Carbon fiber composites application in ITER plasma facing components

NASA Astrophysics Data System (ADS)

Barabash, V.; Akiba, M.; Bonal, J. P.; Federici, G.; Matera, R.; Nakamura, K.; Pacher, H. D.; Rödig, M.; Vieider, G.; Wu, C. H.

1998-10-01

Carbon Fiber Composites (CFCs) are one of the candidate armour materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R&D needs are critically discussed.

Plasma-wall interaction studies within the EUROfusion consortium: progress on plasma-facing components development and qualification

NASA Astrophysics Data System (ADS)

Brezinsek, S.; Coenen, J. W.; Schwarz-Selinger, T.; Schmid, K.; Kirschner, A.; Hakola, A.; Tabares, F. L.; van der Meiden, H. J.; Mayoral, M.-L.; Reinhart, M.; Tsitrone, E.; Ahlgren, T.; Aints, M.; Airila, M.; Almaviva, S.; Alves, E.; Angot, T.; Anita, V.; Arredondo Parra, R.; Aumayr, F.; Balden, M.; Bauer, J.; Ben Yaala, M.; Berger, B. M.; Bisson, R.; Björkas, C.; Bogdanovic Radovic, I.; Borodin, D.; Bucalossi, J.; Butikova, J.; Butoi, B.; Čadež, I.; Caniello, R.; Caneve, L.; Cartry, G.; Catarino, N.; Čekada, M.; Ciraolo, G.; Ciupinski, L.; Colao, F.; Corre, Y.; Costin, C.; Craciunescu, T.; Cremona, A.; De Angeli, M.; de Castro, A.; Dejarnac, R.; Dellasega, D.; Dinca, P.; Dittmar, T.; Dobrea, C.; Hansen, P.; Drenik, A.; Eich, T.; Elgeti, S.; Falie, D.; Fedorczak, N.; Ferro, Y.; Fornal, T.; Fortuna-Zalesna, E.; Gao, L.; Gasior, P.; Gherendi, M.; Ghezzi, F.; Gosar, Ž.; Greuner, H.; Grigore, E.; Grisolia, C.; Groth, M.; Gruca, M.; Grzonka, J.; Gunn, J. P.; Hassouni, K.; Heinola, K.; Höschen, T.; Huber, S.; Jacob, W.; Jepu, I.; Jiang, X.; Jogi, I.; Kaiser, A.; Karhunen, J.; Kelemen, M.; Köppen, M.; Koslowski, H. R.; Kreter, A.; Kubkowska, M.; Laan, M.; Laguardia, L.; Lahtinen, A.; Lasa, A.; Lazic, V.; Lemahieu, N.; Likonen, J.; Linke, J.; Litnovsky, A.; Linsmeier, Ch.; Loewenhoff, T.; Lungu, C.; Lungu, M.; Maddaluno, G.; Maier, H.; Makkonen, T.; Manhard, A.; Marandet, Y.; Markelj, S.; Marot, L.; Martin, C.; Martin-Rojo, A. B.; Martynova, Y.; Mateus, R.; Matveev, D.; Mayer, M.; Meisl, G.; Mellet, N.; Michau, A.; Miettunen, J.; Möller, S.; Morgan, T. W.; Mougenot, J.; Mozetič, M.; Nemanič, V.; Neu, R.; Nordlund, K.; Oberkofler, M.; Oyarzabal, E.; Panjan, M.; Pardanaud, C.; Paris, P.; Passoni, M.; Pegourie, B.; Pelicon, P.; Petersson, P.; Piip, K.; Pintsuk, G.; Pompilian, G. O.; Popa, G.; Porosnicu, C.; Primc, G.; Probst, M.; Räisänen, J.; Rasinski, M.; Ratynskaia, S.; Reiser, D.; Ricci, D.; Richou, M.; Riesch, J.; Riva, G.; Rosinski, M.; Roubin, P.; Rubel, M.; Ruset, C.; Safi, E.; Sergienko, G.; Siketic, Z.; Sima, A.; Spilker, B.; Stadlmayr, R.; Steudel, I.; Ström, P.; Tadic, T.; Tafalla, D.; Tale, I.; Terentyev, D.; Terra, A.; Tiron, V.; Tiseanu, I.; Tolias, P.; Tskhakaya, D.; Uccello, A.; Unterberg, B.; Uytdenhoven, I.; Vassallo, E.; Vavpetič, P.; Veis, P.; Velicu, I. L.; Vernimmen, J. W. M.; Voitkans, A.; von Toussaint, U.; Weckmann, A.; Wirtz, M.; Založnik, A.; Zaplotnik, R.; PFC contributors, WP

2017-11-01

The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, and by modelling codes that simulate edge-plasma conditions and the plasma-material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.

Evaluation of runaway-electron effects on plasma-facing components for NET

NASA Astrophysics Data System (ADS)

Bolt, H.; Calén, H.

1991-03-01

Runaway electrons which are generated during disruptions can cause serious damage to plasma facing components in a next generation device like NET. A study was performed to quantify the response of NET plasma facing components to runaway-electron impact. For the determination of the energy deposition in the component materials Monte Carlo computations were performed. Since the subsurface metal structures can be strongly heated under runaway-electron impact from the computed results damage threshold values for the thermal excursions were derived. These damage thresholds are strongly dependent on the materials selection and the component design. For a carbonmolybdenum divertor with 10 and 20 mm carbon armour thickness and 1 degree electron incidence the damage thresholds are 100 MJ/m 2 and 220 MJ/m 2. The thresholds for a carbon-copper divertor under the same conditions are about 50% lower. On the first wall damage is anticipated for energy depositions above 180 MJ/m 2.

Plasma-Facing Component and Materials Testing for the NSTX-U

NASA Astrophysics Data System (ADS)

Jaworski, Michael; Brooks, A.; Gerhardt, S.; Loesser, D.; Mardenfeld, M.; Menard, J.; Gray, T.; Reinke, M.

2017-10-01

The NSTX-U Recovery Project is developing plasma-facing components for use in the divertor of NSTX-U. The extreme conditions of the NSTX-U divertor make it possible to stress even graphite surfaces to the material limits leading to the possibility of component failures. In addition, the complex, mixed-material environment of the NSTX-U due to the use of boron and lithium wall conditioning techniques creates significant uncertainties in the monitoring of the PFCs. A testing program has been developed to inform on the material and design limitations of the NSTX-U high-heat flux components. These tests include high-heat flux testing in electron beam facilities as well as plasma-based testing. The NSTX-U components could experience perpendicular heat fluxes as high as 45 MW/m2. Parallel heat fluxes onto leading edges could reach 475 MW/m2. The testing program and material survey plan will be presented. Work supported by DOE contract DE-AC02-09CH11466 and DE-AC05-00OR22725.

Graphene as a Coating for Plasma Facing Components

NASA Astrophysics Data System (ADS)

Navarro, Marcos; Zamiri, Marziyeh; Kulcinski, Gerald; Lagally, Max; Santarius, John

2017-10-01

This research explores the protection by graphene of plasma facing materials bombarded with energetic ions of helium. Few studies have shown that graphene can act as a protective layer against sputtering due to energetic ions. In the presence of such irradiation, plasma facing components (PFC's) tend to develop surface morphologies that lead to the sputtering of wall material, potentially diminishing the lifetime of the PFC's and plasma performance. Since plasmas have broad applications and the quality of transferred and grown graphene is different, we have used a chemical vapor deposition method to grow on other substrates. We have also shown that graphene can reduce changes on surface morphology due to energetic helium. After irradiation, in the case of graphene-covered tungsten, our results show that, compared to the uncovered W, graphene suppresses these morphologies that form on the surface of hot W. Using Raman spectroscopy as a diagnostic, the graphene coating shows little sign of damage after being irradiated, indicating that there is little to no sputtering of carbon impurities from the surface. We have determined that the mass losses in W have been reduced significantly, which may lead to an improved plasma performance and longer PFC lifetimes. Supported by DHS Project 2015-DN-077-ARI095 and the Grainger Foundation.

The role and application of ion beam analysis for studies of plasma-facing components in controlled fusion devices

NASA Astrophysics Data System (ADS)

Rubel, Marek; Petersson, Per; Alves, Eduardo; Brezinsek, Sebastijan; Coad, Joseph Paul; Heinola, Kalle; Mayer, Matej; Widdowson, Anna

2016-03-01

First wall materials in controlled fusion devices undergo serious modification by several physical and chemical processes arising from plasma-wall interactions. Detailed information is required for the assessment of material lifetime and accumulation of hydrogen isotopes in wall materials. The intention of this work is to give a concise overview of key issues in the characterization of plasma-facing materials and components in tokamaks, especially in JET with an ITER-Like Wall. IBA techniques play a particularly prominent role here because of their isotope selectivity in the low-Z range (1-10), high sensitivity and combination of several methods in a single run. The role of 3He-based NRA, RBS (standard and micro-size beam) and HIERDA in fuel retention and material migration studies is presented. The use of tracer techniques with rare isotopes (e.g. 15N) or marker layers on wall diagnostic components is described. Special instrumentation, development of equipment to enhance research capabilities and issues in handling of contaminated materials are addressed.

High-Z plasma facing components in fusion devices: boundary conditions and operational experiences

NASA Astrophysics Data System (ADS)

Neu, R.

2006-04-01

In present day fusion devices optimization of the performance and experimental freedom motivates the use of low-Z plasma facing materials (PFMs). However, in a future fusion reactor, for economic reasons, a sufficient lifetime of the first wall components is essential. Additionally, tritium retention has to be small to meet safety requirements. Tungsten appears to be the most realistic material choice for reactor plasma facing components (PFCs) because it exhibits the lowest erosion. But besides this there are a lot of criteria which have to be fulfilled simultaneously in a reactor. Results from present day devices and from laboratory experiments confirm the advantages of high-Z PFMs but also point to operational restrictions, when using them as PFCs. These are associated with the central impurity concentration, which is determined by the sputtering yield, the penetration of the impurities and their transport within the confined plasma. The restrictions could exclude successful operation of a reactor, but concomitantly there exist remedies to ameliorate their impact. Obviously some price has to be paid in terms of reduced performance but lacking of materials or concepts which could substitute high-Z PFCs, emphasis has to be put on the development and optimization of reactor-relevant scenarios which incorporate the experiences and measures.

Analysis of the interaction of deuterium plasmas with tungsten in the Fuego-Nuevo II device

NASA Astrophysics Data System (ADS)

Ramos, Gonzalo; Castillo, Fermín; Nieto, Martín; Martínez, Marco; Rangel, José; Herrera-Velázquez, Julio

2012-10-01

Tungsten is one of the main candidate materials for plasma-facing components in future fusion power plants. The Fuego-Nuevo II, a plasma focus device, which can produce dense magnetized helium and deuterium plasmas, has been adapted to address plasma-facing materials questions. In this paper we present results of tungsten targets exposed to deuterium plasmas in the Fuego Nuevo II device, using different experimental conditions. The plasma generated and accelerated in the coaxial gun is expected to have, before the pinch, energies of the order of hundreds eV and velocities of the order of 40,000 m s-1. At the pinch, the ions are reported to have energies of the order of 1.5 keV at most. The samples, analysed with a scanning electron microscope (SEM) in cross section show a damage profile to depths of the order of 580 nm, which are larger than those expected for ions with 1.5 keV, and may be evidence of ion acceleration. An analysis with the SRIM (Stopping Range of Ions in Matter) package calculations is shown.

Liquid Metals as Plasma-facing Materials for Fusion Energy Systems: From Atoms to Tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stone, Howard A.; Koel, Bruce E.; Bernasek, Steven L.

The objective of our studies was to advance our fundamental understanding of liquid metals as plasma-facing materials for fusion energy systems, with a broad scope: from atoms to tokamaks. The flow of liquid metals offers solutions to significant problems of the plasma-facing materials for fusion energy systems. Candidate metals include lithium, tin, gallium, and their eutectic combinations. However, such liquid metal solutions can only be designed efficiently if a range of scientific and engineering issues are resolved that require advances in fundamental fluid dynamics, materials science and surface science. In our research we investigated a range of significant and timelymore » problems relevant to current and proposed engineering designs for fusion reactors, including high-heat flux configurations that are being considered by leading fusion energy groups world-wide. Using experimental and theoretical tools spanning atomistic to continuum descriptions of liquid metals, and bridging surface chemistry, wetting/dewetting and flow, our research has advanced the science and engineering of fusion energy materials and systems. Specifically, we developed a combined experimental and theoretical program to investigate flows of liquid metals in fusion-relevant geometries, including equilibrium and stability of thin-film flows, e.g. wetting and dewetting, effects of electromagnetic and thermocapillary fields on liquid metal thin-film flows, and how chemical interactions and the properties of the surface are influenced by impurities and in turn affect the surface wetting characteristics, the surface tension, and its gradients. Because high-heat flux configurations produce evaporation and sputtering, which forces rearrangement of the liquid, and any dewetting exposes the substrate to damage from the plasma, our studies addressed such evaporatively driven liquid flows and measured and simulated properties of the different bulk phases and material interfaces. The range of our studies included (i) quantum mechanical calculations that allow inclusion of many thousands of atoms for the characterization of the interface of liquid metals exposed to continuous bombardment by deuterium and tritium as expected in fusion, (ii) molecular dynamics studies of the phase behavior of liquid metals, which (a) utilize thermodynamic properties computed using our quantum mechanical calculations and (b) establish material and wetting properties of the liquid metals, including relevant eutectics, (iii) experimental investigations of the surface science of liquid metals, interacting both with the solid substrate as well as gaseous species, and (iv) fluid dynamical studies that incorporate the material and surface science results of (ii) and (iii) in order to characterize flow in capillary porous materials and the thin-film flow along curved boundaries, both of which are potentially major components of plasma-facing materials. The outcome of these integrated studies was new understanding that enables developing design rules useful for future developments of the plasma-facing components critical to the success of fusion energy systems.« less

Using the tritium plasma experiment to evaluate ITER PFC safety

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Anderl, Robert A.; Bartlit, John R.; Causey, Rion A.; Haines, John R.

1993-06-01

The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore and is being moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capabilty of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 × 1023 ions/m2.s and a plasma temperature of about 15 eV using a plasma that includes tritium. An experimental program has been initiated using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. An industrial consortium led by McDonnell Douglas will design and fabricate the test fixtures.

Plasma-surface interaction in the Be/W environment: Conclusions drawn from the JET-ILW for ITER

NASA Astrophysics Data System (ADS)

Brezinsek, S.; JET-EFDA contributors

2015-08-01

The JET ITER-Like Wall experiment (JET-ILW) provides an ideal test bed to investigate plasma-surface interaction (PSI) and plasma operation with the ITER plasma-facing material selection employing beryllium in the main chamber and tungsten in the divertor. The main PSI processes: material erosion and migration, (b) fuel recycling and retention, (c) impurity concentration and radiation have be1en studied and compared between JET-C and JET-ILW. The current physics understanding of these key processes in the JET-ILW revealed that both interpretation of previously obtained carbon results (JET-C) and predictions to ITER need to be revisited. The impact of the first-wall material on the plasma was underestimated. Main observations are: (a) low primary erosion source in H-mode plasmas and reduction of the material migration from the main chamber to the divertor (factor 7) as well as within the divertor from plasma-facing to remote areas (factor 30 - 50). The energetic threshold for beryllium sputtering minimises the primary erosion source and inhibits multi-step re-erosion in the divertor. The physical sputtering yield of tungsten is low as 10-5 and determined by beryllium ions. (b) Reduction of the long-term fuel retention (factor 10 - 20) in JET-ILW with respect to JET-C. The remaining retention is caused by implantation and co-deposition with beryllium and residual impurities. Outgassing has gained importance and impacts on the recycling properties of beryllium and tungsten. (c) The low effective plasma charge (Zeff = 1.2) and low radiation capability of beryllium reveal the bare deuterium plasma physics. Moderate nitrogen seeding, reaching Zeff = 1.6 , restores in particular the confinement and the L-H threshold behaviour. ITER-compatible divertor conditions with stable semi-detachment were obtained owing to a higher density limit with ILW. Overall JET demonstrated successful plasma operation in the Be/W material combination and confirms its advantageous PSI behaviour and gives strong support to the ITER material selection.

Plasma–wall interaction studies within the EUROfusion consortium: progress on plasma-facing components development and qualification

DOE PAGES

Brezinsek, S.; Coenen, J. W.; Schwarz-Selinger, T.; ...

2017-06-14

The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, andmore » by modelling codes that simulate edge-plasma conditions and the plasma–material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.« less

Plasma–wall interaction studies within the EUROfusion consortium: progress on plasma-facing components development and qualification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brezinsek, S.; Coenen, J. W.; Schwarz-Selinger, T.

The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, andmore » by modelling codes that simulate edge-plasma conditions and the plasma–material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.« less

Protection of tokamak plasma facing components by a capillary porous system with lithium

NASA Astrophysics Data System (ADS)

Lyublinski, I.; Vertkov, A.; Mirnov, S.; Lazarev, V.

2015-08-01

Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.

Investigation of tin-lithium eutectic as a liquid plasma facing material

NASA Astrophysics Data System (ADS)

Ruzic, David; Szott, Matthew; Christenson, Michael; Shchelkanov, Ivan; Kalathiparambil, Kishor Kumar

2016-10-01

Innovative materials and techniques need to be utilized to address the high heat and particle flux incident on plasma facing components in fusion reactors. A liquid metal diverter module developed at UIUC with self circulating lithium has been successfully demonstrated to be capable of handling the relevant heat flux in plasma gun based tests and on operational tokamaks. The proper geometry of the liquid lithium trenches to minimize droplet ejection during transient plasma events have also been identified. Although lithium has proven to be effective in improved plasma performance and contributes to other advantageous factors like reduction in the fuel recycling, impurity gettering and, owing to the low Z, a significantly reduced impact on plasma as compared to the solid wall materials, it still poses several drawbacks related to its high reactivity and high vapor pressure at the relevant tokamak wall temperatures. The evaporation properties of a new eutectic mixture of tin and lithium (20% Sn) shows that lithium segregates to the surface at melting temperatures and hence is an effective replacement for pure lithium. Also, the vapor from the eutectic is dominated by lithium, minimizing the entry of high Z Sn into the plasma. At UIUC experiments for the synthesis and characterization of the eutectic - measurement of the critical wetting parameters and Seebeck coefficients with respect to the trench materials have been performed to ensure lithium wetting and flow in the trenches. The results will be presented. DOE project DEFG02- 99ER54515.

Armour Materials for the ITER Plasma Facing Components

NASA Astrophysics Data System (ADS)

Barabash, V.; Federici, G.; Matera, R.; Raffray, A. R.; ITER Home Teams,

The selection of the armour materials for the Plasma Facing Components (PFCs) of the International Thermonuclear Experimental Reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-à-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R&D.

Addressing the challenges of plasma-surface interactions in NSTX-U*

DOE PAGES

Kaita, Robert; Abrams, Tyler; Jaworski, Michael; ...

2015-04-01

The importance of conditioning plasma-facing components (PFCs) has long been recognized as a critical element in obtaining high-performance plasmas in magnetic confinement devices. Lithium coatings, for example, have been used for decades for conditioning PFCs. Since the initial studies on the Tokamak Fusion Test Reactor, experiments on devices with different aspect ratios and magnetic geometries like the National Spherical Torus Experiment (NSTX) continue to show the relationship between lithium PFCs and good confinement and stability. While such results are promising, their empirical nature do not reflect the detailed relationship between PFCs and the dynamic conditions that occur in the tokamakmore » environment. A first step developing an understanding such complexity will be taken in the upgrade to NSTX (NSTX-U) that is nearing completion. New measurement capabilities include the Materials Analysis and Particle Probe (MAPP) for in situ surface analysis of samples exposed to tokamak plasmas. The OEDGE suite of codes, for example, will provide a new way to model the underlying mechanisms for such material migration in NSTX-U. This will lead to a better understanding of how plasma-facing surfaces evolve during a shot, and how the composition of the plasma facing surface influences the discharge performance we observe. This paper will provide an overview of these capabilities, and highlight their importance for NSTX-U plans to transition from carbon to high-Z PFCs.« less

Long Duration Exposure Facility (LDEF) preliminary findings: LEO space effects on the space plasma-voltage drainage experiment

NASA Technical Reports Server (NTRS)

Blakkolb, Brian K.; Yaung, James Y.; Henderson, Kelly A.; Taylor, William W.; Ryan, Lorraine E.

1992-01-01

The Space Plasma-High Voltage Drainage Experiment (SP-HVDE) provided a unique opportunity to study long term space environmental effects on materials because it was comprised of two identical experimental trays; one tray located on the ram facing side (D-10), and the other on the wake facing side (B-4) of the LDEF. This configuration allows for the comparison of identical materials exposed to two distinctly different environments. The purpose of this work is to document an assessment of the effects of five and three quarters years of low Earth orbital space exposure on materials comprising the SP-HVDE (experiment no. A0054). The findings of the materials investigation reported focus on atomic oxygen effects, micrometeor and debris impact site documentation, thermal property measurements, and environmentally induced contamination.

Development of Advanced Ill-Nitride Materials

DTIC Science & Technology

2008-09-24

have continued to work on InN and related materials. During the last year, we have completed many of our basic materials studies and extended our...conductivity of InN films The origin of bulk electrons in In-face InN has been studied by considering the effects of both unintentionally incorporated... studied in In- and N-face InN films grown on GaN by plasma-assisted molecular beam epitaxy. The TD densities were determined by non-destructive x-ray

Oscillatory vapour shielding of liquid metal walls in nuclear fusion devices.

PubMed

van Eden, G G; Kvon, V; van de Sanden, M C M; Morgan, T W

2017-08-04

Providing an efficacious plasma facing surface between the extreme plasma heat exhaust and the structural materials of nuclear fusion devices is a major challenge on the road to electricity production by fusion power plants. The performance of solid plasma facing surfaces may become critically reduced over time due to progressing damage accumulation. Liquid metals, however, are now gaining interest in solving the challenge of extreme heat flux hitting the reactor walls. A key advantage of liquid metals is the use of vapour shielding to reduce the plasma exhaust. Here we demonstrate that this phenomenon is oscillatory by nature. The dynamics of a Sn vapour cloud are investigated by exposing liquid Sn targets to H and He plasmas at heat fluxes greater than 5 MW m -2 . The observations indicate the presence of a dynamic equilibrium between the plasma and liquid target ruled by recombinatory processes in the plasma, leading to an approximately stable surface temperature.Vapour shielding is one of the interesting mechanisms for reducing the heat load to plasma facing components in fusion reactors. Here the authors report on the observation of a dynamic equilibrium between the plasma and the divertor liquid Sn surface leading to an overall stable surface temperature.

