Russ Doerner

Efficient Plasma Route to Nanostructure Materials: Case Study on the Use of m-WO3 for Solar Water Splitting

One of the main challenges in developing highly efficient nanostructured photoelectrodes is to achieve good control over the desired morphology and good electrical conductivity. We present an efficient plasma-processing technique to form porous structures in tungsten substrates. After an optimized two-step annealling procedure, the mesoporous tungsten transforms into photoactive monoclinic WO3. The excellent control over the feature size and good contact between the crystallites obtained with the plasma technique offers an exciting new synthesis route for nanostructured materials for use in processes such as solar water splitting.

Investigation of plasma exposed W-1% La2O3 tungsten in a high ion flux, low ion energy, low carbon impurity plasma environment for the International Thermonuclear Experimental Reactor

Abstract Previous investigations of tungsten for the International Thermonuclear Experimental Reactor (ITER) were focusing on using energetic ion beams whose energies were over 1 keV. This study presents experimental results of exposed W–1% La 2 O 3 in high ion flux (10 22 m –2 ), low ion energies (about 110 eV) steady-state deuterium plasmas at elevated temperatures (873–1250 K). The tungsten samples are floating during plasma exposure. Using a high-pressure gas analyzer, the residual carbon impurities in the plasma are found to be about 0.25%. No carbon film is detected on the surface by the EDX analysis after plasma exposure. An infrared pyrometer is also used as an in situ detector to monitor the surface emissivities of the substrates during plasma exposure. Using the scanning electron microscopy, microscopic pits of sizes ranging from 0.1 to 5 μm are observed on the plasma exposed tungsten surfaces. These pits are believed to be the results of erupted deuterium gas bubbles, which recombine underneath the surface at defect locations and grain boundaries, leading to substrate damage and erosion loss of the substrate material. Low temperature plasma exposure of a tungsten foil indicates that deuterium gas (D 2 ) is trapped inside the substrate. Macroscopic blisters are observed on the surface. The erosion yield of the W–1% La 2 O 3 increases with temperature and seems to saturate at around 1050 K. Scattered networks of bubble sites are found 5 μm below the substrate surface. High temperature plasma exposure appears to reduce the population as well as the size of the pits. The plasma exposed W–1% La 2 O 3 substrates, exposed above 850 K, retain about 10 19 D/m 2 , which is two orders of magnitude less than those retained by the tungsten foils exposed at 400 K.

Experimentally Determined Neutral Density and Plasma Parameters in a 30cm Ion Engine

range of 1 to 7.9 eV and an electron density range of 4e10 to 1e13 cm-3, throughout the discharge chamber, consistent with the results in the literature [1-2]. Plasma potential estimates, computed from the first derivative of the probe characteristic, indicate potential from 0.5V to 11V above the discharge voltage along the thruster centerline. These values are believed to be excessively high due to poor saturation of the probe IV characteristic in the low-density regions of the plasma. Relative neutral density profiles are also obtained with a fiber optic probe sampling photon flux from the 823.1 nm excited to ground state transition. Plasma parameter measurements and neutral density profiles will be presented as a function of probe location and engine discharge conditions. A discussion of the measured electron energy distribution function will also be presented, with regards to variation from pure maxwellian. It has been found that there is a distinct primary population along the thruster centerline, which causes estimates of electron temperature, electron density, and plasma potential, to err on the high side, due this energetic population. Computation of the energy distribution function of the plasma clearly indicates the presence of primaries, whose presence become less obvious with radial distance from the main discharge plume.

Growth of redeposited carbon and its impact on isotope retention properties on tungsten in a high flux deuterium plasma

Abstract Experiments were performed in a linear magnetized plasma facility (PISCES-B) to simulate carbon re-deposition in the divertor of a fusion reactor such as the International Thermonuclear Experimental Reactor (ITER). The average ion energies are about 100 eV and the ion flux is 2xa0×xa010 22 m −2 s −1 . Tungsten discs and foils were exposed to deuterium plasma for a period of 45–120 min at various substrate temperatures. Carbon impurities were introduced either using graphite sample holder or downstream CD 4 puffing near the targets. In-situ XPS measurements showed a shifting of the binding energy of the carbon in the interlayer between the carbon film and the tungsten surface. Based on AES depth profile, the ratio of tungsten to carbon in the interlayer is about 1.9xa0:xa01. Scanning electron microscopy of plasma-exposed tungsten revealed bubble and pits formation on uncontaminated surfaces. Raman measurements on deposited carbon films were also performed. Deuterium retention measurements were done using Thermal Desorption Spectrometry (TDS). The dominant factor that influences the hydrogen isotope retention is substrate temperature. Measurements indicated a transition from D 2 dominant retention at low temperature exposure to D dominant retention at high temperature exposure. Carbon-contaminated tungsten substrates also showed similar, but a moderate, transition. Total deuterium retention decreases as the exposed temperature increases. The threshold of the carbon impurity concentration in the plasma, under which carbon starts to be deposited on the tungsten surface, is about 0.75% at 850 K. Keeping the impurity concentration at 1%, the temperature threshold is about 750 K.

