discharges with varying lithium wall conditions. Lithium coatings that have been passivated by lengthy exposure to significant levels of impurities contribute to a large O/Li ratio measured during LTX plasma discharges. Furthermore, previous results have indicated that a passivated lithium film on the plasma facing components will function as a stronger impurity source when in the form of a hot liquid layer compared to a solid lithium layer. However, recent TGIS measurements of plasma discharges in LTX with hot stainless steel boundary shells and a fresh liquid lithium coating show lower O/Li impurity line ratios when compared to discharges with a solid lithium film on cool shells. These new measurements help elucidate the somewhat contradictory results of the effects of solid and liquid lithium on plasma confinement observed in previous experiments.

High-resolution grazing-incidence grating spectrometer for temperature measurements of low-Z ions emitting in the 100-300 Å spectral band.

We have constructed a high-resolution grazing-incidence spectrometer designed for measuring the ion temperature of low-Z elements, such as Li(+) or Li(2 +), which radiate near 199 Å and 135 Å, respectively. Based on measurements at the Livermore Electron Beam Ion Trap we have shown that the instrumental resolution is better than 48 mÅ at the 200 Å setting and better than 40 mÅ for the 135-Å range. Such a high spectral resolution corresponds to an instrumental limit for line-width based temperature measurements of about 45 eV for the 199 Å Li(+) and 65 eV for the 135 Å Li(2 +) lines. Recently obtained survey spectra from the Lithium Tokamak Experiment at the Princeton Plasma Physics Laboratory show the presence of these lithium emission lines and the expected core ion temperature of approximately 70 eV is sufficiently high to demonstrate the feasibility of utilizing our high-resolution spectrometer as an ion-temperature diagnostic.

Magnetic diagnostics for equilibrium reconstructions with eddy currents on the Lithium Tokamak eXperiment.

The Lithium Tokamak eXperiment is a spherical tokamak with a close-fitting low-recycling wall composed of thin lithium layers evaporated onto a stainless steel-lined copper shell. Long-lived non-axisymmetric eddy currents are induced in the shell and vacuum vessel by transient plasma and coil currents and these eddy currents influence both the plasma and the magnetic diagnostic signals that are used as constraints for equilibrium reconstruction. A newly installed set of re-entrant magnetic diagnostics and internal saddle flux loops, compatible with high-temperatures and lithium environments, is discussed. Details of the axisymmetric (2D) and non-axisymmetric (3D) treatments of the eddy currents and the equilibrium reconstruction are presented.

Development progress of the Materials Analysis and Particle Probe.

The Materials Analysis and Particle Probe (MAPP) is a compact in vacuo surface science diagnostic, designed to provide in situ surface characterization of plasma facing components in a tokamak environment. MAPP has been implemented for operation on the Lithium Tokamak Experiment at Princeton Plasma Physics Laboratory (PPPL), where all control and analysis systems are currently under development for full remote operation. Control systems include vacuum management, instrument power, and translational/rotational probe drive. Analysis systems include onboard Langmuir probes and all components required for x-ray photoelectron spectroscopy, low-energy ion scattering spectroscopy, direct recoil spectroscopy, and thermal desorption spectroscopy surface analysis techniques.

Overview of Innovative PMI Research on NSTX-U and Associated PMI Facilities at PPPL

Developing a reactor compatible divertor and managing the associated plasma material interaction (PMI) has been identified as a high priority research area for magnetic confinement fusion. Accordingly on NSTXU, the PMI research has received a strong emphasis. With ~15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak projected divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m2. To support the PMI research, a comprehensive set of PMI diagnostic tools are being implemented. The snow-flake configuration can produce exceptionally high divertor flux expansion of up to ~50. Combined with the radiative divertor concept, the snowflake configuration has reduced the divertor heat flux by up to an order of magnitude in NSTX. Another area of active PMI investigation is the effect of divertor lithium coating (both in solid and liquid phases). The overall NSTX lithium PFC coating results suggest exciting opportunities for future magnetic confinement research including significant electron energy confinement improvements, H-mode power threshold reduction, the control of Edge Localized Modes (ELMs), and high heat flux handling. To support the NSTX-U/PPPL PMI research, there are also a number of associated PMI facilities implemented at PPPL/Princeton University including the Liquid Lithium R&D facility, Lithium Tokamak Experiment, and Laboratories for Materials Characterization and Surface Chemistry.