FOREWORD: 13th International Workshop on Plasma-Facing Materials and Components for Fusion Applications/1st International Conference on Fusion Energy Materials Science 13th International Workshop on Plasma-Facing Materials and Components for Fusion Applications/1st International Conference on Fusion Energy Materials Science

NASA Astrophysics Data System (ADS)

Jacob, Wolfgang; Linsmeier, Christian; Rubel, Marek

2011-12-01

The 13th International Workshop on Plasma-Facing Materials and Components (PFMC-13) jointly organized with the 1st International Conference on Fusion Energy Materials Science (FEMaS-1) was held in Rosenheim (Germany) on 9-13 May 2011. PFMC-13 is a successor of the International Workshop on Carbon Materials for Fusion Applications series. Between 1985 and 2003 ten 'Carbon Workshops' were organized in Jülich, Stockholm and Hohenkammer. Then it was time for a change and redefinition of the scope of the symposium to reflect the new requirements of ITER and the ongoing evolution in the field. Under the new name (PFMC-11), the workshop was first organized in 2006 in Greifswald, Germany and PFMC-12 took place in Jülich in 2009. Initially starting in 1985 with about 40 participants as a 1.5 day workshop, the event has continuously grown to about 220 participants at PFMC-12. Due to the joint organization with FEMaS-1, PFMC-13 set a new record with more than 280 participants. The European project Fusion Energy Materials Science, FEMaS, coordinated by the Max-Planck-Institut für Plasmaphysik (IPP), organizes and stimulates cooperative research activities which involve large-scale research facilities as well as other top-level materials characterization laboratories. Five different fields are addressed: benchmarking experiments for radiation damage modelling, the application of micro-mechanical characterization methods, synchrotron and neutron radiation-based techniques and advanced nanoscopic analysis based on transmission electron microscopy. All these fields need to be exploited further by the fusion materials community for timely materials solutions for a DEMO reactor. In order to integrate these materials research fields, FEMaS acted as a co-organizer for the 2011 workshop and successfully introduced a number of participants from research labs and universities into the PFMC community. Plasma-facing materials experience particularly hostile conditions as they are subjected to extremely high heat loads and very high particle and neutron fluxes. They must have high thermal conductivity for efficient heat transport, high cohesive energy for low erosion by particle bombardment and low atomic number to minimize plasma cooling. These contradictory requirements make the development of plasma-facing materials one of the greatest challenges ever faced by materials scientists. The erosion of plasma-facing materials is one of the main factors influencing the operational schedule of experimental fusion reactors and future power plants. A number of materials selected for current designs cannot withstand the presently foreseen plasma scenarios of a power plant for a commercially viable period of time. Therefore, further coordinated development of plasma scenarios and materials is essential for the realization of fusion as an energy source. The design and development of plasma-facing materials requires a detailed understanding of the processes that occur when a material surface is bombarded with an intense flux of heat, particles and neutrons simultaneously. These materials-related topics are the focus of this series of workshops which has established itself as a discussion forum for experts from research institutions and industry dealing with materials for plasma-facing components in present and future thermonuclear fusion devices. During the joint conference PFMC-13/FEMaS-1 recent developments and research results in the following fields were addressed: carbon, beryllium, and tungsten based materials mixed materials erosion and redeposition high heat flux component development benchmarking of radiation damage modelling synchrotron and neutron based characterization techniques application of advanced transmission electron microscopy and micro-/nano-mechanical testing. With the approaching technical realization of ITER, the ITER-related PFMC topics are naturally the main focus of research. In this respect the start of the ITER-like wall experiment at JET is of paramount importance for our community and several presentations were devoted to this topic. The start of the experimental campaign shortly after PFMC-13/FEMaS-1 will most probably bring about many exciting new results and leaves us eagerly awaiting the next PFMC conference. Several other topics which are of significant relevance for the preparation of ITER were addressed. Among them were dust detection and analysis which is a safety concern and the behaviour of beryllium. Due to the toxicity of beryllium dust, great care has to be taken in the handling of beryllium-containing samples and, as a consequence, only a very limited number of places are available worldwide where such samples can be prepared and investigated. For a solid database and a sound understanding of beryllium and beryllium-containing mixed materials much more effort is necessary in the near future. Naturally, traditional PFMC topics such as first-wall lifetime, testing and characterization of plasma-facing components and hydrogen inventory had their appropriate share of the programme. Not to forget carbon, the nucleating material for this workshop series. Although it will, according to present planning, play only a minor role towards the realization of a DEMO reactor, it is still of importance for current machines and was covered in a large number of poster contributions. Topics receiving continuously increasing attention are those related to devices beyond ITER. Such topics are the development of advanced materials, their behaviour under high heat loads and, in particular, the consequences of neutron damage. The issue which was treated in quite a number of contributions was the simulation of neutron damage by implantation of heavy ions and its influence on hydrogen retention. This is presumably a topic which will receive continuous attention in the years to come. As a consequence of the joint organization with the FEMaS project, several presentations addressed advanced characterization techniques. Remarkable examples of 3D tomography images of plasma-facing components using x-ray- or neutron-based techniques were shown. Such methods allow non-destructive and element-resolved analyses of buried interfaces and are therefore a very promising tool for future investigations of plasma-facing components. It would be desirable that many colleagues of the FEMaS community who attended PFMC-13/FEMaS-1 for the first time would also participate in future events of this series. Thirty five invited lectures and oral contributions and 192 posters were presented by participants coming from research laboratories and industrial companies. Two hundered and eighty two researchers from 27 countries from all over the world participated in the lively and intense exchange of results and new ideas. An additional objective of the series of PFMC workshops was and is to encourage the participation of young talented scientists and to spark their interest in this field. For that reason, the workshop started on its first day with a tutorial session. Experts in their respective fields presented in total eight introductory lectures ranging from the basics of plasma-wall interactions to the engineering of plasma-facing components for ITER. Although originally intended for students and newcomers to the field, these tutorial lectures also enjoy great popularity among senior scientist and are in the meantime an indispensable ingredient and a trademark of this workshop series. The event was organized by the IPP, Garching and received substantial financial support from the European Commission through FEMaS. We are very thankful to the staff of IPP who helped with the organization. Our most cordial thanks and gratitude go to Mrs Christina Stahlberg and Mrs Jutta Koser for their help in the organization and at the front desk. Our most sincere words of appreciation go to our colleague Elmar Neitzert who was in charge of administrative organization. The present proceedings of PFMC-13/FEMaS-1 contain in total 83 peer-reviewed publications covering the contents of most of the oral presentations and of a number of poster contributions which were pre-selected by the programme committee. The papers reflect the development and actual status of the field. We thank all participants for their contributions and we particularly thank the referees for their systematic and diligent reviews of the submitted articles. It is due to their commitment and punctual return of reviews that the proceedings can appear in this relative short time after the meeting. In a meeting of the programme committee during the conferences a few changes in the committee composition were decided. Paul Coad retired and has left the programme committee. We cordially thank Paul Coad for his long-time service as a committee member and wish him the very best for the future. We are very happy that Guy Matthews (CCFE, Culham, UK) accepted the invitation to be his successor. Furthermore, to strengthen the international character of the event, it was decided to invite an additional representative from Japan to the programme committee. Noriyasu Ohno from Nagoya University accepted the invitation. To maintain close contact to the FEMaS community the programme committee further decided to invite Christian Linsmeier from IPP, Garching. Another important decision was taken: in view of the size that the event has reached it was decided to change the name from 'workshop' to 'conference'. So the next event in this series will be the PFMC-14 conference. It will be organized by FZ Jülich and most probably take place in spring 2013.

Tungsten-microdiamond composites for plasma facing components

NASA Astrophysics Data System (ADS)

Livramento, V.; Nunes, D.; Correia, J. B.; Carvalho, P. A.; Mardolcar, U.; Mateus, R.; Hanada, K.; Shohoji, N.; Fernandes, H.; Silva, C.; Alves, E.

2011-09-01

Tungsten is considered as one of promising candidate materials for plasma facing component in nuclear fusion reactors due to its resistance to sputtering and high melting point. High thermal conductivity is also a prerequisite for plasma facing components under the unique service environment of fusion reactor characterised by the massive heat load, especially in the divertor area. The feasibility of mechanical alloying of nanodiamond and tungsten, and the consolidation of the composite powders with Spark Plasma Sintering (SPS) was previously demonstrated. In the present research we report on the use of microdiamond instead of nanodiamond in such composites. Microdiamond is more favourable than nanodiamond in view of phonon transport performance leading to better thermal conductivity. However, there is a trade off between densification and thermal conductivity as the SPS temperature increases tungsten carbide formation from microdiamond is accelerated inevitably while the consolidation density would rise.

Coating materials for fusion application in China

NASA Astrophysics Data System (ADS)

Luo, G.-N.; Li, Q.; Liu, M.; Zheng, X. B.; Chen, J. L.; Guo, Q. G.; Liu, X.

2011-10-01

Thick SiC coatings of ˜100 μm on graphite tiles, prepared by chemical vapor infiltration of Si into the tiles and the following reactions between Si and C, are used as plasma facing material (PFM) on HT-7 superconducting tokamak and Experimental Advanced Superconducting Tokamak (EAST). With increase in the heating and driving power in EAST, the present plasma facing component (PFC) of the SiC/C tiles bolted to heat sink will be replaced by W coatings on actively cooled Cu heat sink, prepared by vacuum plasma spraying (VPS) adopting different interlayer. The VPS-W/Cu PFC with built-in cooling channels were prepared and mounted into the HT-7 acting as a movable limiter. Behavior of heat load onto the limiter and the material was studied. The Cu coatings on the Inconel 625 tubes were successfully prepared by high velocity air-fuel (HVAF) thermal spraying, being used as the liquid nitrogen (LN2) shields of the in-vessel cryopump for divertor pumping in EAST.

EU Development of High Heat Flux Components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linke, J.; Lorenzetto, P.; Majerus, P.

2005-04-15

The development of plasma facing components for next step fusion devices in Europe is strongly focused to ITER. Here a wide spectrum of different design options for the divertor target and the first wall have been investigated with tungsten, CFC, and beryllium armor. Electron beam simulation experiments have been used to determine the performance of high heat flux components under ITER specific thermal loads. Beside thermal fatigue loads with power density levels up to 20 MWm{sup -2}, off-normal events are a serious concern for the lifetime of plasma facing components. These phenomena are expected to occur on a time scalemore » of a few milliseconds (plasma disruptions) or several hundred milliseconds (vertical displacement events) and have been identified as a major source for the production of neutron activated metallic or tritium enriched carbon dust which is of serious importance from a safety point of view.The irradiation induced material degradation is another critical concern for future D-T-burning fusion devices. In ITER the integrated neutron fluence to the first wall and the divertor armour will remain in the order of 1 dpa and 0.7 dpa, respectively. This value is low compared to future commercial fusion reactors; nevertheless, a nonnegligible degradation of the materials has been detected, both for mechanical and thermal properties, in particular for the thermal conductivity of carbon based materials. Beside the degradation of individual material properties, the high heat flux performance of actively cooled plasma facing components has been investigated under ITER specific thermal and neutron loads.« less

Failure study of helium-cooled tungsten divertor plasma-facing units tested at DEMO relevant steady-state heat loads

NASA Astrophysics Data System (ADS)

Ritz, G.; Hirai, T.; Norajitra, P.; Reiser, J.; Giniyatulin, R.; Makhankov, A.; Mazul, I.; Pintsuk, G.; Linke, J.

2009-12-01

Tungsten was selected as armor material for the helium-cooled divertor in future DEMO-type fusion reactors and fusion power plants. After realizing the design and testing of them under cyclic thermal loads of up to ~14 MW m-2, the tungsten divertor plasma-facing units were examined by metallography; they revealed failures such as cracks at the thermal loaded and as-machined surfaces, as well as degradation of the brazing layers. Furthermore, in order to optimize the machining processes, the quality of tungsten surfaces prepared by turning, milling and using a diamond cutting wheel were examined. This paper presents a metallographic examination of the tungsten plasma-facing units as well as technical studies and the characterization on machining of tungsten and alternative brazing joints.

Experimental Validation Plan for the Xolotl Plasma-Facing Component Simulator Using Tokamak Sample Exposures

NASA Astrophysics Data System (ADS)

Chan, V. S.; Wong, C. P. C.; McLean, A. G.; Luo, G. N.; Wirth, B. D.

2013-10-01

The Xolotl code under development by PSI-SciDAC will enhance predictive modeling capability of plasma-facing materials under burning plasma conditions. The availability and application of experimental data to compare to code-calculated observables are key requirements to validate the breadth and content of physics included in the model and ultimately gain confidence in its results. A dedicated effort has been in progress to collect and organize a) a database of relevant experiments and their publications as previously carried out at sample exposure facilities in US and Asian tokamaks (e.g., DIII-D DiMES, and EAST MAPES), b) diagnostic and surface analysis capabilities available at each device, and c) requirements for future experiments with code validation in mind. The content of this evolving database will serve as a significant resource for the plasma-material interaction (PMI) community. Work supported in part by the US Department of Energy under GA-DE-SC0008698, DE-AC52-07NA27344 and DE-AC05-00OR22725.

The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions

DOE PAGES

Garrison, L. M.; Zenobia, Samuel J.; Egle, Brian J.; ...

2016-08-01

The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000°C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ionmore » gun can irradiate the samples with ion currents of 20 μA–500 μA; the typical current used is 72 μA, which is an average flux of 9 × 10 14 ions/(cm 2 s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. In conclusion, the MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.« less

The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions.

PubMed

Garrison, L M; Zenobia, S J; Egle, B J; Kulcinski, G L; Santarius, J F

2016-08-01

The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000 °C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ion gun can irradiate the samples with ion currents of 20 μA-500 μA; the typical current used is 72 μA, which is an average flux of 9 × 10(14) ions/(cm(2) s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. The MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.

The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions

NASA Astrophysics Data System (ADS)

Garrison, L. M.; Zenobia, S. J.; Egle, B. J.; Kulcinski, G. L.; Santarius, J. F.

2016-08-01

The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000 °C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ion gun can irradiate the samples with ion currents of 20 μA-500 μA; the typical current used is 72 μA, which is an average flux of 9 × 1014 ions/(cm2 s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. The MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.

Thermal shock tests to qualify different tungsten grades as plasma facing material

NASA Astrophysics Data System (ADS)

Wirtz, M.; Linke, J.; Loewenhoff, Th; Pintsuk, G.; Uytdenhouwen, I.

2016-02-01

The electron beam device JUDITH 1 was used to establish a testing procedure for the qualification of tungsten as plasma facing material. Absorbed power densities of 0.19 and 0.38 GW m-2 for an edge localized mode-like pulse duration of 1 ms were chosen. Furthermore, base temperatures of room temperature, 400 °C and 1000 °C allow investigating the thermal shock performance in the brittle, ductile and high temperature regime. Finally, applying 100 pulses under all mentioned conditions helps qualifying the general damage behaviour while with 1000 pulses for the higher power density the influence of thermal fatigue is addressed. The investigated reference material is a tungsten product produced according to the ITER material specifications. The obtained results provide a general overview of the damage behaviour with quantified damage characteristics and thresholds. In particular, it is shown that the damage strongly depends on the microstructure and related thermo-mechanical properties.

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

NASA Astrophysics Data System (ADS)

Lucia, M.; Kaita, R.; Majeski, R.; Bedoya, F.; Allain, J. P.; Abrams, T.; Bell, R. E.; Boyle, D. P.; Jaworski, M. A.; Schmitt, J. C.

2015-08-01

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 5 h argon GDC or a 50 nm evaporative lithium coating will both significantly improve LTX plasma performance.

High heat flux issues for plasma-facing components in fusion reactors

NASA Astrophysics Data System (ADS)

Watson, Robert D.

1993-02-01

Plasma facing components in tokamak fusion reactors are faced with a number of difficult high heat flux issues. These components include: first wall armor tiles, pumped limiters, diverter plates, rf antennae structure, and diagnostic probes. Peak heat fluxes are 15 - 30 MW/m2 for diverter plates, which will operate for 100 - 1000 seconds in future tokamaks. Disruption heat fluxes can approach 100,000 MW/m2 for 0.1 ms. Diverter plates are water-cooled heat sinks with armor tiles brazed on to the plasma facing side. Heat sink materials include OFHC, GlidcopTM, TZM, Mo-41Re, and niobium alloys. Armor tile materials include: carbon fiber composites, beryllium, silicon carbide, tungsten, and molybdenum. Tile thickness range from 2 - 10 mm, and heat sinks are 1 - 3 mm. A twisted tape insert is used to enhance heat transfer and increase the burnout safety margin from critical heat flux limits to 50 - 60 MW/m2 with water at 10 m/s and 4 MPa. Tests using rastered electron beams have shown thermal fatigue failures from cracks at the brazed interface between tiles and the heat sink after only 1000 cycles at 10 - 15 MW/m2. These fatigue lifetimes need to be increased an order of magnitude to meet future requirements. Other critical issues for plasma facing components include: surface erosion from sputtering and disruption erosion, eddy current forces and runaway electron impact from disruptions, neutron damage, tritium retention and release, remote maintenance of radioactive components, corrosion-erosion, and loss-of-coolant accidents.

RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Part II: Analysis of ITER plasma facing components

NASA Astrophysics Data System (ADS)

Federici, Gianfranco; Raffray, A. René

1997-04-01

The transient thermal model RACLETTE (acronym of Rate Analysis Code for pLasma Energy Transfer Transient Evaluation) described in part I of this paper is applied here to analyse the heat transfer and erosion effects of various slow (100 ms-10 s) high power energy transients on the actively cooled plasma facing components (PFCs) of the International Thermonuclear Experimental Reactor (ITER). These have a strong bearing on the PFC design and need careful analysis. The relevant parameters affecting the heat transfer during the plasma excursions are established. The temperature variation with time and space is evaluated together with the extent of vaporisation and melting (the latter only for metals) for the different candidate armour materials considered for the design (i.e., Be for the primary first wall, Be and CFCs for the limiter, Be, W, and CFCs for the divertor plates) and including for certain cases low-density vapour shielding effects. The critical heat flux, the change of the coolant parameters and the possible severe degradation of the coolant heat removal capability that could result under certain conditions during these transients, for example for the limiter, are also evaluated. Based on the results, the design implications on the heat removal performance and erosion damage of the variuos ITER PFCs are critically discussed and some recommendations are made for the selection of the most adequate protection materials and optimum armour thickness.

High-flux plasma exposure of ultra-fine grain tungsten

DOE PAGES

Kolasinski, R. D.; Buchenauer, D. A.; Doerner, R. P.; ...

2016-05-12

Here we examine the response of an ultra-fine grained (UFG) tungsten material to high-flux deuterium plasma exposure. UFG tungsten has received considerable interest as a possible plasma-facing material in magnetic confinement fusion devices, in large part because of its improved resistance to neutron damage. However, optimization of the material in this manner may lead to trade-offs in other properties. Moreover, we address two aspects of the problem in this work: (a) how high-flux plasmas modify the structure of the exposed surface, and (b) how hydrogen isotopes become trapped within the material. The specific UFG tungsten considered here contains 100 nm-widthmore » Ti dispersoids (1 wt%) that limit the growth of the W grains to a median size of 960 nm. Metal impurities (Fe, Cr) as well as O were identified within the dispersoids; these species were absent from the W matrix. To simulate relevant particle bombardment conditions, we exposed specimens of the W-Ti material to low energy (100 eV), high-flux (> 10 22 m -2 s -1) deuterium plasmas in the PISCES-A facility at the University of California, San Diego. To explore different temperature-dependent trapping mechanisms, we considered a range of exposure temperatures between 200 °C and 500 °C. For comparison, we also exposed reference specimens of conventional powder metallurgy warm-rolled and ITER-grade tungsten at 300 °C. Post-mortem focused ion beam profiling and atomic force microscopy of the UFG tungsten revealed no evidence of near-surface bubbles containing high pressure D2 gas, a common surface degradation mechanism associated with plasma exposure. Thermal desorption spectrometry indicated moderately higher trapping of D in the material compared with the reference specimens, though still within the spread of values for different tungsten grades found in the literature database. Finally, for the criteria considered here, these results do not indicate any significant obstacles to the potential use of UFG tungsten as a plasma-facing material, although further experimental work is needed to assess material response to transient events and high plasma fluence.« less

Sensitivity of WallDYN material migration modeling to uncertainties in mixed-material surface binding energies

DOE PAGES

Nichols, J. H.; Jaworski, M. A.; Schmid, K.

2017-03-09

The WallDYN package has recently been applied to a number of tokamaks to self-consistently model the evolution of mixed-material plasma facing surfaces. A key component of the WallDYN model is the concentration-dependent surface sputtering rate, calculated using SDTRIM.SP. This modeled sputtering rate is strongly influenced by the surface binding energies (SBEs) of the constituent materials, which are well known for pure elements but often are poorly constrained for mixed-materials. This work examines the sensitivity of WallDYN surface evolution calculations to different models for mixed-material SBEs, focusing on the carbon/lithium/oxygen/deuterium system present in NSTX. A realistic plasma background is reconstructed frommore » a high density, H-mode NSTX discharge, featuring an attached outer strike point with local density and temperature of 4 × 10 20 m -3 and 4 eV, respectively. It is found that various mixed-material SBE models lead to significant qualitative and quantitative changes in the surface evolution profile at the outer divertor, with the highest leverage parameter being the C-Li binding model. Uncertainties of order 50%, appearing on time scales relevant to tokamak experiments, highlight the importance of choosing an appropriate mixed-material sputtering representation when modeling the surface evolution of plasma facing components. Lastly, these results are generalized to other fusion-relevant materials with different ranges of SBEs.« less

Sensitivity of WallDYN material migration modeling to uncertainties in mixed-material surface binding energies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nichols, J. H.; Jaworski, M. A.; Schmid, K.

The WallDYN package has recently been applied to a number of tokamaks to self-consistently model the evolution of mixed-material plasma facing surfaces. A key component of the WallDYN model is the concentration-dependent surface sputtering rate, calculated using SDTRIM.SP. This modeled sputtering rate is strongly influenced by the surface binding energies (SBEs) of the constituent materials, which are well known for pure elements but often are poorly constrained for mixed-materials. This work examines the sensitivity of WallDYN surface evolution calculations to different models for mixed-material SBEs, focusing on the carbon/lithium/oxygen/deuterium system present in NSTX. A realistic plasma background is reconstructed frommore » a high density, H-mode NSTX discharge, featuring an attached outer strike point with local density and temperature of 4 × 10 20 m -3 and 4 eV, respectively. It is found that various mixed-material SBE models lead to significant qualitative and quantitative changes in the surface evolution profile at the outer divertor, with the highest leverage parameter being the C-Li binding model. Uncertainties of order 50%, appearing on time scales relevant to tokamak experiments, highlight the importance of choosing an appropriate mixed-material sputtering representation when modeling the surface evolution of plasma facing components. Lastly, these results are generalized to other fusion-relevant materials with different ranges of SBEs.« less

Full toroidal imaging of non-axisymmetric plasma material interaction in the National Spherical Torus Experiment divertor.

PubMed

Scotti, Filippo; Roquemore, A L; Soukhanovskii, V A

2012-10-01

A pair of two dimensional fast cameras with a wide angle view (allowing a full radial and toroidal coverage of the lower divertor) was installed in the National Spherical Torus Experiment in order to monitor non-axisymmetric effects. A custom polar remapping procedure and an absolute photometric calibration enabled the easier visualization and quantitative analysis of non-axisymmetric plasma material interaction (e.g., strike point splitting due to application of 3D fields and effects of toroidally asymmetric plasma facing components).

Preliminary design of laser-induced breakdown spectroscopy for proto-Material Plasma Exposure eXperiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shaw, G., E-mail: shawgc@ornl.gov; University of Tennessee, Knoxville, Tennessee 37996; Martin, M. Z.