Molybdenum and Carbon Cluster Angular Sputtering Distributions Under Low Energy Xenon Ion Bombardment

Molybdenum and carbon cluster (C 2 and C 3) angular sputtering distributions are measured during xenon ion bombardment from a plasma, with incident ion energy EXe ranging between 50 and 225 eV. A quadrupole mass spectrometer (QMS) is used to detect the fraction of sputtered neutrals that is ionized in the plasma, and to obtain the angular distribution by changing the angle between the target and the QMS aperture. The angular sputteri ng distribution for molybdenum presents a maximum at 60°, and this maximum becomes less pronounced as the incident ion energy increases. The dependence of the total sputtering yield on incident ion energy is in good agreement with previous experiments. The re is a large increase of about two orders of magnitude in the sputtering yield from EXe = 50 to 125 eV, and a more moderate increase for higher energies. Sputtered C 2 and C 3 clusters exhibit a similar angular sputtering distribution with a maximum at appr oximately 45 -60°; however, this maximum becomes more pronounced for higher incident energies, in contrast to the molybdenum case. The angular distribution of the sputtered clusters depends on the energy with which they are ejected. The low energy populatio n of sputtered particles has a broad maximum at 45°, while the high energy population has a sharp maximum at 60°. The cluster sputtering yield monotonically increases by less than one order of magnitude from EXe = 50 to 225 eV for all measured sputtering a ngles except for normal sputtering, which has a maximum yield at EXe� 100 eV.

DiMES contributions to PMI understanding

The divertor materials evaluation system (DiMES) program at General Atomics has been using a sample changer mechanism to expose different plasma-facing materials to the lower divertor of DIII-D to study integrated plasma materials interaction effects in a tokamak and to benchmark modeling codes. We found that for carbon divertor plates, a detached plasma eliminates net erosion in DIII-D. A spectroscopic study of chemical sputtering indicated the potential improvement of erosion arising from the aging of the first wall material. We also found the importance of the carbon source from the first wall of DIII-D. When lithium was exposed at the divertor we found significant and complicated MHD interactions between the scrape-off layer current in a tokamak and the conducting liquid. This paper is a report on what we have learned and what we plan to do in response to the needs for the plasma facing components (PFC) design for advanced tokamak machines like ITER. Our plan for the study of liquid surface interaction with the plasma will also be presented.

Quantifying the mechanical effects of He, W and He + W ion irradiation on tungsten with spherical nanoindentation

Recent advances in spherical nanoindentation protocols have proven very useful for capturing the grain-scale mechanical response of different metals. This is achieved by converting the load–displacement response into an effective indentation stress–strain response which reveals latent information such as the elastic–plastic transition or indentation yield strength and work-hardening behavior and subsequently correlating the response with the material structure (e.g., crystal orientation) at the indentation site. Using these protocols, we systematically study and quantify the microscale mechanical effects of He, W, and Hexa0+xa0W ion irradiation on commercially pure, polycrystalline tungsten. The indentation stress–strain response is correlated with the crystal orientation from electron backscatter diffraction, the defect structure from transmission electron microscopy micrographs, and the stopping range of ions in matter calculations of displacement damage and He concentration. He-implanted grains show a much higher indentation yield strength and saturation stress compared to W-ion-irradiated grains for the same displacement damage. There is also good agreement between the dispersed barrier hardening model with a barrier strength of 0.5–0.8 and void models (Bacon–Kochs–Scattergood and Osetsky–Bacon models) with the experimentally observed changes in indentation strength due to the presence of He bubbles. This finding indicates that a high density (~xa09xa0×xa01023xa0m−3) and concentration (~xa01.5 at.%) of small (~xa01xa0nm diameter) He bubbles can be moderate to strong barriers to dislocation slip in tungsten.

Liquid lithium wall experiments in CDX-U

The concept of a flowing lithium first wall for a fusion reactor may lead to a significant advance in reactor design, since it could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls. Sputtering and erosion tests are currently underway in the PISCES device at the University of California at San Diego (UCSD). To complement this effort, plasma interaction questions in a toroidal plasma geometry will be addressed by a proposed new ground breaking experiment in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The CDX-U plasma is intensely heated and well diagnosed, and an extensive liquid lithium plasma-facing surface will be used for the first time with a toroidal plasma. Since CDX-U is a small ST, only /spl ap/1 liter or less of lithium is required to produce a toroidal liquid lithium limiter target, leading to a quick and cost-effective experiment.