2014-11-15

Laser-induced breakdown spectroscopy (LIBS) is a technique for measuring surface matter composition. LIBS is performed by focusing laser radiation onto a target surface, ablating the surface, forming a plasma, and analyzing the light produced. LIBS surface analysis is a possible diagnostic for characterizing plasma-facing materials in ITER. Oak Ridge National Laboratory has enabled the initial installation of a laser-induced breakdown spectroscopy diagnostic on the prototype Material-Plasma Exposure eXperiment (Proto-MPEX), which strives to mimic the conditions found at the surface of the ITER divertor. This paper will discuss the LIBS implementation on Proto-MPEX, preliminary design of the fiber optic LIBS collectionmore » probe, and the expected results.« less

Final case for a stainless steel diagnostic first wall on ITER

NASA Astrophysics Data System (ADS)

Pitts, R. A.; Bazylev, B.; Linke, J.; Landman, I.; Lehnen, M.; Loesser, D.; Loewenhoff, Th.; Merola, M.; Roccella, R.; Saibene, G.; Smith, M.; Udintsev, V. S.

2015-08-01

In 2010 the ITER Organization (IO) proposed to eliminate the beryllium armour on the plasma-facing surface of the diagnostic port plugs and instead to use bare stainless steel (SS), simplifying the design and providing significant cost reduction. Transport simulations at the IO confirmed that charge-exchange sputtering of the SS surfaces would not affect burning plasma operation through core impurity contamination, but a second key issue is the potential melt damage/material loss inflicted by the intense photon radiation flashes expected at the thermal quench of disruptions mitigated by massive gas injection. This paper addresses this second issue through a combination of ITER relevant experimental heat load tests and qualitative theoretical arguments of melt layer stability. It demonstrates that SS can be employed as material for the port plug plasma-facing surface and this has now been adopted into the ITER baseline.

Modeling Electrothermal Plasma with Boundary Layer Effects

NASA Astrophysics Data System (ADS)

AlMousa, Nouf Mousa A.

Electrothermal plasma sources produce high-density (1023-10 28 /m3) and high temperature (1-5 eV) plasmas that are of interest for a variety of applications such as hypervelocity launch devices, fusion reactor pellet injectors, and pulsed thrusters for small satellites. Also, the high heat flux (up to 100 GW/m2) and high pressure (100s MPa) of electrothermal (ET) plasmas allow for the use of such facilities as a source of high heat flux to simulate off-normal events in Tokamak fusion reactors. Off-normal events like disruptions, thermal and current quenches, are the perfect recipes for damage of plasma facing components (PFC). Successful operation of a fusion reactor requires comprehensive understanding of material erosion behavior. The extremely high heat fluxes deposited in PFCs melt and evaporate or directly sublime the exposed surfaces, which results in a thick vapor/melt boundary layer adjacent to the solid wall structure. The accumulating boundary layers provide a self-protecting nature by attenuating the radiant energy transport to the PFCs. The ultimate goal of this study is to develop a reliable tool to adequately simulate the effect of the boundary layers on the formation and flow of the energetic ET plasma and its impact on exposed surfaces erosion under disruption like conditions. This dissertation is a series of published journals/conferences papers. The first paper verified the existence of the vapor shield that evolved at the boundary layer under the typical operational conditions of the NC State University ET plasma facilities PIPE and SIRENS. Upon the verification of the vapor shield, the second paper proposed novel model to simulate the evolution of the boundary layer and its effectiveness in providing a self-protecting nature for the exposed plasma facing surfaces. The developed models simulate the radiant heat flux attenuation through an optically thick boundary layer. The models were validated by comparing the simulation results to experimental data taken from the ET plasma facilities. Upon validation of the boundary layer models, computational experiments were conducted with the purpose of evaluation the PFCs' erosion during plasma disruption in Tokamak fusion reactors. Erosion of a set of selected low-Z and high-Z materials were analyzed and discussed. For metallic plasma facing materials under the impact of hard and long time-scale disruption events, melting and melt-layer splashing become dominate erosion mechanisms during plasma-material interaction. In order to realistically assess the erosion of the metallic fusion reactor components, the fourth paper accounts for the various mechanisms by which material evolved from PFCs due to melting and vaporization, with a developed melting and splattering/splashing model incorporated in the ET plasma code. Also, the shielding effect associated with melt-layer and vapor-layer is investigated. The quantitative results of material erosion with the boundary layer effects including a vapor layer, melt layer and splashing effects is a new model and an important step towards achieving a better understanding of plasma-material interactions under exposure to such high heat flux conditions.

Plasma Surface Interactions Common to Advanced Fusion Wall Materials and EUV Lithography - Lithium and Tin

NASA Astrophysics Data System (ADS)

Ruzic, D. N.; Alman, D. A.; Jurczyk, B. E.; Stubbers, R.; Coventry, M. D.; Neumann, M. J.; Olczak, W.; Qiu, H.

2004-09-01

Advanced plasma facing components (PFCs) are needed to protect walls in future high power fusion devices. In the semiconductor industry, extreme ultraviolet (EUV) sources are needed for next generation lithography. Lithium and tin are candidate materials in both areas, with liquid Li and Sn plasma material interactions being critical. The Plasma Material Interaction Group at the University of Illinois is leveraging liquid metal experimental and computational facilities to benefit both fields. The Ion surface InterAction eXperiment (IIAX) has measured liquid Li and Sn sputtering, showing an enhancement in erosion with temperature for light ion bombardment. Surface Cleaning of Optics by Plasma Exposure (SCOPE) measures erosion and damage of EUV mirror samples, and tests cleaning recipes with a helicon plasma. The Flowing LIquid surface Retention Experiment (FLIRE) measures the He and H retention in flowing liquid metals, with retention coefficients varying between 0.001 at 500 eV to 0.01 at 4000 eV.

To Demonstrate an Integrated Solution for Plasma-Material Interfaces Compatible with an Optimized Core Plasma

NASA Astrophysics Data System (ADS)

Goldston, Robert; Brooks, Jeffrey; Hubbard, Amanda; Leonard, Anthony; Lipschultz, Bruce; Maingi, Rajesh; Ulrickson, Michael; Whyte, Dennis

2009-11-01

The plasma facing components in a Demo reactor will face much more extreme boundary plasma conditions and operating requirements than any present or planned experiment. These include 1) Power density a factor of four or more greater than in ITER, 2) Continuous operation resulting in annual energy and particle throughput 100-200 times larger than ITER, 3) Elevated surface operating temperature for efficient electricity production, 4) Tritium fuel cycle control for safety and breeding requirements, and 5) Steady state plasma confinement and control. Consistent with ReNeW Thrust 12, design options are being explored for a new moderate-scale facility to assess core-edge interaction issues and solutions. Key desired features include high power density, sufficient pulse length and duty cycle, elevated wall temperature, steady-state control of an optimized core plasma, and flexibility in changing boundary components as well as access for comprehensive measurements.

Formation of He-Rich Layers Observed by Neutron Reflectometry in the He-Ion-Irradiated Cr/W Multilayers: Effects of Cr/W Interfaces on the He-Trapping Behavior.

PubMed

Chen, Feida; Tang, Xiaobin; Huang, Hai; Li, Xinxi; Wang, Yan; Huang, Chaoqiang; Liu, Jian; Li, Huan; Chen, Da

2016-09-21

Cr/W multilayer nanocomposites were presented in the paper as potential candidate materials for the plasma facing components in fusion reactors. We used neutron reflectometry to measure the depth profile of helium in the multienergy He ions irradiated [Cr/W (50 nm)]3 multilayers. Results showed that He-rich layers with low neutron scattering potential energy form at the Cr/W interfaces, which is in great agreement with previous modeling results of other multilayers. This phenomenon provided a strong evidence for the He trapping effects of Cr/W interfaces and implied the possibility of using the Cr/W multilayer nanocomposites as great He-tolerant plasma facing materials.

Preparation of tungsten fiber reinforced-tungsten/copper composite for plasma facing component

NASA Astrophysics Data System (ADS)

He, Gang; Xu, Kunyuan; Guo, Shibin; Qian, Xueqiang; Yang, Zengchao; Liu, Guanghua; Li, Jiangtao

2014-12-01

W fiber reinforced-W/Cu composite is designed as a transition layer between CuCrZr heat sink material and W plasma facing material. A novel method was developed for the preparation of W fiber reinforced-W/Cu composite by combining combustion synthesis with centrifugal infiltration. Cu melt with a transient temperature over 2000 °C produced by the thermite reaction was infiltrated into the W powder and fiber bed with the assistance of a high gravity field. It was found that the W particles were sintered and bonded to the W fibers due to the high temperature produced by the thermite reaction. The bending strength of W/Cu composite improved 12.7% through W fibers reinforcement.

The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garrison, L. M., E-mail: garrisonlm@ornl.gov; Egle, B. J.; Fusion Technology Institute, University of Wisconsin-Madison, 1500 Engineering Drive, Madison, Wisconsin 53706

2016-08-15

The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000 °C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ionmore » gun can irradiate the samples with ion currents of 20 μA–500 μA; the typical current used is 72 μA, which is an average flux of 9 × 10{sup 14} ions/(cm{sup 2} s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. The MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.« less

Analysis of heat transfer and erosion effects on ITER divertor plasma facing components induced by slow high-power transients

DOE Office of Scientific and Technical Information (OSTI.GOV)

Federici, G.; Raffray, A.R.; Chiocchio, S.

1995-12-31

This paper presents the results of an analysis carried out to investigate the thermal response of ITER divertor plasma facing components (PFC`s) clad with Be, W, and CFC, to high-recycling, high-power thermal transients (i.e. 10--30 MW/m{sup 2}) which are anticipated to last up to a few seconds. The armour erosion and surface melting are estimated for the different plasma facing materials (PFM`s) together with the maximum heat flux to the coolant, and armour/heat-sink interface temperature. The analysis assumes that intense target evaporation will lead to high radiative power losses in the plasma in front of the target which self-protects themore » target. The cases analyzed clarify the influence of several key parameters such as the plasma heat flux to the target, the loss of the melt layer, the duration of the event, the thickness of the armour, and comparison is made with cases without vapor shielding. Finally, some implications for the performance and lifetime of divertor PFC`s clad with different PFM`s are discussed.« less

Nuclear Fusion Blast and Electrode Lifetimes in a PJMIF Reactor

NASA Astrophysics Data System (ADS)

Thio, Y. C. Francis; Witherspoon, F. D.; Case, A.; Brockington, S.; Cruz, E.; Luna, M.; Hsu, S. C.

2017-10-01

We present an analysis and numerical simulation of the nuclear blast from the micro-explosion following the completion of the fusion burn for a baseline design of a PJMIF fusion reactor with a fusion gain of 20. The stagnation pressure from the blast against the chamber wall defines the engineering requirement for the structural design of the first wall and the plasma guns. We also present an analysis of the lifetimes of the electrodes of the plasma guns which are exposed to (1) the high current, and (2) the neutron produced by the fusion reactions. We anticipate that the gun electrodes are made of tungsten alloys as plasma facing components reinforced structurally by appropriate steel alloys. Making reasonable assumptions about the electrode erosion rate (100 ng/C transfer), the electrode lifetime limited by the erosion rate is estimated to be between 19 and 24 million pulses before replacement. Based on known neutron radiation effects on structural materials such as steel alloys and plasma facing component materials such as tungsten alloys, the plasma guns are expected to survive some 22 million shots. At 1 Hz, this equal to about 6 months of continuous operation before they need to be replaced. Work supported by Strong Atomics, LLC.

Hydrogen isotope retention in beryllium for tokamak plasma-facing applications

NASA Astrophysics Data System (ADS)

Anderl, R. A.; Causey, R. A.; Davis, J. W.; Doerner, R. P.; Federici, G.; Haasz, A. A.; Longhurst, G. R.; Wampler, W. R.; Wilson, K. L.

Beryllium has been used as a plasma-facing material to effect substantial improvements in plasma performance in the Joint European Torus (JET), and it is planned as a plasma-facing material for the first wall (FW) and other components of the International Thermonuclear Experimental Reactor (ITER). The interaction of hydrogenic ions, and charge-exchange neutral atoms from plasmas, with beryllium has been studied in recent years with widely varying interpretations of results. In this paper we review experimental data regarding hydrogenic atom inventories in experiments pertinent to tokamak applications and show that with some very plausible assumptions, the experimental data appear to exhibit rather predictable trends. A phenomenon observed in high ion-flux experiments is the saturation of the beryllium surface such that inventories of implanted particles become insensitive to increased flux and to continued implantation fluence. Methods for modeling retention and release of implanted hydrogen in beryllium are reviewed and an adaptation is suggested for modeling the saturation effects. The TMAP4 code used with these modifications has succeeded in simulating experimental data taken under saturation conditions where codes without this feature have not. That implementation also works well under more routine conditions where the conventional recombination-limited release model is applicable. Calculations of tritium inventory and permeation in the ITER FW during the basic performance phase (BPP) using both the conventional recombination model and the saturation effects assumptions show a difference of several orders of magnitude in both inventory and permeation rate to the coolant.

Synergistic effects of surface erosion on tritium inventory and permeation in metallic plasma facing armours

NASA Astrophysics Data System (ADS)

Federici, G.; Holland, D. F.; Matera, R.

1996-10-01

In the next generation of DT fuelled tokamaks, i.e., the International Thermonuclear Experimental Reactor (ITER) implantation of energetic DT particles on some portions of the plasma facing components (PFCs) will take place along with significant erosion of the armour surfaces. As a result of the simultaneous removal of material from the front surface, the build-up of tritium inventory and the start of permeation originating in the presence of large densities of neutron-induced traps is expected to be influenced considerably and special provisions could be required to minimise the consequences on the design. This paper reports on the results of a tritium transport modelling study based on a new model which describes the migration of implanted tritium across the bulk of metallic plasma facing materials containing neutron-induced traps which can capture it and includes the synergistic effects of surface erosion. The physical basis of the model is summarised, but emphasis is on the discussion of the results of a comparative study performed for beryllium and tungsten armours for ranges of design and operation conditions similar to those anticipated in the divertor of ITER.

Design and Construction of Field Reversed Configuration Plasma Chamber for Plasma Material Interaction Studies

NASA Astrophysics Data System (ADS)

Smith, DuWayne L.

A Field Reversed Configuration (FRC) plasma source was designed and constructed to conduct high energy plasma-materials interaction studies. The purpose of these studies is the development of advanced materials for use in plasma based electric propulsion systems and nuclear fusion containment vessels. Outlined within this thesis is the basic concept of FRC plasmoid creation, an overview of the device design and integration of various diagnostics systems for plasma conditions and characterization, discussion on the variety of material defects resulting from the plasma exposure with methods and tools designed for characterization. Using a Michelson interferometer it was determined that the FRC plasma densities are on the order of ~1021 m-3. A novel dynamic pressure probe was created to measure ion velocities averaging 300 km/s. Compensating flux loop arrays were used to measure magnetic field strength and verify the existence of the FRC plasmoid and when used in combination with density measurements it was determined that the average ion temperatures are ~130 eV. X-ray Photoelectron Spectroscopy (XPS) was employed as a means of characterizing the size and shape of the plasma jet in the sample exposure positions. SEM results from preliminary studies reveal significant morphological changes on plasma facing material surfaces, and use of XRD to elucidate fuel gas-ion implantation strain rates correlated to plasma exposure energies.

Integrated Prediction and Mitigation Methods of Materials Damage and Lifetime Assessment during Plasma Operation and Various Instabilities in Fusion Devices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hassanein, Ahmed

2015-03-31

This report describes implementation of comprehensive and integrated models to evaluate plasma material interactions during normal and abnormal plasma operations. The models in full3D simulations represent state-of-the art worldwide development with numerous benchmarking of various tokamak devices and plasma simulators. In addition, significant number of experimental work has been performed in our center for materials under extreme environment (CMUXE) at Purdue to benchmark the effect of intense particle and heat fluxes on plasma-facing components. This represents one-year worth of work and resulted in more than 23 Journal Publications and numerous conferences presentations. The funding has helped several students to obtainmore » their M.Sc. and Ph.D. degrees and many of them are now faculty members in US and around the world teaching and conducting fusion research. Our work has also been recognized through many awards.« less

Relativistic electron beam device

DOEpatents

Freeman, J.R.; Poukey, J.W.; Shope, S.L.; Yonas, G.

1975-07-01

A design is given for an electron beam device for irradiating spherical hydrogen isotope bearing targets. The accelerator, which includes hollow cathodes facing each other, injects an anode plasma between the cathodes and produces an approximately 10 nanosecond, megajoule pulse between the anode plasma and the cathodes. Targets may be repetitively positioned within the plasma between the cathodes, and accelerator diode arrangement permits materials to survive operation in a fusion power source. (auth)

Plasma facing materials performance under ITER-relevant mitigated disruption photonic heat loads

NASA Astrophysics Data System (ADS)

Klimov, N. S.; Putrik, A. B.; Linke, J.; Pitts, R. A.; Zhitlukhin, A. M.; Kuprianov, I. B.; Spitsyn, A. V.; Ogorodnikova, O. V.; Podkovyrov, V. L.; Muzichenko, A. D.; Ivanov, B. V.; Sergeecheva, Ya. V.; Lesina, I. G.; Kovalenko, D. V.; Barsuk, V. A.; Danilina, N. A.; Bazylev, B. N.; Giniyatulin, R. N.

2015-08-01

PFMs (Plasma-facing materials: ITER grade stainless steel, beryllium, and ferritic-martensitic steels) as well as deposited erosion products of PFCs (Be-like, tungsten, and carbon based) were tested in QSPA under photonic heat loads relevant to those expected from photon radiation during disruptions mitigated by massive gas injection in ITER. Repeated pulses slightly above the melting threshold on the bulk materials eventually lead to a regular, "corrugated" surface, with hills and valleys spaced by 0.2-2 mm. The results indicate that hill growth (growth rate of ∼1 μm per pulse) and sample thinning in the valleys is a result of melt-layer redistribution. The measurements on the 316L(N)-IG indicate that the amount of tritium absorbed by the sample from the gas phase significantly increases with pulse number as well as the modified layer thickness. Repeated pulses significantly below the melting threshold on the deposited erosion products lead to a decrease of hydrogen isotopes trapped during the deposition of the eroded material.

Characterization and damaging law of CFC for high heat flux actively cooled plasma facing components

NASA Astrophysics Data System (ADS)

Chevet, G.; Martin, E.; Boscary, J.; Camus, G.; Herb, V.; Schlosser, J.; Escourbiac, F.; Missirlian, M.

2011-10-01

The carbon fiber reinforced carbon composite (CFC) Sepcarb N11 has been used in the Tore Supra (TS) tokamak (Cadarache, France) as armour material for the plasma facing components. For the fabrication of the Wendelstein 7-X (W7-X) divertor (Greifswald, Germany), the NB31 material was chosen. For the fabrication of the ITER divertor, two potential CFC candidates are the NB31 and NB41 materials. In the case of Tore Supra, defects such as microcracks or debonding were found at the interface between CFC tile and copper heat sink. A mechanical characterization of the behaviour of N11 and NB31 was undertaken, allowing the identification of a damage model and finite element calculations both for flat tiles (TS and W7-X) and monoblock (ITER) armours. The mechanical responses of these CFC materials were found almost linear under on-axis tensile tests but highly nonlinear under shear tests or off-axis tensile tests. As a consequence, damage develops within the high shear-stress zones.

Ti-doped isotropic graphite: A promising armour material for plasma-facing components

NASA Astrophysics Data System (ADS)

García-Rosales, C.; López-Galilea, I.; Ordás, N.; Adelhelm, C.; Balden, M.; Pintsuk, G.; Grattarola, M.; Gualco, C.

2009-04-01

Finely dispersed Ti-doped isotropic graphites with 4 at.% Ti have been manufactured using synthetic mesophase pitch 'AR' as raw material. These new materials show a thermal conductivity at room temperature of ˜200 W/mK and flexural strength close to 100 MPa. Measurement of the total erosion yield by deuterium bombardment at ion energies and sample temperatures for which pure carbon shows maximum values, resulted in a reduction of at least a factor of 4, mainly due to dopant enrichment at the surface caused by preferential erosion of carbon. In addition, ITER relevant thermal shock loads were applied with an energetic electron beam at the JUDITH facility. The results demonstrated a significantly improved performance of Ti-doped graphite compared to pure graphite. Finally, Ti-doped graphite was successfully brazed to a CuCrZr block using a Mo interlayer. These results let assume that Ti-doped graphite can be a promising armour material for divertor plasma-facing components.

Erosion resistant nozzles for laser plasma extreme ultraviolet (EUV) sources

DOEpatents

Kubiak, Glenn D.; Bernardez, II, Luis J.

2000-01-04

A gas nozzle having an increased resistance to erosion from energetic plasma particles generated by laser plasma sources. By reducing the area of the plasma-facing portion of the nozzle below a critical dimension and fabricating the nozzle from a material that has a high EUV transmission as well as a low sputtering coefficient such as Be, C, or Si, it has been shown that a significant reduction in reflectance loss of nearby optical components can be achieved even after exposing the nozzle to at least 10.sup.7 Xe plasma pulses.

Neutron-Irradiated Samples as Test Materials for MPEX

DOE PAGES

Ellis, Ronald James; Rapp, Juergen

2015-10-09

Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by fast neutron irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. This paper presents assessments of the calculated induced radioactivity and resulting radiation dose rates of a variety of potential fusion reactor plasma-facing materials (such as tungsten). The scientific code packages MCNP and SCALE were used to simulate irradiation of themore » samples in HFIR including the generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. A challenge of the MPEX project is to minimize the radioactive inventory in the preparation of the samples and the sample dose rates for inclusion in the MPEX facility.« less

Overview of the recent DiMES and MiMES experiments in DIII-D

NASA Astrophysics Data System (ADS)

Rudakov, D. L.; Wong, C. P. C.; Litnovsky, A.; Wampler, W. R.; Boedo, J. A.; Brooks, N. H.; Fenstermacher, M. E.; Groth, M.; Hollmann, E. M.; Jacob, W.; Krasheninnikov, S. I.; Krieger, K.; Lasnier, C. J.; Leonard, A. W.; McLean, A. G.; Marot, M.; Moyer, R. A.; Petrie, T. W.; Philipps, V.; Smirnov, R. D.; Stangeby, P. C.; Watkins, J. G.; West, W. P.; Yu, J. H.

2009-12-01

Divertor and midplane material evaluation systems (DiMES and MiMES) in the DIII-D tokamak are used to address a variety of plasma-material interaction (PMI) issues relevant to ITER. Among the topics studied are carbon erosion and re-deposition, hydrogenic retention in the gaps between plasma-facing components (PFCs), deterioration of diagnostic mirrors from carbon deposition and techniques to mitigate that deposition, and dynamics and transport of dust. An overview of the recent experimental results is presented.

Spark plasma sintering of pure and doped tungsten as plasma facing material

NASA Astrophysics Data System (ADS)

Autissier, E.; Richou, M.; Minier, L.; Naimi, F.; Pintsuk, G.; Bernard, F.

2014-04-01

In the current water cooled divertor concept, tungsten is an armour material and CuCrZr is a structural material. In this work, a fabrication route via a powder metallurgy process such as spark plasma sintering is proposed to fully control the microstructure of W and W composites. The effect of chemical composition (additives) and the powder grain size was investigated. To reduce the sintering temperature, W powders doped with a nano-oxide dispersion of Y2O3 are used. Consequently, the sintering temperature for W-oxide dispersed strengthened (1800 °C) is lower than for pure W powder. Edge localized mode tests were performed on pure W and compared to other preparation techniques and showed promising results.

Manufacturing and characterization of PIM-W materials as plasma facing materials

NASA Astrophysics Data System (ADS)

Pintsuk, G.; Antusch, S.; Rieth, M.; Wirtz, M.

2016-02-01

Powder injection molding (PIM) was used to produce pure and particle reinforced W materials to be qualified for the use as plasma facing material. As alloying elements La2O3, Y2O3, TiC, and TaC were chosen with a particle size between 50 nm and 2.5 μm, depending on the alloying element. The fabrication of alloyed materials was done for different compositions using powder mixtures. Final sintering was performed in H2 atmosphere at 2400 °C resulting in plates of 55 × 22 × 4 mm3 with ˜98% theoretical density. The qualification of the materials was done via high heat flux testing in the electron beam facility JUDITH-1. Thereby, ELM-like 1000 thermal shock loads of 0.38 GW m-2 for 1 ms and 100 disruption like loads of 1.13 GW m-2 for 1 ms at a base temperature of 1000 °C were applied. The obtained damage characteristics, i.e. surface roughening and crack formation, were qualified versus an industrially manufactured pure reference tungsten material and linked to the material’s microstructure and mechanical properties.

Recent advances in modeling and simulation of the exposure and response of tungsten to fusion energy conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marian, Jaime; Becquart, Charlotte S.; Domain, Christophe

2017-06-09

Under the anticipated operating conditions for demonstration magnetic fusion reactors beyond ITER, structural materials will be exposed to unprecedented conditions of irradiation, heat flux, and temperature. While such extreme environments remain inaccessible experimentally, computational modeling and simulation can provide qualitative and quantitative insights into materials response and complement the available experimental measurements with carefully validated predictions. For plasma facing components such as the first wall and the divertor, tungsten (W) has been selected as the best candidate material due to its superior high-temperature and irradiation properties. In this paper we provide a review of recent efforts in computational modeling ofmore » W both as a plasma-facing material exposed to He deposition as well as a bulk structural material subjected to fast neutron irradiation. We use a multiscale modeling approach –commonly used as the materials modeling paradigm– to define the outline of the paper and highlight recent advances using several classes of techniques and their interconnection. We highlight several of the most salient findings obtained via computational modeling and point out a number of remaining challenges and future research directions« less

Investigation of Plasma Surface Interactions with the PISCES ELM Laser System

NASA Astrophysics Data System (ADS)

Umstadter, K. R.; Baldwin, M.; Hanna, J.; Doerner, R.; Lynch, T.; Palmer, T.; Tynan, G. R.

2007-11-01

When an ELM occurs in tokamaks, up to 30% of the pedestal energy can be deposited on the wall of the tokamak causing heating & material loss due to sublimation, evaporation and melt splashing of plasma facing components (PFCs) and expansion of the ejected material into the plasma. We have explored heat pulses using an electrical power circuit to draw electrons from the plasma to heat samples ohmically. This system is limited in power to ˜250kJ/m^2 at the minimum pulse width of 10ms and depletes the plasma column, complicating spectroscopy. We have completed calculations that indicate that a pulsed laser system can be used to simulate the heat pulse of ELMs. We are integrating laser systems into the existing PFC research program in PISCES, a laboratory facility capable of reproducing plasma-materials interactions expected during normal operation of large tokamaks. Two Nd:YAG lasers capable of delivering up to 50J of energy over various pulsewidths are used for the experiments. Laser heat pulse only, H+/D+ plasma only, and laser+plasma experiments were conducted and initial results indicate that metals behave very differently while exposed to plasma and simultaneous heat pulses. We will also discuss initial results for carbon PFCs and material transport into the plasma. Supported by US DoE grant DE-FG02-07ER-54912.

Deuterium retention and release from molybdenum exposed to a Penning discharge

NASA Astrophysics Data System (ADS)

Causey, R. A.; Kunz, C. L.; Cowgill, D. F.

2005-03-01

Both molybdenum and tungsten are candidate materials for plasma-facing applications in fusion reactors. While tungsten has a higher melting point and a higher threshold for sputtering, it is a brittle material that is difficult to machine into shapes required for fusion applications. For this reason, molybdenum is now receiving serious consideration as an alternative for tungsten. If molybdenum is to be used as a plasma-facing material, the hydrogen retention and recycling characteristics must be known. In this report, we present experimental results on deuterium retention in molybdenum after exposure to a Penning discharge at temperatures from 573 to 773 K. D2+ ions with energies of 1.2 keV were implanted into the 50 mm diameter molybdenum samples at fluxes of 10 20 D/m 2 s. Thermal desorption spectroscopy was used to determine both the amount of retained deuterium and the release kinetics. Low retention values similar to those measured previously for tungsten were observed.

Interaction of plasmas with lithium and tungsten fusion plasma facing components

NASA Astrophysics Data System (ADS)

Fiflis, Peter Robert

One of the largest outstanding issues in magnetic confinement fusion is the interaction of the fusion plasma with the first wall of the device; an interaction which is strongest in the divertor region. Erosion, melting, sputtering, and deformation are all concerns which inform choices of divertor material. Of the many materials proposed for use in the divertor, only a few remain as promising choices. Tungsten has been chosen as the material for the ITER divertor, and liquid lithium stands poised as its replacement in higher heat flux devices. As a refractory metal, tungsten's large melting point and thermal conductivity as well as its low sputtering yield have led to its selection as the material of choice of the ITER divertor. Experiments have reinforced this choice demonstrating tungsten's ability to withstand large heat fluxes when adequately cooled. However, tungsten has shown a propensity to nanostructure under exposure within a certain temperature range to large fluxes of helium ions. These nanostructures if disrupted into the plasma as dust by an off-normal event would cause quenching of the plasma from the generated dust. Liquid lithium, meanwhile, has gathered growing interest within the fusion community in recent years as a divertor, limiter, and alternative first wall material. Liquid lithium is attractive as a low-Z material replacement for refractory metals due to its ability to getter impurities, while also being self-healing in nature. However, concerns exist about the stability of a liquid metal surface at the edge of a fusion device. Liquid metal pools, such as the Li-DiMes probe, have shown evidence of macroscopic lithium displacement as well as droplet formation and ejection into the plasma. These issues must be mitigated in future implementations of liquid lithium divertor concepts. Rayleigh-Taylor-like (RT) and Kelvin-Helmholtz-like (KH) instabilities have been claimed as the initiators of droplet ejection, yet not enough data exists to delineate a stability boundary. The influences of plasma pressure and current driven instabilities on lithium surfaces that lead to droplet ejection are investigated to determine which of the two effects is dominant for a given set of plasma conditions. This work studies the influence of large plasma fluxes on these two materials to better inform the selection and design of plasma facing components (PFCs). The nanostructuring of tungsten was investigated to determine the mechanisms by which tungsten nanostructures so that its formation may be mitigated. Experiments investigated the dependence of nanostructuring on temperature, looked at the morphological evolution, and grew nanostructures on a variety of metals to examine their similarity to tungsten. Additionally, a computational model is presented for the initial stages of fuzz formation showing good quantitative and qualitative agreement with experimental observations. The influences of RT and KH instabilities on the surface of liquid lithium were experimentally observed and quantified on the ThermoElectric-driven Liquid-metal plasma-facing Structures (TELS) chamber at the University of Illinois at Urbana-Champaign and the stabilizing effect of surface tension, an effect employed by the LiMIT concept as well as other liquid lithium concepts, was studied, and the stability boundary afforded by surface tension was compared between experiment, computational simulation, and theory.

Optimisation and characterisation of tungsten thick coatings on copper based alloy substrates

NASA Astrophysics Data System (ADS)

Riccardi, B.; Montanari, R.; Casadei, M.; Costanza, G.; Filacchioni, G.; Moriani, A.

2006-06-01

Tungsten is a promising armour material for plasma facing components of nuclear fusion reactors because of its low sputter rate and favourable thermo-mechanical properties. Among all the techniques able to realise W armours, plasma spray looks particularly attractive owing to its simplicity and low cost. The present work concerns the optimisation of spraying parameters aimed at 4-5 mm thick W coating on copper-chromium-zirconium (Cu,Cr,Zr) alloy substrates. Characterisation of coatings was performed in order to assess microstructure, impurity content, density, tensile strength, adhesion strength, thermal conductivity and thermal expansion coefficient. The work performed has demonstrated the feasibility of thick W coatings on flat and curved geometries. These coatings appear as a reliable armour for medium heat flux plasma facing component.

New oxidation-resistant tungsten alloys for use in the nuclear fusion reactors

NASA Astrophysics Data System (ADS)

Litnovsky, A.; Wegener, T.; Klein, F.; Linsmeier, Ch; Rasinski, M.; Kreter, A.; Tan, X.; Schmitz, J.; Coenen, J. W.; Mao, Y.; Gonzalez-Julian, J.; Bram, M.

2017-12-01

Smart tungsten-based alloys are under development as plasma-facing components for a future fusion power plant. Smart alloys are planned to adjust their properties depending on environmental conditions: acting as a sputter-resistant plasma-facing material during plasma operation and suppressing the sublimation of radioactive tungsten oxide in case of an accident on the power plant. New smart alloys containing yttrium are presently in the focus of research. Thin film smart alloys are featuring an remarkable 105-fold suppression of mass increase due to an oxidation as compared to that of pure tungsten at 1000 °C. Newly developed bulk smart tungsten alloys feature even better oxidation resistance compared to that of thin films. First plasma test of smart alloys under DEMO-relevant conditions revealed the same mass removal as for pure tungsten due to sputtering by plasma ions. Exposed smart alloy samples demonstrate the superior oxidation performance as compared to tungsten-chromium-titanium systems developed earlier.

Implementation of a diffusion convection surface evolution model in WallDYN

NASA Astrophysics Data System (ADS)

Schmid, K.

2013-07-01

In thermonuclear fusion experiments with multiple plasma facing materials the formation of mixed materials is inevitable. The formation of these mixed material layers is a dynamic process driven the tight interaction between transport in the plasma scrape off layer and erosion/(re-) deposition at the surface. To track this global material erosion/deposition balance and the resulting formation of mixed material layers the WallDYN code has been developed which couples surface processes and plasma transport. The current surface model in WallDYN cannot fully handle the growth of layers nor does it include diffusion. However at elevated temperatures diffusion is a key process in the formation of mixed materials. To remedy this shortcoming a new surface model has been developed which, for the first time, describes both layer growth/recession and diffusion in a single continuous diffusion/convection equation. The paper will detail the derivation of the new surface model and compare it to TRIDYN calculations.

Innovative divertor concept development on DIII-D and EAST

DOE PAGES

Guo, H. Y.; Allen, S.; Canik, J.; ...

2016-06-02

A critical issue facing the design and operation of next-step high-power steady-state fusion devices is the control of heat fluxes and erosion at the plasma-facing components, in particular, the divertor target plates. A new initiative has been launched on DIII-D to develop and demonstrate innovative boundary plasma-materials interface solutions. The central purposes of this new initiative are to advance scientific understanding in this critical area and develop an advanced divertor concept for application to next-step fusion devices. Finally, DIII-D will leverage strong collaborative efforts on the EAST superconducting tokamak for extending integrated high performance advanced divertor solutions to true steady-state.

Measurements of line-averaged electron density of pulsed plasmas using a He-Ne laser interferometer in a magnetized coaxial plasma gun device

NASA Astrophysics Data System (ADS)

Iwamoto, D.; Sakuma, I.; Kitagawa, Y.; Kikuchi, Y.; Fukumoto, N.; Nagata, M.

2012-10-01

In next step of fusion devices such as ITER, lifetime of plasma-facing materials (PFMs) is strongly affected by transient heat and particle loads during type I edge localized modes (ELMs) and disruption. To clarify damage characteristics of the PFMs, transient heat and particle loads have been simulated by using a plasma gun device. We have performed simulation experiments by using a magnetized coaxial plasma gun (MCPG) device at University of Hyogo. The line-averaged electron density measured by a He-Ne interferometer is 2x10^21 m-3 in a drift tube. The plasma velocity measured by a time of flight technique and ion Doppler spectrometer was 70 km/s, corresponding to the ion energy of 100 eV for helium. Thus, the ion flux density is 1.4x10^26 m-2s-1. On the other hand, the MCPG is connected to a target chamber for material irradiation experiments. It is important to measure plasma parameters in front of target materials in the target chamber. In particular, a vapor cloud layer in front of the target material produced by the pulsed plasma irradiation has to be characterized in order to understand surface damage of PFMs under ELM-like plasma bombardment. In the conference, preliminary results of application of the He-Ne laser interferometer for the above experiment will be shown.

Particle and heat flux estimates in Proto-MPEX in Helicon Mode with IR imaging

NASA Astrophysics Data System (ADS)

Showers, M. A.; Biewer, T. M.; Caughman, J. B. O.; Donovan, D. C.; Goulding, R. H.; Rapp, J.

2016-10-01

The Prototype Material Plasma Exposure eXperiment (Proto-MPEX) at Oak Ridge National Laboratory (ORNL) is a linear plasma device developing the plasma source concept for the Material Plasma Exposure eXperiment (MPEX), which will address plasma material interaction (PMI) science for future fusion reactors. To better understand how and where energy is being lost from the Proto-MPEX plasma during ``helicon mode'' operations, particle and heat fluxes are quantified at multiple locations along the machine length. Relevant diagnostics include infrared (IR) cameras, four double Langmuir probes (LPs), and in-vessel thermocouples (TCs). The IR cameras provide temperature measurements of Proto-MPEX's plasma-facing dump and target plates, located on either end of the machine. The change in surface temperature is measured over the duration of the plasma shot to determine the heat flux hitting the plates. The IR cameras additionally provide 2-D thermal load distribution images of these plates, highlighting Proto-MPEX plasma behaviors, such as hot spots. The LPs and TCs provide additional plasma measurements required to determine particle and heat fluxes. Quantifying axial variations in fluxes will help identify machine operating parameters that will improve Proto-MPEX's performance, increasing its PMI research capabilities. This work was supported by the U.S. D.O.E. contract DE-AC05-00OR22725.

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. Ono; Jaworski, M.; Kaita, R.

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 NSTX-U, the PMI research has received a strong emphasis. Moreover, with ˜15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m 2.

Tungsten as a plasma-facing material in fusion devices: impact of helium high-temperature irradiation on hydrogen retention and damages in the material

NASA Astrophysics Data System (ADS)

Bernard, E.; Sakamoto, R.; Kreter, A.; Barthe, M. F.; Autissier, E.; Desgardin, P.; Yamada, H.; Garcia-Argote, S.; Pieters, G.; Chêne, J.; Rousseau, B.; Grisolia, C.

2017-12-01

Plasma-facing materials for next generation fusion devices, like ITER and DEMO, have to withstand intense fluxes of light elements (notably helium and hydrogen isotopes). For tungsten (W), helium (He) irradiation leads to major changes in the material morphology, rising concerns about properties such as material structure conservation and hydrogen (H) retention. The impact of preceeding He irradiation conditions (temperature, flux and fluence) on H trapping were investigated on a set of W samples exposed to the linear plasma device PSI-2. Positron annihilation spectroscopy (PAS) was carried out to probe the free volume of defects created by the He exposure in the W structure at the atomic scale. In parallel, tritium (T) inventory after exposure was evaluated through T gas loading and desorption at the Saclay Tritium Lab. First, we observed that the material preparation prior to He irradiation was crucial, with a major reduction of the T trapping when W was annealed at 1773 K for 2 h compared to the as-received material. PAS study confirms the presence of He in the bubbles created in the material surface layer, whose dimensions were previously characterized by transmission electron microscopy and grazing-incidence small-angle x-ray scattering, and demonstrates that even below the minimal energy for displacement of He in W, defects are created in almost all He irradiation conditions. The T loading study highlights that increasing the He fluence leads to higher T inventory. Also, for a given fluence, increasing the He flux reduces the T trapping. The very first steps of a parametric study were set to understand the mechanisms at stake in those observed material modifications, confirming the need to pursue the study with a more complete set of surface and irradiation conditions.

Assessing material properties for fusion applications by ion beams

NASA Astrophysics Data System (ADS)

Catarino, N.; Dias, M.; Jepu, I.; Alves, E.

2017-10-01

The plasma-facing materials in the ITER divertor area must withstand unusual events, such as the edge-localized modes (ELMS). At the point when an ELM occurs, up to 30% of the energy can be deposited on the plasma-facing boundary in the form of the heat and particle load causing material loss due to sublimation. Tungsten is a promising candidate as a plasma-facing material in the ITER divertor area since it has a high melting point, good thermal conductivity and low sputtering yield, which minimizes the plasma contamination. However their brittleness at low temperatures which is worsened by irradiation is an issue. One strategy to modulate the properties of tungsten is alloying this element with other refractory metals, such as tantalum that shows higher toughness, lower activation and higher radiation resistance. In the present study tungsten-tantalum alloys (W-Ta) were produced by Ta implantation. The fundamental mechanisms which govern the behaviour of defect dynamics in W-Ta materials under reactor conditions, were simulated by the implantation of He and D. The microstructure observations of the W plates that after single Ta implantation revealed crater-like cavities and a more severe effect after D implantation. The effect increase with the increasing of D fluence. However at fluences higher than 1021D/m the effect is reduced. In addition, blistering was observed in W-Ta plates implanted with He. The D retention in the W-Ta alloys increases with the implanted fluence with tendency for saturation for high fluences. Moreover the results show that D retention is higher after sequential He and D implantation than for single D implantation. The diffractogram of W-Ta alloys implanted with He evidenced the presence of broadened W peaks associated with stress induced by irradiation, which may cause internal stress field resulting in a distortion of the crystal lattice. These irradiation defects can be observed in the D release spectra where three peaks are associated with three types of defects in W and W-Ta implanted with He and D.

Commissioning and experimental validation of SST-1 plasma facing components

NASA Astrophysics Data System (ADS)

Paravastu, Yuvakiran; Raval, Dilip; Khan, Ziauddin; Patel, Hitesh; Biswas, Prabal; Parekh, Tejas; George, Siju; Santra, Prosenjit; Ramesh, Gattu; ArunPrakash, A.; Thankey, Prashant; Semwal, Pratibha; Dhanani, Kalpeshkumar R.; Jaiswal, Snehal; Chauhan, Pradeep; Pradhan, Subrata

2017-04-01

Plasma facing components of SST-1 are designed to withstand an input heat load of 1.0 MW/m2. They protect vacuum vessel, auxiliary heating source i.e. RF antennas, NBI and other in-vessel diagnostic from the plasma particles and high radiative heat loads. PFC’s are positioned symmetric to mid-plane to accommodate with circular, single and double null configuration. Graphite is used as plasma facing material, back made of copper alloy and SS cooling/baking tubes are brazed on copper alloy back plates for efficient heat removal of incident heat flux. Benchmarking of PFC assembly was first carried out in prototype vacuum vessel of SST-1 to develop understanding and methodology of co-ordinate measurements. Based on such hands-on-experience, the final assembly of PFC’s in vacuum vessel of SST-1 was carried out. Initially, PFC’s are to be baked at 250 °C for wall conditioning followed with cooling for heat removal of incident heat flux during long pulse plasma operation. For this purpose, the supply and return headers are designed and installed inside the vacuum vessel in such a way that it will cater water as well as hot nitrogen gas depending up on the cycle. This paper will discuss the successful installation of PFC’s and its plasma operation respecting all design criteria.

System for the production of plasma

DOEpatents

Bakken, George S.

1978-01-01

The present invention provides a system for the production of a plasma by concentrating and focusing a laser beam on the plasma-forming material with a lightfocusing member which comprises a parabolic axicon in conjunction with a coaxial conical mirror. The apex of the conical mirror faces away from the focus of the parabolic axicon such that the conical mirror serves to produce a virtual line source along the axis of the cone. Consequently, irradiation from a laser parallel to the axis toward the apex of the conical mirror will be concentrated at the focus of the parabolic axicon, impinging upon the plasma-forming material there introduced to produce a plasma. The system is adaptable to irradiation of a target pellet introduced at the focus of the parabolic axicon and offers an advantage in that the target pellet can be irradiated with a high degree of radial and spherical symmetry.

High-Z material erosion and its control in DIII-D carbon divertor

DOE PAGES

Ding, Rui; Rudakov, Dimitry L.; Stangeby, Peter C.; ...

2017-03-16

It is expected that high-Z materials will be used as plasma-facing components (PFCs) in future fusion devices, making the erosion of high-Z material a key issue for high-power, long pulse operation. High-Z material erosion and redeposition have been studied using tungsten and molybdenum coated samples exposed in well-diagnosed DIII-D divertor plasma discharges. By coupling dedicated experiments and modelling using the 3D Monte Carlo code ERO, the roles of sheath potential and background carbon impurities in determining high-Z material erosion are identified. Different methods suggested by modelling have been investigated to control high-Z material erosion in DIII-D experiments. The erosion ofmore » Mo and W are found to be strongly suppressed by local injection of methane and deuterium gases. The 13C deposition resulting from local 13CH 4 injection also provides information on radial transport due to E×B drifts and cross field diffusion. Finally, D 2 gas puffing is found to cause 2 local plasma perturbation, suppressing W erosion because of the lower effective sputtering yield of W at lower plasma temperature and for higher carbon concentration in the mixed surface layer.« less

Global modeling of wall material migration following boronization in NSTX-U

NASA Astrophysics Data System (ADS)

Nichols, J. H.; Jaworski, M. A.; Skinner, C. H.; Bedoya, F.; Scotti, F.; Soukhanovskii, V. A.; Schmid, K.

2017-10-01

NSTX-U operated in 2016 with graphite plasma facing components, periodically conditioned with boron to improve plasma performance. Following each boronization, spectroscopic diagnostics generally observed a decrease in oxygen influx from the walls, and an in-vacuo material probe (MAPP) observed a corresponding decrease in surface oxygen concentration at the lower divertor. However, oxygen levels tended to return to a pre-boronization state following repeated plasma exposure. This behavior is interpretively modeled using the WallDYN mixed-material migration code, which couples local erosion and deposition processes with plasma impurity transport in a non-iterative, self-consistent manner that maintains overall material balance. A spatially inhomogenous model of the thin films produced by the boronization process is presented. Plasma backgrounds representative of NSTX-U conditions are reconstructed from a combination of NSTX-U and NSTX datasets. Low-power NSTX-U fiducial discharges, which led to less apparent surface degradation than normal operations, are also modeled with WallDYN. Likely mechanisms driving the observed evolution of surface oxygen are examined, as well as remaining discrepancies between model and experiment and potential improvements to the model. Work supported by US DOE contract DE-AC02-09CH11466.

Transpiration cooled electrodes and insulators for MHD generators

DOEpatents

Hoover, Jr., Delmer Q.

1981-01-01

Systems for cooling the inner duct walls in a magnetohydrodynamic (MHD) generator. The inner face components, adjacent the plasma, are formed of a porous material known as a transpiration material. Selected cooling gases are transpired through the duct walls, including electrically insulating and electrode segments, and into the plasma. A wide variety of structural materials and coolant gases at selected temperatures and pressures can be utilized and the gases can be drawn from the generation system compressor, the surrounding environment, and combustion and seed treatment products otherwise discharged, among many other sources. The conduits conducting the cooling gas are electrically insulated through low pressure bushings and connectors so as to electrically isolate the generator duct from the ground.

Hydrogen permeation properties of plasma-sprayed tungsten*1

NASA Astrophysics Data System (ADS)

Anderl, R. A.; Pawelko, R. J.; Hankins, M. R.; Longhurst, G. R.; Neiser, R. A.

1994-09-01

Tungsten has been proposed as a plasma-facing component material for advanced fusion facilities. This paper reports on laboratory-scale studies that were done to assess the hydrogen permeation properties of plasma-sprayed tungsten for such applications. The work entailed deuterium permeation measurements for plasma-sprayed (PS) tungsten coatings, sputter-deposited (SP) tungsten coatings, and steel substrate material using a mass-analyzed, 3 keV D 3+ ion beam with fluxes of ˜6.5 × 10 19 D/m 2 s. Extensive characterization analyses for the plasma-sprayed tungsten coatings were made using Auger spectrometry and scanning electron microscopy (SEM). Observed permeation rates through composite PS-tungsten/steel specimens were several orders of magnitude below the permeation levels observed for SP-tungsten/steel composite specimens and pure steel specimens. Characterization analyses indicated that the plasma-sprayed tungsten coating had a nonhomogeneous microstructure that consisted of splats with columnar solidification, partially-melted particles with grain boundaries, and void regions. Reduced permeation levels can be attributed to the complex microstructure and a substantial surface-connected porosity.

Hydrogen transport behavior of beryllium

NASA Astrophysics Data System (ADS)

Anderl, R. A.; Hankins, M. R.; Longhurst, G. R.; Pawelko, R. J.; Macaulay-Newcombe, R. G.

1992-12-01

Beryllium is being evaluated for use as a plasma-facing material in the International Thermonuclear Experimental Reactor (ITER). One concern in the evaluation is the retention and permeation of tritium implanted into the plasma-facing surface. We performed laboratory-scale studies to investigate mechanisms that influence hydrogen transport and retention in beryllium foil specimens of rolled powder metallurgy product and rolled ingot cast beryllium. Specimen characterization was accomplished using scanning electron microscopy, Auger electron spectroscopy, and Rutherford backscattering spectrometry (RBS) techniques. Hydrogen transport was investigated using ion-beam permeation experiments and nuclear reaction analysis (NRA). Results indicate that trapping plays a significant role in permeation, re-emission, and retention, and that surface processes at both upstream and downstream surfaces are also important.

Monte Carlo simulation of ion-material interactions in nuclear fusion devices

NASA Astrophysics Data System (ADS)

Nieto Perez, M.; Avalos-Zuñiga, R.; Ramos, G.

2017-06-01

One of the key aspects regarding the technological development of nuclear fusion reactors is the understanding of the interaction between high-energy ions coming from the confined plasma and the materials that the plasma-facing components are made of. Among the multiple issues important to plasma-wall interactions in fusion devices, physical erosion and composition changes induced by energetic particle bombardment are considered critical due to possible material flaking, changes to surface roughness, impurity transport and the alteration of physicochemical properties of the near surface region due to phenomena such as redeposition or implantation. A Monte Carlo code named MATILDA (Modeling of Atomic Transport in Layered Dynamic Arrays) has been developed over the years to study phenomena related to ion beam bombardment such as erosion rate, composition changes, interphase mixing and material redeposition, which are relevant issues to plasma-aided manufacturing of microelectronics, components on object exposed to intense solar wind, fusion reactor technology and other important industrial fields. In the present work, the code is applied to study three cases of plasma material interactions relevant to fusion devices in order to highlight the code's capabilities: (1) the Be redeposition process on the ITER divertor, (2) physical erosion enhancement in castellated surfaces and (3) damage to multilayer mirrors used on EUV diagnostics in fusion devices due to particle bombardment.

Tritium Plasma Experiment Upgrade and Improvement of Surface Diagnostic Capabilities at STAR Facility for Enhancing Tritium and Nuclear PMI Sciences

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shimada, M.; Taylor, C. N.; Pawelko, R. J.

2016-04-01

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 directly study tritium retention and permeation in neutron-irradiated materials with tritium [M. Shimada et.al., Rev. Sci. Instru. 82 (2011) 083503 and and M. Shimada, et.al., Nucl. Fusion 55 (2015) 013008]. The plasma-material-interaction (PMI) determines a boundary condition for diffusing tritium into bulk PFCs, and the tritium PMI is crucial for enhancing fundamental sciences that dictate tritium fuel cycles and safety and are high importance to an FNSF and DEMO. Recentlymore » the TPE has undergone major upgrades in its electrical and control systems. 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. We discuss the electrical upgrade, enhanced operational safety, improved plasma performance, and development of optical spectrometer system. 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, Fusion Nuclear Science Facility (FNSF), and Demonstration reactor (DEMO). This work was prepared for the U.S. Department of Energy, Office of Fusion Energy Sciences, under the DOE Idaho Field Office contract number DE-AC07-05ID14517.« less

DUCTILE-PHASE TOUGHENED TUNGSTEN FOR PLASMA-FACING MATERIALS IN FUSION REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henager, Charles H.; Setyawan, Wahyu; Roosendaal, Timothy J.

2017-05-01

Tungsten (W) and W-alloys are the leading candidates for plasma-facing components in nuclear fusion reactor designs because of their high melting point, strength retention at high temperatures, high thermal conductivity, and low sputtering yield. However, tungsten is brittle and does not exhibit the required fracture toughness for licensing in nuclear applications. A promising approach to increasing fracture toughness of W-alloys is by ductile-phase toughening (DPT). In this method, a ductile phase is included in a brittle matrix to prevent on inhibit crack propagation by crack blunting, crack bridging, crack deflection, and crack branching. Model examples of DPT tungsten are exploredmore » in this study, including W-Cu and W-Ni-Fe powder product composites. Three-point and four-point notched and/or pre-cracked bend samples were tested at several strain rates and temperatures to help understand deformation, cracking, and toughening in these materials. Data from these tests are used for developing and calibrating crack-bridging models. Finite element damage mechanics models are introduced as a modeling method that appears to capture the complexity of crack growth in these materials.« less

High heat flux testing of CFC composites for the tokamak physics experiment

NASA Astrophysics Data System (ADS)

Valentine, P. G.; Nygren, R. E.; Burns, R. W.; Rocket, P. D.; Colleraine, A. P.; Lederich, R. J.; Bradley, J. T.

1996-10-01

High heat flux (HHF) testing of carbon fiber reinforced carbon composites (CFC's) was conducted under the General Atomics program to develop plasma-facing components (PFC's) for Princeton Plasma Physics Laboratory's tokamak physics experiment (TPX). As part of the process of selecting TPX CFC materials, a series of HHF tests were conducted with the 30 kW electron beam test system (EBTS) facility at Sandia National Laboratories, and with the plasma disruption simulator I (PLADIS-I) facility at the University of New Mexico. The purpose of the tests was to make assessments of the thermal performance and erosion behavior of CFC materials. Tests were conducted with 42 different CFC materials. In general, the CFC materials withstood the rapid thermal pulse environments without fracturing, delaminating, or degrading in a non-uniform manner; significant differences in thermal performance, erosion behavior, vapor evolution, etc. were observed and preliminary findings are presented below. The CFC's exposed to the hydrogen plasma pulses in PLADIS-I exhibited greater erosion rates than the CFC materials exposed to the electron-beam pulses in EBTS. The results obtained support the continued consideration of a variety of CFC composites for TPX PFC components.

FOREWORD: 12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications 12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications

NASA Astrophysics Data System (ADS)

Kreter, Arkadi; Linke, Jochen; Rubel, Marek

2009-12-01

The 12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications (PFMC-12) was held in Forschungszentrum Jülich (FZJ) in Germany in May 2009. This symposium is the successor to the International Workshop on Carbon Materials for Fusion Applications series. Between 1985 and 2003, 10 'Carbon Workshops' were organized in Jülich, Stockholm and Hohenkammer. After this time, the scope of the symposium was redefined to reflect the new requirements of ITER and the ongoing evolution of the field. The workshop was first organized under its new name in 2006 in Greifswald, Germany. The main objective of this conference series is to provide a discussion forum for experts from research institutions and industry dealing with materials for plasma-facing components in present and future controlled fusion devices. The operation of ASDEX-Upgrade with tungsten-coated wall, the fast progress of the ITER-Like Wall Project at JET, the plans for the EAST tokamak to install tungsten, the start of ITER construction and a discussion about the wall material for DEMO all emphasize the importance of plasma-wall interactions and component behaviour, and give much momentum to the field. In this context, the properties and behaviour of beryllium, carbon and tungsten under plasma impact are research topics of foremost relevance and importance. Our community realizes both the enormous advantages and serious drawbacks of all the candidate materials. As a result, discussion is in progress as to whether to use carbon in ITER during the initial phase of operation or to abandon this element and use only metal components from the start. There is broad knowledge about carbon, both in terms of its excellent power-handling capabilities and the drawbacks related to chemical reactivity with fuel species and, as a consequence, about problems arising from fuel inventory and dust formation. We are learning continuously about beryllium and tungsten under fusion conditions, but our knowledge is still limited, especially in relation to the behaviour of these metals in environments containing multiple species. There are many appealing issues related to material mixing and fuel retention that call for robust and comprehensive studies. In this sense, the aim of the workshop is not only to discuss hot topics, but also to identify the most important research areas and those that need urgent solutions. Another topic of foremost relevance to ITER is the development of plasma-facing components that are able to withstand extreme power fluxes, in particular, those during transient phases. Materials and production methods for high-heat-flux components have to be further developed and industrialized. A key requirement in this field is the development of non-destructive testing methods for the qualification of methods and quality assessment during production. Invited talks and contributed presentations therefore dealt with aspects of fundamental processes, experimental findings, advanced modelling and the technology of fusion reactor components. Several areas were selected as the major topics of PFMC-12: materials for the ITER-divertor (erosion, redeposition, fuel retention) carbon-based materials tungsten and tungsten coatings beryllium mixed materials (intentional and non-intentional) the ITER-Like Wall Project materials under high-heat-flux loads including transients (ELMs, disruptions) technology and testing of plasma-facing components neutron effects in plasma-facing materials. 26 invited lectures and oral contributions, and 131 posters were presented by participants from research laboratories and industrial companies. 210 researchers from 24 countries from all over the world participated in a lively and intense exchange of knowledge and ideas. The workshop was hosted by Forschungszentrum Jülich (FZJ), a centre where the integration of science and technology for fusion reactor materials has been a focus for decades. This is reflected by the operation of several devices vital for progress in fusion research. TEXTOR (Toroidal EXperiment for Technology Oriented Research) is a mission-oriented tokamak for the study of plasma-wall interactions and testing of materials in fusion environments. JUDITH-1 (JÜlich DIvertor Test facility in Hot-cell) and the recently started JUDITH-2 are the most powerful test beds for studies of material performance under steady-state or pulsed power loads. The results of testing in JUDITH establish the background for material qualification. The expertize of FZJ in fusion engineering is vital for the construction of the Wendelstein-7X stellarator in Greifswald and the diagnostics for the ITER plasma. Finally, there is a group of eminent theoreticians and modellers at work in FZJ. As a consequence, FZJ is the home of the supercomputer, High Performance Computing-For Fusion (HPC-FF). During the workshop, special guided laboratory tours were organized to get the participants acquainted with the experimental facilities at FZJ: TEXTOR, JUDITH and HPC-FF. The quality of the talks, posters and discussions, and the comfortable conference facilities were of great importance but activities outside fusion science also formed part of the workshop. A guided tour in the Old Town of Aachen was very much appreciated by all participants; a stroll in this beautiful place was not only a relaxing moment but also put participants in touch with a great deal of European history. Big and long-term projects always attract young, ambitious people. The recruitment of talented scientists is a conditio sine qua non for the future success and progress of fusion science and engineering. The enthusiasm of students is very important but not sufficient; it is the responsibility of older colleagues to get students acquainted with the major issues and challenges. For this reason, the workshop was preceded by a series of tutorials on plasma-wall interactions and properties, and testing of relevant materials. The lectures were met with a great response: not only did over thirty young colleagues register but also senior scientists registered for the course and were very active in discussions. The workshop was supported financially by Forschungszentrum Jülich and the ExtreMat Integrated Project, a programme for the development and study of new materials for extreme environments. We are very grateful to the staff of Forschungszentrum who helped with the organization. Our most cordial thanks and gratitude go to Yasmin Fattah, Angelika Hallmanns, Gabriele Knauf and Gerd Boeling for all their kindness and efficiency, which helped all of us to enjoy the meeting. We thank most sincerely our colleagues Gerald Pintsuk, Takeshi Hirai and Andrey Litnovsky for their most professional work in the construction and operation of the conference webpage, the preparation of the sessions and for all other elements that were vital for the smooth running of the meeting. We thank very much Marliese Felden and Ralf-Uwe Limbach who very kindly and professionally took care of the photographic documentation of the workshop. The proceedings of this workshop contains 67 peer-reviewed articles covering the contents of most of the invited presentations and a number of poster contributions which were pre-selected by the programme committee. The papers reflect the development and actual status of the field. We thank all participants for their contributions and the referees for their smooth and efficient peer-review. Thank you all for your hard work and co-operation. We are looking forward to seeing you at the next meeting; we invite you to come, though we are not yet able to say 'when' and 'where' we will meet next time. It is a special feature of this conference series that a new meeting is announced only when the community feels that there is substantial new material to be presented and discussed.

A multiscale microstructural approach to ductile-phase toughened tungsten for plasma-facing materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nguyen, Ba Nghiep; Henager, Jr., Charles H.; Overman, Nicole R.

Increasing fracture toughness and modifying the ductile-brittle transition temperature of a tungsten-alloy relative to pure tungsten has been shown to be feasible by ductile-phase toughening (DPT) of tungsten for future plasma-facing materials for fusion energy. In DPT, a ductile phase is included in a brittle tungsten matrix to increase the overall work of fracture for the material. This research models the deformation behavior of DPT tungsten materials, such as tungsten-copper composites, using a multiscale modeling approach that involves a microstructural dual-phase (copper-tungsten) region of interest where the constituent phases are finely discretized and are described by a continuum damage mechanicsmore » model. Large deformation, damage, and fracture are allowed to occur and are modeled in a region that is connected to adjacent homogenized elastic regions to form a macroscopic structure, such as a test specimen. The present paper illustrates this multiscale modeling approach to analyze unnotched and single-edge notched (SENB) tungsten-copper composite specimens subjected to three-point bending. The predicted load-displacement responses and crack propagation patterns are compared to the corresponding experimental results to validate the model. Furthermore, such models may help design future DPT composite configurations for fusion materials, including volume fractions of ductile phase and microstructural optimization.« less

A multiscale microstructural approach to ductile-phase toughened tungsten for plasma-facing materials

DOE PAGES

Nguyen, Ba Nghiep; Henager, Jr., Charles H.; Overman, Nicole R.; ...

2018-05-23

Increasing fracture toughness and modifying the ductile-brittle transition temperature of a tungsten-alloy relative to pure tungsten has been shown to be feasible by ductile-phase toughening (DPT) of tungsten for future plasma-facing materials for fusion energy. In DPT, a ductile phase is included in a brittle tungsten matrix to increase the overall work of fracture for the material. This research models the deformation behavior of DPT tungsten materials, such as tungsten-copper composites, using a multiscale modeling approach that involves a microstructural dual-phase (copper-tungsten) region of interest where the constituent phases are finely discretized and are described by a continuum damage mechanicsmore » model. Large deformation, damage, and fracture are allowed to occur and are modeled in a region that is connected to adjacent homogenized elastic regions to form a macroscopic structure, such as a test specimen. The present paper illustrates this multiscale modeling approach to analyze unnotched and single-edge notched (SENB) tungsten-copper composite specimens subjected to three-point bending. The predicted load-displacement responses and crack propagation patterns are compared to the corresponding experimental results to validate the model. Furthermore, such models may help design future DPT composite configurations for fusion materials, including volume fractions of ductile phase and microstructural optimization.« less

Effects of in situ dual ion beam (He+ and D+) irradiation with simultaneous pulsed heat loading on surface morphology evolution of tungsten-tantalum alloys

NASA Astrophysics Data System (ADS)

Gonderman, S.; Tripathi, J. K.; Sinclair, G.; Novakowski, T. J.; Sizyuk, T.; Hassanein, A.

2018-02-01

The strong thermal and mechanical properties of tungsten (W) are well suited for the harsh fusion environment. However, increasing interest in using tungsten as plasma-facing components (PFCs) has revealed several key issues. These potential roadblocks necessitate more investigation of W and other alternative W based materials exposed to realistic fusion conditions. In this work, W and tungsten-tantalum (W-Ta) alloys were exposed to single (He+) and dual (He+  +  D+) ion irradiations with simultaneous pulsed heat loading to elucidate PFCs response under more realistic conditions. Laser only exposer revealed significantly more damage in W-Ta samples as compared to pure W samples. This was due to the difference in the mechanical properties of the two different materials. Further erosion studies were conducted to evaluate the material degradation due to transient heat loading in both the presence and absence of He+ and/or D+ ions. We concluded that erosion of PFC materials was significantly enhanced due to the presence of ion irradiation. This is important as it demonstrates that there are key synergistic effects resulting from more realistic fusion loading conditions that need to be considered when evaluating the response of plasma facing materials.

Erosion products of plasma facing materials formed under ITER-like transient load and deuterium retention in them

DOE Office of Scientific and Technical Information (OSTI.GOV)

Putrik, A. B., E-mail: putrik@triniti.ru; Klimov, N. S.; Gasparyan, Yu. M., E-mail: yura@plasma.mephi.ru

2015-12-15

Erosion of the plasma-facing materials in particular evaporation of the materials in a fusion reactor under intense transient events is one of the problems of the ITER. The current experimental data are insufficient to predict the properties of the erosion products, a significant part of which will be formed during transient events (edge-localized modes (ELMs) and disruptions). The paper concerns the experimental investigation of the graphite and tungsten erosion products deposited under pulsed plasma load at the QSPA-T: heat load on the target was 2.6 MJ/m{sup 2} with 0.5 ms pulse duration. The designed diagnostics for measuring the deposition ratemore » made it possible to determine that the deposition of eroded material occurs during discharge, and the deposition rate is in the range (0.1–100) × 10{sup 19} at/(cm{sup 2} s), which is much higher than that for stationary processes. It is found that the relative atomic concentrations D/C and D/(W + C) in the erosion products deposited during the pulse process are on the same level as for the stationary processes. An exposure of erosion products to photonic energy densities typical of those expected at mitigated disruptions in the ITER (pulse duration of 0.5–1 ms, integral energy density of radiation of 0.1–0.5 MJ/m2) significantly decreases the concentration of trapped deuterium.« less

Advanced simulation of mixed-material erosion/evolution and application to low and high-Z containing plasma facing components

NASA Astrophysics Data System (ADS)

Brooks, J. N.; Hassanein, A.; Sizyuk, T.

2013-07-01

Plasma interactions with mixed-material surfaces are being analyzed using advanced modeling of time-dependent surface evolution/erosion. Simulations use the REDEP/WBC erosion/redeposition code package coupled to the HEIGHTS package ITMC-DYN mixed-material formation/response code, with plasma parameter input from codes and data. We report here on analysis for a DIII-D Mo/C containing tokamak divertor. A DIII-D/DiMES probe experiment simulation predicts that sputtered molybdenum from a 1 cm diameter central spot quickly saturates (˜4 s) in the 5 cm diameter surrounding carbon probe surface, with subsequent re-sputtering and transport to off-probe divertor regions, and with high (˜50%) redeposition on the Mo spot. Predicted Mo content in the carbon agrees well with post-exposure probe data. We discuss implications and mixed-material analysis issues for Be/W mixing at the ITER outer divertor, and Li, C, Mo mixing at an NSTX divertor.

Modeling of material erosion and redeposition for dedicated DiMES experiments on DIII-D

NASA Astrophysics Data System (ADS)

Ding, R.; Abrams, T.; Chrobak, C. P.; Guo, H. Y.; Snyder, P. B.; Chan, V. S.; Rudakov, D. L.; Stangeby, P. C.; Elder, J. D.; Tskhakaya, D.; Wampler, W. R.; Kirschner, A.; McLean, A. G.

2015-11-01

Erosion and redeposition of plasma facing materials is a key issue for high-power, long pulse tokamak operation. A series of experiments has been carried out on DIII-D in which well-characterized samples of different materials were exposed to divertor plasma using DiMES. Such experiments provide a good benchmark for PMI codes, such as ERO. It was found that the erosion and redeposition are strongly determined by the impurity content in the plasma and sheath properties near the surface. The principal experimental results (net erosion rate and profile, net/gross erosion ratio) are reproduced by ERO simulations to within the uncertainties, indicating that the controlling physics has likely been identified. New techniques suggested by modeling such as external biasing and local gas injection for suppressing material erosion are planned to be tested in DiMES/DIII-D experiments. Work supported by US DOE DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC04-94AL85000, DE-AC52-07NA27344.

Development of a plasma driven permeation experiment for TPE

DOE PAGES

Buchenauer, Dean; Kolasinski, Robert; Shimada, Masa; ...

2014-04-18

Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800 ?C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1,2]. To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000 ?C, a new TPE membrane holder has been built to hold test specimens (=1 mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ionmore » chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE’s vacuum chamber has been demonstrated by sealing tests performed up to 1000 ?C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (~700 Pa) with good flow uniformity (at 1000 sccm). Thermal transport simulations indicate that temperatures up to 1000 ?C are expected at the highest TPE fluxes.« less

An in situ accelerator-based diagnostic for plasma-material interactions science on magnetic fusion devices.

PubMed

Hartwig, Zachary S; Barnard, Harold S; Lanza, Richard C; Sorbom, Brandon N; Stahle, Peter W; Whyte, Dennis G

2013-12-01

This paper presents a novel particle accelerator-based diagnostic that nondestructively measures the evolution of material surface compositions inside magnetic fusion devices. The diagnostic's purpose is to contribute to an integrated understanding of plasma-material interactions in magnetic fusion, which is severely hindered by a dearth of in situ material surface diagnosis. The diagnostic aims to remotely generate isotopic concentration maps on a plasma shot-to-shot timescale that cover a large fraction of the plasma-facing surface inside of a magnetic fusion device without the need for vacuum breaks or physical access to the material surfaces. Our instrument uses a compact (~1 m), high-current (~1 milliamp) radio-frequency quadrupole accelerator to inject 0.9 MeV deuterons into the Alcator C-Mod tokamak at MIT. We control the tokamak magnetic fields--in between plasma shots--to steer the deuterons to material surfaces where the deuterons cause high-Q nuclear reactions with low-Z isotopes ~5 μm into the material. The induced neutrons and gamma rays are measured with scintillation detectors; energy spectra analysis provides quantitative reconstruction of surface compositions. An overview of the diagnostic technique, known as accelerator-based in situ materials surveillance (AIMS), and the first AIMS diagnostic on the Alcator C-Mod tokamak is given. Experimental validation is shown to demonstrate that an optimized deuteron beam is injected into the tokamak, that low-Z isotopes such as deuterium and boron can be quantified on the material surfaces, and that magnetic steering provides access to different measurement locations. The first AIMS analysis, which measures the relative change in deuterium at a single surface location at the end of the Alcator C-Mod FY2012 plasma campaign, is also presented.

Can tokamaks PFC survive a single event of any plasma instabilities?

NASA Astrophysics Data System (ADS)

Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

2013-07-01

Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

NASA Astrophysics Data System (ADS)

Guo, H. Y.; Hill, D. N.; Leonard, A. W.; Allen, S. L.; Stangeby, P. C.; Thomas, D.; Unterberg, E. A.; Abrams, T.; Boedo, J.; Briesemeister, A. R.; Buchenauer, D.; Bykov, I.; Canik, J. M.; Chrobak, C.; Covele, B.; Ding, R.; Doerner, R.; Donovan, D.; Du, H.; Elder, D.; Eldon, D.; Lasa, A.; Groth, M.; Guterl, J.; Jarvinen, A.; Hinson, E.; Kolemen, E.; Lasnier, C. J.; Lore, J.; Makowski, M. A.; McLean, A.; Meyer, B.; Moser, A. L.; Nygren, R.; Owen, L.; Petrie, T. W.; Porter, G. D.; Rognlien, T. D.; Rudakov, D.; Sang, C. F.; Samuell, C.; Si, H.; Schmitz, O.; Sontag, A.; Soukhanovskii, V.; Wampler, W.; Wang, H.; Watkins, J. G.

2016-12-01

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

DOE PAGES

Guo, H. Y.; Hill, D. N.; Leonard, A. W.; ...

2016-09-14

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, whichmore » we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.« less

High-heat-flux testing of irradiated tungsten-based materials for fusion applications using infrared plasma arc lamps

DOE PAGES

Sabau, Adrian S.; Ohriner, Evan K.; Kiggans, Jim; ...

2014-11-01

Testing of advanced materials and component mock-ups under prototypical fusion high-heat-flux conditions, while historically a mainstay of fusion research, has proved to be quite challenging, especially for irradiated materials. A new high-heat-flux–testing (HHFT) facility based on water-wall plasma arc lamps (PALs) is now introduced for materials and small-component testing. Two PAL systems, utilizing a 12 000°C plasma arc contained in a quartz tube cooled by a spiral water flow over the inside tube surface, provide maximum incident heat fluxes of 4.2 and 27 MW/m 2 over areas of 9×12 and 1×10 cm 2, respectively. This paper will present the overallmore » design and implementation of a PAL-based irradiated material target station (IMTS). The IMTS is primarily designed for testing the effects of heat flux or thermal cycling on material coupons of interest, such as those for plasma-facing components. Temperature results are shown for thermal cycling under HHFT of tungsten coupon specimens that were neutron irradiated in HFIR. Finally, radiological surveys indicated minimal contamination of the 36×36×18 cm test section, demonstrating the capability of the new facility to handle irradiated specimens at high temperature.« less

GEM detector development for tokamak plasma radiation diagnostics: SXR poloidal tomography

NASA Astrophysics Data System (ADS)

Chernyshova, Maryna; Malinowski, Karol; Ziółkowski, Adam; Kowalska-Strzeciwilk, Ewa; Czarski, Tomasz; Poźniak, Krzysztof T.; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Wojeński, Andrzej; Kolasiński, Piotr; Krawczyk, Rafał D.

2015-09-01

An increased attention to tungsten material is related to a fact that it became a main candidate for the plasma facing material in ITER and future fusion reactor. The proposed work refers to the studies of W influence on the plasma performances by developing new detectors based on Gas Electron Multiplier GEM) technology for tomographic studies of tungsten transport in ITER-oriented tokamaks, e.g. WEST project. It presents current stage of design and developing of cylindrically bent SXR GEM detector construction for horizontal port implementation. Concept to overcome an influence of constraints on vertical port has been also presented. It is expected that the detecting unit under development, when implemented, will add to the safe operation of tokamak bringing creation of sustainable nuclear fusion reactors a step closer.

Development progress of the Materials Analysis and Particle Probe

NASA Astrophysics Data System (ADS)

Lucia, M.; Kaita, R.; Majeski, R.; Bedoya, F.; Allain, J. P.; Boyle, D. P.; Schmitt, J. C.; Onge, D. A. St.

2014-11-01

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.

Development progress of the Materials Analysis and Particle Probe.

PubMed

Lucia, M; Kaita, R; Majeski, R; Bedoya, F; Allain, J P; Boyle, D P; Schmitt, J C; Onge, D A St

2014-11-01

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.

Measurement of thickness of film deposited on the plasma-facing wall in the QUEST tokamak by colorimetry.

PubMed

Wang, Z; Hanada, K; Yoshida, N; Shimoji, T; Miyamoto, M; Oya, Y; Zushi, H; Idei, H; Nakamura, K; Fujisawa, A; Nagashima, Y; Hasegawa, M; Kawasaki, S; Higashijima, A; Nakashima, H; Nagata, T; Kawaguchi, A; Fujiwara, T; Araki, K; Mitarai, O; Fukuyama, A; Takase, Y; Matsumoto, K

2017-09-01

After several experimental campaigns in the Kyushu University Experiment with Steady-state Spherical Tokamak (QUEST), the originally stainless steel plasma-facing wall (PFW) becomes completely covered with a deposited film composed of mixture materials, such as iron, chromium, carbon, and tungsten. In this work, an innovative colorimetry-based method was developed to measure the thickness of the deposited film on the actual QUEST wall. Because the optical constants of the deposited film on the PFW were position-dependent and the extinction coefficient k 1 was about 1.0-2.0, which made the probing light not penetrate through some thick deposited films, the colorimetry method developed can only provide a rough value range of thickness of the metal-containing film deposited on the actual PFW in QUEST. However, the use of colorimetry is of great benefit to large-area inspections and to radioactive materials in future fusion devices that will be strictly prohibited from being taken out of the limited area.

Lithium-based surfaces controlling fusion plasma behavior at the plasma-material interfacea)

NASA Astrophysics Data System (ADS)

Allain, Jean Paul; Taylor, Chase N.

2012-05-01

The plasma-material interface and its impact on the performance of magnetically confined thermonuclear fusion plasmas are considered to be one of the key scientific gaps in the realization of nuclear fusion power. At this interface, high particle and heat flux from the fusion plasma can limit the material's lifetime and reliability and therefore hinder operation of the fusion device. Lithium-based surfaces are now being used in major magnetic confinement fusion devices and have observed profound effects on plasma performance including enhanced confinement, suppression and control of edge localized modes (ELM), lower hydrogen recycling and impurity suppression. The critical spatial scale length of deuterium and helium particle interactions in lithium ranges between 5-100 nm depending on the incident particle energies at the edge and magnetic configuration. Lithium-based surfaces also range from liquid state to solid lithium coatings on a variety of substrates (e.g., graphite, stainless steel, refractory metal W/Mo/etc., or porous metal structures). Temperature-dependent effects from lithium-based surfaces as plasma facing components (PFC) include magnetohydrodynamic (MHD) instability issues related to liquid lithium, surface impurity, and deuterium retention issues, and anomalous physical sputtering increase at temperatures above lithium's melting point. The paper discusses the viability of lithium-based surfaces in future burning-plasma environments such as those found in ITER and DEMO-like fusion reactor devices.

Development of a double plasma gun device for investigation of effects of vapor shielding on erosion of PFC materials under ELM-like pulsed plasma bombardment

NASA Astrophysics Data System (ADS)

Sakuma, I.; Iwamoto, D.; Kitagawa, Y.; Kikuchi, Y.; Fukumoto, N.; Nagata, M.

2012-10-01

It is considered that thermal transient events such as type I edge localized modes (ELMs) could limit the lifetime of plasma-facing components (PFCs) in ITER. We have investigated surface damage of tungsten (W) materials under transient heat and particle loads by using a magnetized coaxial plasma gun (MCPG) device at University of Hyogo. The capacitor bank energy for the plasma discharge is 144 kJ (2.88 mF, 10 kVmax). Surface melting of a W material was clearly observed at the energy density of ˜2 MJ/m2. It is known that surface melting and evaporation during a transient heat load could generate a vapor cloud layer in front of the target material [1]. Then, the subsequent erosion could be reduced by the vapor shielding effect. In this study, we introduce a new experiment using two MCPG devices (MCPG-1, 2) to understand vapor shielding effects of a W surface under ELM-like pulsed plasma bombardment. The capacitor bank energy of MCPG-2 is almost same as that of MCPG-1. The second plasmoid is applied with a variable delay time after the plasmoid produced by MCPG-1. Then, a vapor cloud layer could shield the second plasma load. To verify the vapor shielding effects, surface damage of a W material is investigated by changing the delay time. In the conference, the preliminary experimental results will be shown.[4pt] [1] A. Hassanein et al., J. Nucl. Mater. 390-391, pp. 777-780 (2009).

Response of plasma facing components in Tokamaks due to intense energy deposition using Particle-In-Cell (PIC) methods

NASA Astrophysics Data System (ADS)

Genco, Filippo

Damage to plasma-facing components (PFC) due to various plasma instabilities is still a major concern for the successful development of fusion energy and represents a significant research obstacle in the community. It is of great importance to fully understand the behavior and lifetime expectancy of PFC under both low energy cycles during normal events and highly energetic events as disruptions, Edge-Localized Modes (ELM), Vertical Displacement Events (VDE), and Run-away electron (RE). The consequences of these high energetic dumps with energy fluxes ranging from 10 MJ/m2 up to 200 MJ/m 2 applied in very short periods (0.1 to 5 ms) can be catastrophic both for safety and economic reasons. Those phenomena can cause a) large temperature increase in the target material b) consequent melting, evaporation and erosion losses due to the extremely high heat fluxes c) possible structural damage and permanent degradation of the entire bulk material with probable burnout of the coolant tubes; d) plasma contamination, transport of target material into the chamber far from where it was originally picked. The modeling of off-normal events such as Disruptions and ELMs requires the simultaneous solution of three main problems along time: a) the heat transfer in the plasma facing component b) the interaction of the produced vapor from the surface with the incoming plasma particles c) the transport of the radiation produced in the vapor-plasma cloud. In addition the moving boundaries problem has to be considered and solved at the material surface. Considering the carbon divertor as target, the moving boundaries are two since for the given conditions, carbon doesn't melt: the plasma front and the moving eroded material surface. The current solution methods for this problem use finite differences and moving coordinates system based on the Crank-Nicholson method and Alternating Directions Implicit Method (ADI). Currently Particle-In-Cell (PIC) methods are widely used for solving complex dynamics problems involving distorted plasma hydrodynamic problems and plasma physics. The PIC method solves the hydrodynamic equations solving all field equations tracking at the same time "sample particles" or pseudo-particles (representative of the much more numerous real ones) as the move under the influence of diffusion or magnetic force. The superior behavior of the PIC techniques over the more classical Lagrangian finite difference methods stands in the fact that detailed information about the particles are available at all times as well as mass and momentum transport values are constantly provided. This allows with a relative small number of particles to well describe the behavior of plasma even in presence of highly distorted flows without losing accuracy. The radiation transport equation is solved at each time step calculating for each cell the opacity and emissivity coefficients. Photon radiation continuum and line fluxes are also calculated per the entire domain and provide useful information for the entire energetic calculation of the system which in the end provides the total values of erosion and lifetime of the target material. In this thesis, a new code named HEIGHTS-PIC code has been created and modified using a new approach of the PIC technique to solve the three physics problems involved integrating each of them as a continuum providing insight on the plasma behavior, evolution along time and physical understanding of the very complex phenomena taking place. The results produced with the models are compared with the well-known and benchmarked HEIGHTS package and also with existing experimental results especially produced in Russia at the TRINITI facility. Comparisons with LASER experiments are also discussed.

Dust particles in controlled fusion devices: morphology, observations in the plasma and influence on the plasma performance

NASA Astrophysics Data System (ADS)

Rubel, M.; Cecconello, M.; Malmberg, J. A.; Sergienko, G.; Biel, W.; Drake, J. R.; Hedqvist, A.; Huber, A.; Philipps, V.

2001-08-01

The formation and release of particle agglomerates, i.e. debris and dusty objects, from plasma facing components and the impact of such materials on plasma operation in controlled fusion devices has been studied in the Extrap T2 reversed field pinch and the TEXTOR tokamak. Several plasma diagnostic techniques, camera observations and surface analysis methods were applied for in situ and ex situ investigation. The results are discussed in terms of processes that are decisive for dust transfer: localized power deposition connected with wall locked modes causing emission of carbon granules, brittle destruction of graphite and detachment of thick flaking co-deposited layers. The consequences for large next step devices are also addressed.

Modeling of surface temperature effects on mixed material migration in NSTX-U

NASA Astrophysics Data System (ADS)

Nichols, J. H.; Jaworski, M. A.; Schmid, K.

2016-10-01

NSTX-U will initially operate with graphite walls, periodically coated with thin lithium films to improve plasma performance. However, the spatial and temporal evolution of these films during and after plasma exposure is poorly understood. The WallDYN global mixed-material surface evolution model has recently been applied to the NSTX-U geometry to simulate the evolution of poloidally inhomogenous mixed C/Li/O plasma-facing surfaces. The WallDYN model couples local erosion and deposition processes with plasma impurity transport in a non-iterative, self-consistent manner that maintains overall material balance. Temperature-dependent sputtering of lithium has been added to WallDYN, utilizing an adatom sputtering model developed from test stand experimental data. Additionally, a simplified temperature-dependent diffusion model has been added to WallDYN so as to capture the intercalation of lithium into a graphite bulk matrix. The sensitivity of global lithium migration patterns to changes in surface temperature magnitude and distribution will be examined. The effect of intra-discharge increases in surface temperature due to plasma heating, such as those observed during NSTX Liquid Lithium Divertor experiments, will also be examined. Work supported by US DOE contract DE-AC02-09CH11466.

RF models for plasma-surface interactions

NASA Astrophysics Data System (ADS)

Jenkins, Thomas; Smithe, David; Lin, Ming-Chieh; Kruger, Scott; Stoltz, Peter

2013-09-01

Computational models for DC and oscillatory (RF-driven) sheath potentials, arising at metal or dielectric-coated surfaces in contact with plasma, are developed within the VSim code and applied in parameter regimes characteristic of fusion plasma experiments and plasma processing scenarios. Results from initial studies quantifying the effects of various dielectric wall coating materials and thicknesses on these sheath potentials, as well as on the ensuing flux of plasma particles to the wall, are presented. As well, the developed models are used to model plasma-facing ICRF antenna structures in the ITER device; we present initial assessments of the efficacy of dielectric-coated antenna surfaces in reducing sputtering-induced high-Z impurity contamination of the fusion reaction. Funded by U.S. DoE via a Phase I SBIR grant, award DE-SC0009501.

Tungsten coating by ATC plasma spraying on CFC for WEST tokamak

NASA Astrophysics Data System (ADS)

Firdaouss, M.; Desgranges, C.; Hernandez, C.; Mateus, C.; Maier, H.; Böswirth, B.; Greuner, H.; Samaille, F.; Bucalossi, J.; Missirlian, M.

2017-12-01

In the field of fusion experiments using a tokamak, the plasma facing components (PFC) are the closest object to the hot plasma. Due to the plasma-wall interaction, the material composing the PFC may enter the plasma and disturb the experiments. In the past, the main material for PFC was carbon (CFC, graphite), while the future reactors like ITER will be fully metallic, in particular tungsten. The Tore Supra tokamak has been transformed in an x-point divertor fusion device within the frame of the WEST (W (tungsten) Environment in Steady-state Tokamak) project in order to have plasma conditions close to those expected in ITER. The PFC other than the divertor has been coated with W to transform Tore Supra into a fully metallic environment. Different coating techniques have been selected for different kind of PFC. This paper gives an overview on the coating process used for the antennae protection limiter, the associated validation programme and concludes on the adequacy of the W coating with the WEST experimental programme requirements and gives perspectives on the development to be pursued.

Investigation Of A Tin-Lithium Alloy As A Liquid Plasma-Facing Material

NASA Astrophysics Data System (ADS)

Sandefur, Heather; Ruzic, David; Kolasinski, Robert; Buchenauer, Dean; Sandia National Laboratories Collaboration; University of Illinois Collaboration

2017-10-01

Sn-Li is a low melting-point alloy that has been identified as a material with favorable performance in plasma material interaction studies. While lithium is a low Z material with a demonstrated ability to absorb impinging ions, pure lithium is plagued by high evaporation rates in the liquid phase. The Sn-Li alloy is a more stable alternative that provides a lower rate of evaporative flux due to the high vapor pressure of tin. In the liquid phase, the bulk segregation of lithium to the surface of the material has also been observed. While the alloy is of considerable interest, little data has been collected on its surface chemistry in a plasma environment. In order to expand the existing body of knowledge in this area, samples of an 80 percent Sn-20 percent Li alloy were prepared and analyzed in order to assess the surface composition and degree of lithium segregation in the liquid phase. The Angle-Resolved Ion Energy Spectrometer (ARIES) at Sandia National Laboratories was used to probe the surfaces of the alloy using the low energy ion scattering method. The lithium coverage at the surface was measured, and the material's affinity for hydrogen chemisorption was investigated.

Overview of innovative PMI research on NSTX-U and associated PMI facilities at PPPL

DOE PAGES

M. Ono; Jaworski, M.; Kaita, R.; ...

2013-05-01

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 NSTX-U, the PMI research has received a strong emphasis. Moreover, with ˜15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m 2.

Liquid metals as a divertor plasma-facing material explored using the Pilot-PSI and Magnum-PSI linear devices

NASA Astrophysics Data System (ADS)

Morgan, T. W.; Rindt, P.; van Eden, G. G.; Kvon, V.; Jaworksi, M. A.; Lopes Cardozo, N. J.

2018-01-01

For DEMO and beyond, liquid metal plasma-facing components are considered due to their resilience to erosion through flowed replacement, potential for cooling beyond conduction and inherent immunity to many of the issues of neutron loading compared to solid materials. The development curve of liquid metals is behind that of e.g. tungsten however, and tokamak-based research is currently somewhat limited in scope. Therefore, investigation into linear plasma devices can provide faster progress under controlled and well-diagnosed conditions in assessing many of the issues surrounding the use of liquid metals. The linear plasma devices Magnum-PSI and Pilot-PSI are capable of producing DEMO-relevant plasma fluxes, which well replicate expected divertor conditions, and the exploration of physics issues for tin (Sn) and lithium (Li) such as vapour shielding, erosion under high particle flux loading and overall power handling are reviewed here. A deeper understanding of erosion and deposition through this work indicates that stannane formation may play an important role in enhancing Sn erosion, while on the other hand the strong hydrogen isotope affinity reduces the evaporation rate and sputtering yields for Li. In combination with the strong redeposition rates, which have been observed under this type of high-density plasma, this implies that an increase in the operational temperature range, implying a power handling range of 20-25 MW m-2 for Sn and up to 12.5 MW m-2 for Li could be achieved. Vapour shielding may be expected to act as a self-protection mechanism in reducing the heat load to the substrate for off-normal events in the case of Sn, but may potentially be a continual mode of operation for Li.

Secondary electron emission yield from high aspect ratio carbon velvet surfaces

DOE PAGES

Jin, Chenggang; Ottaviano, Angelica; Raitses, Yevgeny

2017-11-01

The plasma electrons bombarding a plasma-facing wall surface can induce secondary electron emission (SEE) from the wall. A strong SEE can enhance the power losses by reducing the wall sheath potential and thereby increasing the electron flux from the plasma to the wall. The use of the materials with surface roughness and the engineered materials with surface architecture is known to reduce the effective SEE by trapping the secondary electrons. In this work, we demonstrate a 65% reduction of SEE yield using a velvet material consisting of high aspect ratio carbon fibers. The measurements of SEE yield for different velvetmore » samples using the electron beam in vacuum demonstrate the dependence of the SEE yield on the fiber length and the packing density, which is strongly affected by the alignment of long velvet fibers with respect to the electron beam impinging on the velvet sample. Furthermore, the results of SEE measurements support the previous observations of the reduced SEE measured in Hall thrusters.« less

Secondary electron emission yield from high aspect ratio carbon velvet surfaces

NASA Astrophysics Data System (ADS)

Jin, Chenggang; Ottaviano, Angelica; Raitses, Yevgeny

2017-11-01

The plasma electrons bombarding a plasma-facing wall surface can induce secondary electron emission (SEE) from the wall. A strong SEE can enhance the power losses by reducing the wall sheath potential and thereby increasing the electron flux from the plasma to the wall. The use of the materials with surface roughness and the engineered materials with surface architecture is known to reduce the effective SEE by trapping the secondary electrons. In this work, we demonstrate a 65% reduction of SEE yield using a velvet material consisting of high aspect ratio carbon fibers. The measurements of SEE yield for different velvet samples using the electron beam in vacuum demonstrate the dependence of the SEE yield on the fiber length and the packing density, which is strongly affected by the alignment of long velvet fibers with respect to the electron beam impinging on the velvet sample. The results of SEE measurements support the previous observations of the reduced SEE measured in Hall thrusters.

Study of plasma-facing components in the Lithium Tokamak Experiment with the Materials Analysis and Particle Probe

NASA Astrophysics Data System (ADS)

Lucia, M.; Kaita, R.; Majeski, R.; Boyle, D. P.; Granstedt, E. M.; Jacobson, C. M.; Schmitt, J. C.; Allain, J. P.; Bedoya, F.; Gonderman, S.

2013-10-01

The Lithium Tokamak Experiment (LTX) is a spherical torus designed to accommodate solid or liquid lithium as the primary plasma-facing component (PFC). We present initial results from the implementation on LTX of the Materials Analysis and Particle Probe (MAPP) diagnostic, a collaboration among PPPL, Purdue University, and the University of Illinois. MAPP is a compact in vacuo surface science diagnostic, and its operation on LTX will provide the first ever in situ surface measurements of a tokamak first wall environment. With MAPP's analysis techniques, we will study the evolution of the surface chemistry of LTX's first wall as a function of varied temperature and lithium coating. During its 2013 run campaign, LTX will use an electron beam to evaporate lithium onto the first wall from an in-vessel reservoir. We will use two quartz crystal microbalances to estimate thickness of lithium coatings thus applied to the MAPP probe. We have recently installed a set of triple Langmuir probes on LTX, and they will be used to relate LTX edge plasma parameters to MAPP results. We will combine data from MAPP and the triple probes to estimate the local edge recycling coefficient based on desorption of retained hydrogen. This work was supported by U.S. DOE contract DE-AC02-09CH11466.

Simulation of tokamak armour erosion and plasma contamination at intense transient heat fluxes in ITER

NASA Astrophysics Data System (ADS)

Landman, I. S.; Bazylev, B. N.; Garkusha, I. E.; Loarte, A.; Pestchanyi, S. E.; Safronov, V. M.

2005-03-01

For ITER, the potential material damage of plasma facing tungsten-, CFC-, or beryllium components during transient processes such as ELMs or mitigated disruptions are simulated numerically using the MHD code FOREV-2D and the melt motion code MEMOS-1.5D for a heat deposition in the range of 0.5-3 MJ/m 2 on the time scale of 0.1-1 ms. Such loads can cause significant evaporation at the target surface and a contamination of the SOL by the ions of evaporated material. Results are presented on carbon plasma dynamics in toroidal geometry and on radiation fluxes from the SOL carbon ions obtained with FOREV-2D. The validation of MEMOS-1.5D against the plasma gun tokamak simulators MK-200UG and QSPA-Kh50, based on the tungsten melting threshold, is described. Simulations with MEMOS-1.5D for a beryllium first wall that provide important details about the melt motion dynamics and typical features of the damage are reported.

Analysis of Helium Segregation on Surfaces of Plasma-Exposed Tungsten

NASA Astrophysics Data System (ADS)

Maroudas, Dimitrios; Hu, Lin; Hammond, Karl; Wirth, Brian

2015-11-01

We report a systematic theoretical and atomic-scale computational study of implanted helium segregation on surfaces of tungsten, which is considered as a plasma facing component in nuclear fusion reactors. We employ a hierarchy of atomic-scale simulations, including molecular statics to understand the origin of helium surface segregation, targeted molecular-dynamics (MD) simulations of near-surface cluster reactions, and large-scale MD simulations of implanted helium evolution in plasma-exposed tungsten. We find that small, mobile helium clusters (of 1-7 He atoms) in the near-surface region are attracted to the surface due to an elastic interaction force. This thermodynamic driving force induces drift fluxes of these mobile clusters toward the surface, facilitating helium segregation. Moreover, the clusters' drift toward the surface enables cluster reactions, most importantly trap mutation, at rates much higher than in the bulk material. This cluster dynamics has significant effects on the surface morphology, near-surface defect structures, and the amount of helium retained in the material upon plasma exposure.

EDITORIAL: Plasma Surface Interactions for Fusion

NASA Astrophysics Data System (ADS)

2006-05-01

Because plasma-boundary physics encompasses some of the most important unresolved issues for both the International Thermonuclear Experimental Reactor (ITER) project and future fusion power reactors, there is a strong interest in the fusion community for better understanding and characterization of plasma wall interactions. Chemical and physical sputtering cause the erosion of the limiters/divertor plates and vacuum vessel walls (made of C, Be and W, for example) and degrade fusion performance by diluting the fusion fuel and excessively cooling the core, while carbon redeposition could produce long-term in-vessel tritium retention, degrading the superior thermo-mechanical properties of the carbon materials. Mixed plasma-facing materials are proposed, requiring optimization for different power and particle flux characteristics. Knowledge of material properties as well as characteristics of the plasma material interaction are prerequisites for such optimizations. Computational power will soon reach hundreds of teraflops, so that theoretical and plasma science expertise can be matched with new experimental capabilities in order to mount a strong response to these challenges. To begin to address such questions, a Workshop on New Directions for Advanced Computer Simulations and Experiments in Fusion-Related Plasma Surface Interactions for Fusion (PSIF) was held at the Oak Ridge National Laboratory from 21 to 23 March, 2005. The purpose of the workshop was to bring together researchers in fusion related plasma wall interactions in order to address these topics and to identify the most needed and promising directions for study, to exchange opinions on the present depth of knowledge of surface properties for the main fusion-related materials, e.g., C, Be and W, especially for sputtering, reflection, and deuterium (tritium) retention properties. The goal was to suggest the most important next steps needed for such basic computational and experimental work to be facilitated by researchers in fusion, material, and physical sciences. Representatives from many fusion research laboratories attended, and 25 talks were given, the majority of them making up the content of these Workshop proceedings. The presentations of all talks and further information on the Workshop are available at http://www-cfadc.phy.ornl.gov/psif/home.html. The workshop talks dealt with identification of needs from the perspective of integrated fusion simulation and ITER design, recent developments and perspectives on computation of plasma-facing surface properties using the current and expected new generation of computation capability, and with the status of dedicated laboratory experiments which characterize the underlying processes of PSIF. The Workshop summary and conclusions are being published in Nuclear Fusion 45 (2005). We are indebted to Lynda Saddiq and Fay Ownby, secretaries in the Physics Division of ORNL, whose special efforts, devotion, and expertise made possible both the Workshop and these Proceedings. J T Hogan, P S Krstic and F W Meyer Physics Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6372, USA

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

DOE PAGES

Scotti, F.; Soukhanovskii, V. A.

2015-12-09

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 band pass and neutral density filters on each optical path allowing for simultaneous imaging of emission at wavelengths differing in brightness up to 3more » 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). Furthermore, the diagnostic setup and initial results from its application on the lithium tokamak experiment are presented.« less

Hydrogen in tungsten as plasma-facing material

NASA Astrophysics Data System (ADS)

Roth, Joachim; Schmid, Klaus

2011-12-01

Materials facing plasmas in fusion experiments and future reactors are loaded with high fluxes (1020-1024 m-2 s-1) of H, D and T fuel particles at energies ranging from a few eV to keV. In this respect, the evolution of the radioactive T inventory in the first wall, the permeation of T through the armour into the coolant and the thermo-mechanical stability after long-term exposure are key parameters determining the applicability of a first wall material. Tungsten exhibits fast hydrogen diffusion, but an extremely low solubility limit. Due to the fast diffusion of hydrogen and the short ion range, most of the incident ions will quickly reach the surface and recycle into the plasma chamber. For steady-state operation the solute hydrogen for the typical fusion reactor geometry and wall conditions can reach an inventory of about 1 kg. However, in short-pulse operation typical of ITER, solute hydrogen will diffuse out after each pulse and the remaining inventory will consist of hydrogen trapped in lattice defects, such as dislocations, grain boundaries and irradiation-induced traps. In high-flux areas the hydrogen energies are too low to create displacement damage. However, under these conditions the solubility limit will be exceeded within the ion range and the formation of gas bubbles and stress-induced damage occurs. In addition, simultaneous neutron fluxes from the nuclear fusion reaction D(T,n)α will lead to damage in the materials and produce trapping sites for diffusing hydrogen atoms throughout the bulk. The formation and diffusive filling of these different traps will determine the evolution of the retained T inventory. This paper will concentrate on experimental evidence for the influence different trapping sites have on the hydrogen inventory in W as studied in ion beam experiments and low-temperature plasmas. Based on the extensive experimental data, models are validated and applied to estimate the contribution of different traps to the tritium inventory in future fusion reactors.

Damage of actively cooled plasma facing components of magnetic confinement controlled fusion machines

NASA Astrophysics Data System (ADS)

Chevet, G.; Schlosser, J.; Martin, E.; Herb, V.; Camus, G.

2009-03-01

Plasma facing components (PFCs) of magnetic fusion machines have high manufactured residual stresses and have to withstand important stress ranges during operation. These actively cooled PFCs have a carbon fibre composite (CFC) armour and a copper alloy heat sink. Cracks mainly appear in the CFC near the composite/copper interface. In order to analyse damage mechanisms, it is important to well simulate the damage mechanisms both of the CFC and the CFC/Cu interface. This study focuses on the mechanical behaviour of the N11 material for which the scalar ONERA damage model was used. The damage parameters of this model were identified by similarity to a neighbour material, which was extensively analysed, according to the few characterization test results available for the N11. The finite elements calculations predict a high level of damage of the CFC at the interface zone explaining the encountered difficulties in the PFCs fabrication. These results suggest that the damage state of the CFC cells is correlated with a conductivity decrease to explain the temperature increase of the armour surface under fatigue heat load.

An Overview of Recent PISCES Program PMI Results

NASA Astrophysics Data System (ADS)

Tynan, George; Doerner, Russell; Abe, Shota; Baldwin, Matthew; Barton, Joseph; Chen, Renkun; Gosselin, Jordan; Hollmann, Eric; Nishijima, Daisuke; Simmonds, Michael; Wang, Yong; Yu, Jonathan

2015-11-01

The PISCES Program is focused on fundamental PMI studies of Be and W-based solid plasma facing components under steady-state and transient conditions. We will show results from studies in W, Be and mixed W-Be material systems. Topics of investigation include formation of near-surface nanobubbles from He plasma ion implantation, growth of W-fuzz from these bubbles in steady-state and transient conditions, D retention in Be and W and development of a D-retention model for both H/D isotope exchange and displacement damage experiments. Initial studies of PMI in displacement damaged W are also presented, showing the effect of damage and exposure temperature on D retention, D diffusion, W thermal conductivity. Be-based results include morphology evolution under high plasma flux exposure, Be erosion mechanisms, and retention in Be-based materials. Future plans and connections to fusion energy system requirements will be discussed. This work supported by grant DE-FG02-07ER54912.

Direct depth distribution measurement of deuterium in bulk tungsten exposed to high-flux plasma

NASA Astrophysics Data System (ADS)

Taylor, C. N.; Shimada, M.

2017-05-01

Understanding tritium retention and permeation in plasma-facing components is critical for fusion safety and fuel cycle control. Glow discharge optical emission spectroscopy (GD-OES) is shown to be an effective tool to reveal the depth profile of deuterium in tungsten. Results confirm the detection of deuterium. A ˜46 μm depth profile revealed that the deuterium content decreased precipitously in the first 7 μm, and detectable amounts were observed to depths in excess of 20 μm. The large probing depth of GD-OES (up to 100s of μm) enables studies not previously accessible to the more conventional techniques for investigating deuterium retention. Of particular applicability is the use of GD-OES to measure the depth profile for experiments where high deuterium concentration in the bulk material is expected: deuterium retention in neutron irradiated materials, and ultra-high deuterium fluences in burning plasma environment.

Plasma-wall interaction in laser inertial fusion reactors: novel proposals for radiation tests of first wall materials

NASA Astrophysics Data System (ADS)

Alvarez Ruiz, J.; Rivera, A.; Mima, K.; Garoz, D.; Gonzalez-Arrabal, R.; Gordillo, N.; Fuchs, J.; Tanaka, K.; Fernández, I.; Briones, F.; Perlado, J.

2012-12-01

Dry-wall laser inertial fusion (LIF) chambers will have to withstand strong bursts of fast charged particles which will deposit tens of kJ m-2 and implant more than 1018 particles m-2 in a few microseconds at a repetition rate of some Hz. Large chamber dimensions and resistant plasma-facing materials must be combined to guarantee the chamber performance as long as possible under the expected threats: heating, fatigue, cracking, formation of defects, retention of light species, swelling and erosion. Current and novel radiation resistant materials for the first wall need to be validated under realistic conditions. However, at present there is a lack of facilities which can reproduce such ion environments. This contribution proposes the use of ultra-intense lasers and high-intense pulsed ion beams (HIPIB) to recreate the plasma conditions in LIF reactors. By target normal sheath acceleration, ultra-intense lasers can generate very short and energetic ion pulses with a spectral distribution similar to that of the inertial fusion ion bursts, suitable to validate fusion materials and to investigate the barely known propagation of those bursts through background plasmas/gases present in the reactor chamber. HIPIB technologies, initially developed for inertial fusion driver systems, provide huge intensity pulses which meet the irradiation conditions expected in the first wall of LIF chambers and thus can be used for the validation of materials too.

Measurements of tungsten migration in the DIII-D divertor

NASA Astrophysics Data System (ADS)

Wampler, W. R.; Rudakov, D. L.; Watkins, J. G.; McLean, A. G.; Unterberg, E. A.; Stangeby, P. C.

2017-12-01

An experimental study of migration of tungsten in the DIII-D divertor is described, in which the outer strike point of L-mode plasmas was positioned on a toroidal ring of tungsten-coated metal inserts. Net deposition of tungsten on the divertor just outside the strike point was measured on graphite samples exposed to various plasma durations using the divertor materials evaluation system. Tungsten coverage, measured by Rutherford backscattering spectroscopy (RBS), was found to be low and nearly independent of both radius and exposure time closer to the strike point, whereas farther from the strike point the W coverage was much larger and increased with exposure time. Depth profiles from RBS show this was due to accumulation of thicker mixed-material deposits farther from the strike point where the plasma temperature is lower. These results are consistent with a low near-surface steady-state coverage on graphite undergoing net erosion, and continuing accumulation in regions of net deposition. This experiment provides data needed to validate, and further improve computational simulations of erosion and deposition of material on plasma-facing components and transport of impurities in magnetic fusion devices. Such simulations are underway and will be reported later.

Permanent Magnet Ecr Plasma Source With Magnetic Field Optimization

DOEpatents

Doughty, Frank C.; Spencer, John E.

2000-12-19

In a plasma-producing device, an optimized magnet field for electron cyclotron resonance plasma generation is provided by a shaped pole piece. The shaped pole piece adjusts spacing between the magnet and the resonance zone, creates a convex or concave resonance zone, and decreases stray fields between the resonance zone and the workpiece. For a cylindrical permanent magnet, the pole piece includes a disk adjacent the magnet together with an annular cylindrical sidewall structure axially aligned with the magnet and extending from the base around the permanent magnet. The pole piece directs magnetic field lines into the resonance zone, moving the resonance zone further from the face of the magnet. Additional permanent magnets or magnet arrays may be utilized to control field contours on a local scale. Rather than a permeable material, the sidewall structure may be composed of an annular cylindrical magnetic material having a polarity opposite that of the permanent magnet, creating convex regions in the resonance zone. An annular disk-shaped recurve section at the end of the sidewall structure forms magnetic mirrors keeping the plasma off the pole piece. A recurve section composed of magnetic material having a radial polarity forms convex regions and/or magnetic mirrors within the resonance zone.

Smart tungsten alloys as a material for the first wall of a future fusion power plant

NASA Astrophysics Data System (ADS)

Litnovsky, A.; Wegener, T.; Klein, F.; Linsmeier, Ch.; Rasinski, M.; Kreter, A.; Unterberg, B.; Coenen, J. W.; Du, H.; Mayer, J.; Garcia-Rosales, C.; Calvo, A.; Ordas, N.

2017-06-01

Tungsten is currently deemed as a promising plasma-facing material (PFM) for the future power plant DEMO. In the case of an accident, air can get into contact with PFMs during the air ingress. The temperature of PFMs can rise up to 1200 °C due to nuclear decay heat in the case of damaged coolant supply. Heated neutron-activated tungsten forms a volatile radioactive oxide which can be mobilized into the atmosphere. New self-passivating ‘smart’ alloys can adjust their properties to the environment. During plasma operation the preferential sputtering of lighter alloying elements will leave an almost pure tungsten surface facing the plasma. During an accident the alloying elements in the bulk are forming oxides thus protecting tungsten from mobilization. Good plasma performance and the suppression of oxidation are required for smart alloys. Bulk tungsten (W)-chroimum (Cr)-titanium (Ti) alloys were exposed together with pure tungsten (W) samples to the steady-state deuterium plasma under identical conditions in the linear plasma device PSI 2. The temperature of the samples was ~576 °C-715 °C, the energy of impinging ions was 210 eV matching well the conditions expected at the first wall of DEMO. Weight loss measurements demonstrated similar mass decrease of smart alloys and pure tungsten samples. The oxidation of exposed samples has proven no effect of plasma exposure on the oxidation resistance. The W-Cr-Ti alloy demonstrated advantageous 3-fold lower mass gain due to oxidation than that of pure tungsten. New yttrium (Y)-containing thin film systems are demonstrating superior performance in comparison to that of W-Cr-Ti systems and of pure W. The oxidation rate constant of W-Cr-Y thin film is 105 times less than that of pure tungsten. However, the detected reactivity of the bulk smart alloy in humid atmosphere is calling for a further improvement.

Erosion and deposition in the JET divertor during the second ITER-like wall campaign

NASA Astrophysics Data System (ADS)

Mayer, M.; Krat, S.; Baron-Wiechec, A.; Gasparyan, Yu; Heinola, K.; Koivuranta, S.; Likonen, J.; Ruset, C.; de Saint-Aubin, G.; Widdowson, A.; Contributors, JET

2017-12-01

Erosion of plasma-facing materials and successive transport and redeposition of eroded material are crucial processes determining the lifetime of plasma-facing components and the trapped tritium inventory in redeposited material layers. Erosion and deposition in the JET divertor were studied during the second JET ITER-like wall campaign ILW-2 in 2013-2014 by using a poloidal row of specially prepared divertor marker tiles including the tungsten bulk tile 5. The marker tiles were analyzed using elastic backscattering with 3-4.5 MeV incident protons and nuclear reaction analysis using 0.8-4.5 MeV 3He ions before and after the campaign. The erosion/deposition pattern observed during ILW-2 is qualitatively comparable to the first campaign ILW-1 in 2011-2012: deposits consist mainly of beryllium with 5-20 at.% of carbon and oxygen and small amounts of Ni and W. The highest deposition with deposited layer thicknesses up to 30 μm per campaign is still observed on the upper and horizontal parts of the inner divertor. Outer divertor tiles 5, 6, 7 and 8 are net W erosion areas. The observed D inventory is roughly comparable to the inventory observed during ILW-1. The results obtained during ILW-2 therefore confirm the positive results observed in ILW-1 with respect to reduced material deposition and hydrogen isotopes retention in the divertor.

Theoretical investigation of crack formation in tungsten after heat loads

NASA Astrophysics Data System (ADS)

Arakcheev, A. S.; Huber, A.; Wirtz, M.; Sergienko, G.; Steudel, I.; Burdakov, A. V.; Coenen, J. W.; Kreter, A.; Linke, J.; Mertens, Ph.; Shoshin, A. A.; Unterberg, B.; Vasilyev, A. A.

2015-08-01

Transient events such as ELMs in large plasma devices lead to significant heat load on plasma-facing components (PFCs). ELMs cause mechanical damage of PFCs (e.g. cracks). The cracks appear due to stresses caused by thermal extension. Analytical calculations of the stresses are carried out for tungsten. The model only takes into account the basic features of solid body mechanics without material modifications (e.g. fatigue or recrystallization). The numerical results of the model demonstrate good agreement with experimental data obtained at the JUDITH-1, PSI-2 and GOL-3 facilities.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lucia, M., E-mail: mlucia@pppl.gov; Kaita, R.; Majeski, R.

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.

Material impacts and heat flux characterization of an electrothermal plasma source with an applied magnetic field

NASA Astrophysics Data System (ADS)

Gebhart, T. E.; Martinez-Rodriguez, R. A.; Baylor, L. R.; Rapp, J.; Winfrey, A. L.

2017-08-01

To produce a realistic tokamak-like plasma environment in linear plasma device, a transient source is needed to deliver heat and particle fluxes similar to those seen in an edge localized mode (ELM). ELMs in future large tokamaks will deliver heat fluxes of ˜1 GW/m2 to the divertor plasma facing components at a few Hz. An electrothermal plasma source can deliver heat fluxes of this magnitude. These sources operate in an ablative arc regime which is driven by a DC capacitive discharge. An electrothermal source was configured with two pulse lengths and tested under a solenoidal magnetic field to determine the resulting impact on liner ablation, plasma parameters, and delivered heat flux. The arc travels through and ablates a boron nitride liner and strikes a tungsten plate. The tungsten target plate is analyzed for surface damage using a scanning electron microscope.

Helium segregation on surfaces of plasma-exposed tungsten

DOE PAGES

Maroudas, Dimitrios; Blondel, Sophie; Hu, Lin; ...

2016-01-21

Here we report a hierarchical multi-scale modeling study of implanted helium segregation on surfaces of tungsten, considered as a plasma facing component in nuclear fusion reactors. We employ a hierarchy of atomic-scale simulations based on a reliable interatomic interaction potential, including molecular-statics simulations to understand the origin of helium surface segregation, targeted molecular-dynamics (MD) simulations of near-surface cluster reactions, and large-scale MD simulations of implanted helium evolution in plasma-exposed tungsten. We find that small, mobile He-n (1 <= n <= 7) clusters in the near-surface region are attracted to the surface due to an elastic interaction force that provides themore » thermodynamic driving force for surface segregation. Elastic interaction force induces drift fluxes of these mobile Hen clusters, which increase substantially as the migrating clusters approach the surface, facilitating helium segregation on the surface. Moreover, the clusters' drift toward the surface enables cluster reactions, most importantly trap mutation, in the near-surface region at rates much higher than in the bulk material. Moreover, these near-surface cluster dynamics have significant effects on the surface morphology, near-surface defect structures, and the amount of helium retained in the material upon plasma exposure. We integrate the findings of such atomic-scale simulations into a properly parameterized and validated spatially dependent, continuum-scale reaction-diffusion cluster dynamics model, capable of predicting implanted helium evolution, surface segregation, and its near-surface effects in tungsten. This cluster-dynamics model sets the stage for development of fully atomistically informed coarse-grained models for computationally efficient simulation predictions of helium surface segregation, as well as helium retention and surface morphological evolution, toward optimal design of plasma facing components.« less

Helium segregation on surfaces of plasma-exposed tungsten

NASA Astrophysics Data System (ADS)

Maroudas, Dimitrios; Blondel, Sophie; Hu, Lin; Hammond, Karl D.; Wirth, Brian D.

2016-02-01

We report a hierarchical multi-scale modeling study of implanted helium segregation on surfaces of tungsten, considered as a plasma facing component in nuclear fusion reactors. We employ a hierarchy of atomic-scale simulations based on a reliable interatomic interaction potential, including molecular-statics simulations to understand the origin of helium surface segregation, targeted molecular-dynamics (MD) simulations of near-surface cluster reactions, and large-scale MD simulations of implanted helium evolution in plasma-exposed tungsten. We find that small, mobile He n (1  ⩽  n  ⩽  7) clusters in the near-surface region are attracted to the surface due to an elastic interaction force that provides the thermodynamic driving force for surface segregation. This elastic interaction force induces drift fluxes of these mobile He n clusters, which increase substantially as the migrating clusters approach the surface, facilitating helium segregation on the surface. Moreover, the clusters’ drift toward the surface enables cluster reactions, most importantly trap mutation, in the near-surface region at rates much higher than in the bulk material. These near-surface cluster dynamics have significant effects on the surface morphology, near-surface defect structures, and the amount of helium retained in the material upon plasma exposure. We integrate the findings of such atomic-scale simulations into a properly parameterized and validated spatially dependent, continuum-scale reaction-diffusion cluster dynamics model, capable of predicting implanted helium evolution, surface segregation, and its near-surface effects in tungsten. This cluster-dynamics model sets the stage for development of fully atomistically informed coarse-grained models for computationally efficient simulation predictions of helium surface segregation, as well as helium retention and surface morphological evolution, toward optimal design of plasma facing components.

Modelling deuterium release from tungsten after high flux high temperature deuterium plasma exposure

NASA Astrophysics Data System (ADS)

Grigorev, Petr; Matveev, Dmitry; Bakaeva, Anastasiia; Terentyev, Dmitry; Zhurkin, Evgeny E.; Van Oost, Guido; Noterdaeme, Jean-Marie

2016-12-01

Tungsten is a primary candidate for plasma facing materials for future fusion devices. An important safety concern in the design of plasma facing components is the retention of hydrogen isotopes. Available experimental data is vast and scattered, and a consistent physical model of retention of hydrogen isotopes in tungsten is still missing. In this work we propose a model of non-equilibrium hydrogen isotopes trapping under fusion relevant plasma exposure conditions. The model is coupled to a diffusion-trapping simulation tool and is used to interpret recent experiments involving high plasma flux exposures. From the computational analysis performed, it is concluded that high flux high temperature exposures (T = 1000 K, flux = 1024 D/m2/s and fluence of 1026 D/m2) result in generation of sub-surface damage and bulk diffusion, so that the retention is driven by both sub-surface plasma-induced defects (bubbles) and trapping at natural defects. On the basis of the non-equilibrium trapping model we have estimated the amount of H stored in the sub-surface region to be ∼10-5 at-1, while the bulk retention is about 4 × 10-7 at-1, calculated by assuming the sub-surface layer thickness of about 10 μm and adjusting the trap concentration to comply with the experimental results for the integral retention.

Development of lithium and tungsten limiters for test on T-10 tokamak at high heat load condition

NASA Astrophysics Data System (ADS)

Lyublinski, I. E.; Vertkov, A. V.; Zharkov, M. Yu; Vershkov, V. A.; Mirnov, S. V.

2016-04-01

Application of a complex of powerful (up to 3 MW) ECR plasma heating in T-10 tokamak is pulled down with a problem of the strong plasma pollution at power input more than 2 MW. For the solution of these problems the new W and Li limiters is developed and prepared to implementation. As it is supposed, application of W as a plasma facing material will allow excluding carbon influx into vacuum chamber. An additional Li limiter arranged in a shadow of W one will be used as a Li source for plasma periphery cooling due to a reradiation on Li that will lead to decrease in power deposition on W limiters. Parameters and design of limiters are presented. Plasma facing surface of a limiter is made of capillary-porous system (CPS) with Li. Porous matrix of CPS (W felt) provides stability of liquid Li surface under MHD force effect and an opportunity of its constant renewal due to capillary forces. The necessary Li flux from a Li limiter surface is estimated for maintenance of normal operation mode of W limiters at ECRH power of 3 MW during 400 ms. It is shown, that upgrade of limiters in tokamak T-10 will allow providing of ECR plasma heating with power up to 3 MW at reasonable Li flux.

Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Federici, G.; Skinner, C.H.; Brooks, J.N.

2001-01-10

The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of themore » important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.« less

Developing a compact toroid injector in the ThermoElectric driven Liquid metal plasma facing Structures device

NASA Astrophysics Data System (ADS)

Christenson, Michael; Szott, Matthew; Kalathiparambil, Kishor; Sovinec, Carl; Ruzic, David

2016-10-01

The ThermoElectric-driven Liquid-metal plasma-facing Structures (TELS) device at the University of Illinois is a theta-pinched, plasma-material interaction test stand used to simulate extreme events in the edge and divertor regions of a tokamak plasma. Previous measurements of the electron and ion temperatures have shown that the isotropic heat load on target ranges between 0.1 and 0.2 MJ m-2 over a pulse lasting 0.2 ms. While this compares well to the heat loads from Type 1 ELMs in larger toroidal devices, it is still much less than the energy deposition from Type 1 ELMs expected in ITER, which are in excess of 1 MJ m-2. To this end, a compact toroid (CT) injector has been proposed as a modification to the existing TELS device. By using an externally applied bias field to force reconnection at the muzzle of the coaxial plasma accelerator source that drives ionization, NIMROD MHD simulations have shown a peak magnetic flux of 3.5 mWb is reached 0.025 ms into the pulse - more than sufficient to form a CT. Early calorimetry and magnetic field measurements indicate that a new plasma structure has been formed in the magnetized coaxial plasma source. This work presents the current results of CT generation with respect to the bias field strength as well as the coaxial source geometry. DOE OFES DE-SC0008587, DE-SC0008658, DE-FG02-99ER54515.

Modelling structural and plasma facing materials for fusion power plants: Recent advances and outstanding issues in the EURATOM fusion materials programme

NASA Astrophysics Data System (ADS)

Boutard, Jean-Louis; Dudarev, Sergei; Rieth, Michael

2011-10-01

EFDA Fusion Materials Topical Group was established at the end of 2007 to coordinate the EU effort on the development of structural and protection materials able to withstand the very demanding operating conditions of a future DEMO power plant. Focusing on a selection of well identified materials issues, including the behaviour of Reduced Activation Ferritic-Martensitic steels, and W-alloys under the foreseen operation conditions in a future DEMO, this paper describes recent advances in physical modelling and experimental validation, contributing to the definition of chemical composition and microstructure of materials with improved in-service stability at high temperature, high neutron flux and intense ion bombardment.

Counter-facing plasma guns for efficient extreme ultra-violet plasma light source

NASA Astrophysics Data System (ADS)

Kuroda, Yusuke; Yamamoto, Akiko; Kuwabara, Hajime; Nakajima, Mitsuo; Kawamura, Tohru; Horioka, Kazuhiko

2013-11-01

A plasma focus system composed of a pair of counter-facing coaxial guns was proposed as a long-pulse and/or repetitive high energy density plasma source. We applied Li as the source of plasma for improvement of the conversion efficiency, the spectral purity, and the repetition capability. For operation of the system with ideal counter-facing plasma focus mode, we changed the system from simple coaxial geometry to a multi-channel configuration. We applied a laser trigger to make synchronous multi-channel discharges with low jitter. The results indicated that the configuration is promising to make a high energy density plasma with high spectral efficiency.

Beryllium processing technology review for applications in plasma-facing components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Castro, R.G.; Jacobson, L.A.; Stanek, P.W.

1993-07-01

Materials research and development activities for the International Thermonuclear Experimental Reactor (ITER), i.e., the next generation fusion reactor, are investigating beryllium as the first-wall containment material for the reactor. Important in the selection of beryllium is the ability to process, fabricate and repair beryllium first-wall components using existing technologies. Two issues that will need to be addressed during the engineering design activity will be the bonding of beryllium tiles in high-heat-flux areas of the reactor, and the in situ repair of damaged beryllium tiles. The following review summarizes the current technology associated with welding and joining of beryllium to itselfmore » and other materials, and the state-of-the-art in plasma-spray technology as an in situ repair technique for damaged beryllium tiles. In addition, a review of the current status of beryllium technology in the former Soviet Union is also included.« less

Proof-of-concept experiment for on-line laser induced breakdown spectroscopy analysis of impurity layer deposited on optical window and other plasma facing components of Aditya tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maurya, Gulab Singh; Kumar, Rohit; Rai, Awadhesh Kumar, E-mail: awadheshkrai@rediffmail.com

2015-12-15

In the present manuscript, we demonstrate the design of an experimental setup for on-line laser induced breakdown spectroscopy (LIBS) analysis of impurity layers deposited on specimens of interest for fusion technology, namely, plasma-facing components (PFCs) of a tokamak. For investigation of impurities deposited on PFCs, LIBS spectra of a tokamak wall material like a stainless steel sample (SS304) have been recorded through contaminated and cleaned optical windows. To address the problem of identification of dust and gases present inside the tokamak, we have shown the capability of the apparatus to record LIBS spectra of gases. A new approach known asmore » “back collection method” to record LIBS spectra of impurities deposited on the inner surface of optical window is presented.« less

Non-uniform Erosion and Surface Evolution of Plasma-Facing Materials for Electric Propulsion

NASA Astrophysics Data System (ADS)

Matthes, Christopher Stanley Rutter

A study regarding the surface evolution of plasma-facing materials is presented. Experimental efforts were performed in the UCLA Pi Facility, designed to explore the physics of plasma-surface interactions. The influence of micro-architectured surfaces on the effects of plasma sputtering is compared with the response of planar samples. Ballistic deposition of sputtered atoms as a result of geometric re-trapping is observed. This provides a self-healing mechanism of micro-architectured surfaces during plasma exposure. This result is quantified using a QCM to demonstrate the evolution of surface features and the corresponding influence on the instantaneous sputtering yield. The sputtering yield of textured molybdenum samples exposed to 300 eV Ar plasma is found to be roughly 1 of the 2 corresponding value of flat samples, and increases with ion fluence. Mo samples exhibited a sputtering yield initially as low as 0.22+/-8%, converging to 0.4+/-8% at high fluence. Although the yield is dependent on the initial surface structure, it is shown to be transient, reaching a steady-state value that is independent of initial surface conditions. A continuum model of surface evolution resulting from sputtering, deposition and surface diffusion is also derived to resemble the damped Kuramoto-Sivashinsky (KS) equation of non-linear dynamics. Linear stability analysis of the evolution equation provides an estimate of the selected wavelength, and its dependence on the ion energy and angle of incidence. The analytical results are confirmed by numerical simulations of the equation with a Fast Fourier Transform method. It is shown that for an initially flat surface, small perturbations lead to the evolution of a selected surface pattern that has nano- scale wavelength. When the surface is initially patterned by other means, the final resulting pattern is a competition between the "templated" pattern and the "self-organized" structure. Potential future routes of research are also discussed, corresponding to a design analysis of the current experimental study.

Design of a digital holography system for PFC erosion measurements on Proto-MPEX.

PubMed

Thomas, C E Tommy; Biewer, T M; Baylor, L R; Combs, S K; Meitner, S J; Rapp, J; Hillis, D L; Granstedt, E M; Majeski, R; Kaita, R

2016-11-01

A project has been started at ORNL to develop a dual-wavelength digital holography system for plasma facing component erosion measurements on prototype material plasma exposure experiment. Such a system will allow in situ real-time measurements of component erosion. Initially the system will be developed with one laser, and first experimental laboratory measurements will be made with the single laser system. In the second year of development, a second CO 2 laser will be added and measurements with the dual wavelength system will begin. Adding the second wavelength allows measurements at a much longer synthetic wavelength.

Manufacturing of reliable actively cooled fusion components - a challenge for non-destructive inspections

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reheis, N.; Zabernig, A.; Ploechl, L.

1994-12-31

Actively cooled in-vessel components like divertors or limiters require high quality and reliability to ensure safe operation during long term use. Such components are subjected to very severe thermal and mechanical cyclic loads and high power densities. Key requirements for materials in question are e.g. high melting point and thermal conductivity and low atomic mass number. Since no single material can simultaneously meet all of these requirements the selection of materials to be combined in composite components as well as of manufacturing and non-destructive inspection (NDI) methods is a particularly challenging task. Armour materials like graphite intended to face themore » plasma and help to maintain its desired properties, are bonded to metallic substrates like copper, molybdenum or stainless steel providing cooling and mechanical support. Several techniques such as brazing and active metal casting have been developed and successfully applied for joining materials with different thermophysical properties, pursuing the objective of sufficient heat dissipation from the hot, plasma facing surface to the coolant. NDI methods are an integral part of the manufacturing schedule of these components, starting in the design phase and ending in the final inspection. They apply all kinds of divertor types (monobloc and flat-tile concept). Particular focus is put on the feasibility of detecting small flaws and defects in complex interfaces and on the limits of these techniques. Special test pieces with defined defects acting as standards were inspected. Accompanying metallographic investigations were carried out to compare actual defects with results recorded during NDI.« less

Plasma wall interaction, a key issue on the way to a steady state burning fusion device

NASA Astrophysics Data System (ADS)

Philipps, V.

2006-04-01

The International Tokamak Experimental Reactor (ITER), the first burning fusion plasma experiment based on the tokamak principle, is ready for construction. It is based on many years of fusion research resulting in a robust design in most of the areas. Present day fusion research concentrates on the remaining critical issues which are, to a large extent, connected with processes of plasma wall interaction. This is mainly due to extended duty cycle and the increase of the plasma stored energy in comparison with present-day machines. Critical topics are the lifetime of the plasma facing components (PFC) and the long-term tritium retention. These processes are controlled mainly by material erosion, both during steady state operation and transient power losses (disruptions and edge localized modes (ELMs)) and short- and long-range material migration and re-deposition. The extrapolation from present-day 'full carbon wall' devices suggests that the long-term tritium retention in a burning fusion device would be unacceptably high under these conditions allowing for only an unacceptable limited number of pulses in a D T mixture. As a consequence of this, research activities have been strengthened to understand in more detail the underlying processes of material erosion and re-deposition, to develop methods to remove retained tritium from the PFCs and remote areas of a fusion device and to explore these processes and the plasma performance in more detail with metallic PFC, such as beryllium (Be) and tungsten (W), which are foreseen for the ITER experiment. This paper outlines the main physical mechanisms leading to material erosion, migration and re-deposition and the associated fuel retention. It addresses the experimental database in these areas and describes the further research strategies that will be needed to tackle critical issues.

Modeling of high-Z materials erosion and its suppression in DIII-D

NASA Astrophysics Data System (ADS)

Ding, Rui; Guo, H. Y.; Chan, V. S.; Snyder, P. B.; Rudakov, D. L.; Stangeby, P. C.; Elder, J. D.; Tskhakaya, D.; Wampler, W. R.; Kirschner, A.; McLean, A. G.

2015-11-01

Erosion of plasma facing components is a key issue for high-power, long pulse operation. The 3D Monte Carlo code ERO has been used to simulate the erosion/redeposition of Mo and W samples exposed to DIII-D divertor plasma using the DiMES. The net erosion rate is significantly reduced due to the high local re-deposition ratio of eroded materials, which is mainly controlled by the electric field and plasma density within the Chodura sheath as indicated by ERO modeling. Similar re-deposition ratios were obtained from the modeling using three sheath models for small inclined magnetic field angle, all being close to the measured value. ERO modeling shows that local CH4 injection can create a carbon coating on the Mo sample to mitigate Mo erosion; the local decrease of electron temperature due to gas injection also suppresses net erosion, consistent with experimental observation. Supported by the US DOE under DE-FC02-04ER54698 and PSI-SciDAC project.

Conceptual design and development of GEM based detecting system for tomographic tungsten focused transport monitoring

NASA Astrophysics Data System (ADS)

Chernyshova, M.; Czarski, T.; Malinowski, K.; Kowalska-Strzęciwilk, E.; Poźniak, K.; Kasprowicz, G.; Zabołotny, W.; Wojeński, A.; Kolasiński, P.; Mazon, D.; Malard, P.

2015-10-01

Implementing tungsten as a plasma facing material in ITER and future fusion reactors will require effective monitoring of not just its level in the plasma but also its distribution. That can be successfully achieved using detectors based on Gas Electron Multiplier (GEM) technology. This work presents the conceptual design of the detecting unit for poloidal tomography to be tested at the WEST project tokamak. The current stage of the development is discussed covering aspects which include detector's spatial dimensions, gas mixtures, window materials and arrangements inside and outside the tokamak ports, details of detector's structure itself and details of the detecting module electronics. It is expected that the detecting unit under development, when implemented, will add to the safe operation of tokamak bringing the creation of sustainable nuclear fusion reactors a step closer. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

Quality control of FWC during assembly and commissioning in SST-1 Tokamak

NASA Astrophysics Data System (ADS)

Patel, Hitesh; Santra, Prosenjit; Parekh, Tejas; Biswas, Prabal; Jayswal, Snehal; Chauhan, Pradeep; Paravastu, Yuvakiran; George, Siju; Semwal, Pratibha; Thankey, Prashant; Ramesh, Gattu; Prakash, Arun; Dhanani, Kalpesh; Raval, D. C.; Khan, Ziauddin; Pradhan, Subrata

2017-04-01

First Wall Components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma, comprises of limiters, divertors, baffles, passive stabilizers designed to operate long duration (∼1000 s) discharges of elongated plasma. All FWC consist of copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at inter-connected ring & port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 undergoing a rigorous quality control and checks at every stage of the assembly process. This paper will present the quality control aspects and checks of FWC from commencement of assembly procedure, namely material test reports, leak testing of high temperature baked components, assembled dimensional tolerances, leak testing of all welded joints, graphite tile tightening torques, electrical continuity and electrical isolation of passive stabilizers from vacuum vessel, baking and cooling hydraulic connections inside vacuum vessel.

Liquid-metal plasma-facing component research on the National Spherical Torus Experiment

NASA Astrophysics Data System (ADS)

Jaworski, M. A.; Khodak, A.; Kaita, R.

2013-12-01

Liquid metal plasma-facing components (PFCs) have been proposed as a means of solving several problems facing the creation of economically viable fusion power reactors. Liquid metals face critical issues in three key areas: free-surface stability, material migration and demonstration of integrated scenarios. To date, few demonstrations exist of this approach in a diverted tokamak and we here provide an overview of such work on the National Spherical Torus Experiment (NSTX). The liquid lithium divertor (LLD) was installed and operated for the 2010 run campaign using evaporated coatings as the filling method. Despite a nominal liquid level exceeding the capillary structure and peak current densities into the PFCs exceeding 100 kA m-2, no macroscopic ejection events were observed. The stability can be understood from a Rayleigh-Taylor instability analysis. Capillary restraint and thermal-hydraulic considerations lead to a proposed liquid-metal PFCs scheme of actively-supplied, capillary-restrained systems. Even with state-of-the-art cooling techniques, design studies indicate that the surface temperature with divertor-relevant heat fluxes will still reach temperatures above 700 °C. At this point, one would expect significant vapor production from a liquid leading to a continuously vapor-shielded regime. Such high-temperature liquid lithium PFCs may be possible on the basis of momentum-balance arguments.

In situ measurements of fuel retention by laser induced desorption spectroscopy in TEXTOR

NASA Astrophysics Data System (ADS)

Zlobinski, M.; Philipps, V.; Schweer, B.; Huber, A.; Stoschus, H.; Brezinsek, S.; Samm, U.; TEXTOR Team

2011-12-01

In future fusion devices such as ITER tritium retention due to tritium co-deposition in mixed material layers can be a serious safety problem. Laser induced desorption spectroscopy (LIDS) can measure the hydrogen content of hydrogenic carbon layers locally on plasma-facing components, while hydrogen is used as a tritium substitute. For several years, this method has been applied in the TEXTOR tokamak in situ during plasma operation to monitor the hydrogen content in space and time. This work shows the LIDS signal reproducibility and studies the effects of different plasma conditions, desorption distances from the plasma and different laser energies using a dedicated sample with constant hydrogen amount. Also the LIDS signal evaluation procedure is described in detail and the detection limits for different conditions in the TEXTOR tokamak are estimated.

Study of ion-irradiated tungsten in deuterium plasma

NASA Astrophysics Data System (ADS)

Khripunov, B. I.; Gureev, V. M.; Koidan, V. S.; Kornienko, S. N.; Latushkin, S. T.; Petrov, V. B.; Ryazanov, A. I.; Semenov, E. V.; Stolyarova, V. G.; Danelyan, L. S.; Kulikauskas, V. S.; Zatekin, V. V.; Unezhev, V. N.

2013-07-01

Experimental study aimed at investigation of neutron induced damage influence on fusion reactor plasma facing materials is reported. Displacement damage was produced in tungsten by high-energy helium and carbon ions at 3-10 MeV. The reached level of displacement damage ranged from several dpa to 600 dpa. The properties of the irradiated tungsten were studied in steady-state deuterium plasma on the LENTA linear divertor simulator. Plasma exposures were made at 250 eV of ion energy to fluence 1021-1022 ion/сm2. Erosion dynamics of the damaged layer and deuterium retention were observed. Surface microstructure modifications and important damage of the 5 μm layer shown. Deuterium retention in helium-damaged tungsten (ERD) showed its complex behavior (increase or decrease) depending on implanted helium quantity and the structure of the surface layer.

Nanochannel structures in W enhance radiation tolerance

DOE PAGES

Qin, Wenjing; Ren, Feng; Doerner, Russell P.; ...

2018-04-23

Developing high performance plasma facing materials (PFMs) is one of the greatest challenges for fusion reactors, because PFMs face unprecedented harsh environments including high flux plasma exposure, fast neutron irradiation and large transmutation gas. Tungsten (W) is considered as one of the most promising PFMs. Rapid accumulation of helium (He) atoms in such environments can lead to the He bubbles nucleation and even the formation of nano- to micro-scale “fuzz” on W surface, which greatly degrade the properties of W itself. The possible ejection of large W particulates into the core plasma can cause plasma instabilities. In this paper, wemore » present a new strategy to address the root causes of bubble nucleation and “fuzz” formation by concurrently releasing He outside of W matrix through the nano-engineered channel structure (nanochannels). Comparing to ordinary bulk W, nanochannel W films with high surface-to-volume ratios are found to not only delay the growth of He bubbles, but also suppress the formation of “fuzz” (less than a half of the “fuzz” thickness formation in bulk W). Finally, molecular dynamic (MD) simulation results elucidate that low vacancy formation energy and high He binding energy in the nanochannel surface effectively help He release and affect He clusters distribution in W during He ion irradiation.« less

Nanochannel structures in W enhance radiation tolerance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qin, Wenjing; Ren, Feng; Doerner, Russell P.

Developing high performance plasma facing materials (PFMs) is one of the greatest challenges for fusion reactors, because PFMs face unprecedented harsh environments including high flux plasma exposure, fast neutron irradiation and large transmutation gas. Tungsten (W) is considered as one of the most promising PFMs. Rapid accumulation of helium (He) atoms in such environments can lead to the He bubbles nucleation and even the formation of nano- to micro-scale “fuzz” on W surface, which greatly degrade the properties of W itself. The possible ejection of large W particulates into the core plasma can cause plasma instabilities. In this paper, wemore » present a new strategy to address the root causes of bubble nucleation and “fuzz” formation by concurrently releasing He outside of W matrix through the nano-engineered channel structure (nanochannels). Comparing to ordinary bulk W, nanochannel W films with high surface-to-volume ratios are found to not only delay the growth of He bubbles, but also suppress the formation of “fuzz” (less than a half of the “fuzz” thickness formation in bulk W). Finally, molecular dynamic (MD) simulation results elucidate that low vacancy formation energy and high He binding energy in the nanochannel surface effectively help He release and affect He clusters distribution in W during He ion irradiation.« less

Material impacts and heat flux characterization of an electrothermal plasma source with an applied magnetic field

DOE PAGES

Gebhart, T. E.; Martinez-Rodriguez, R. A.; Baylor, L. R.; ...

2017-08-11

To produce a realistic tokamak-like plasma environment in linear plasma device, a transient source is needed to deliver heat and particle fluxes similar to those seen in an edge localized mode (ELM). ELMs in future large tokamaks will deliver heat fluxes of ~1 GW/m 2 to the divertor plasma facing components at a few Hz. An electrothermal plasma source can deliver heat fluxes of this magnitude. These sources operate in an ablative arc regime which is driven by a DC capacitive discharge. An electrothermal source was configured in this paper with two pulse lengths and tested under a solenoidal magneticmore » field to determine the resulting impact on liner ablation, plasma parameters, and delivered heat flux. The arc travels through and ablates a boron nitride liner and strikes a tungsten plate. Finally, the tungsten target plate is analyzed for surface damage using a scanning electron microscope.« less

Material impacts and heat flux characterization of an electrothermal plasma source with an applied magnetic field

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gebhart, T. E.; Martinez-Rodriguez, R. A.; Baylor, L. R.

To produce a realistic tokamak-like plasma environment in linear plasma device, a transient source is needed to deliver heat and particle fluxes similar to those seen in an edge localized mode (ELM). ELMs in future large tokamaks will deliver heat fluxes of ~1 GW/m 2 to the divertor plasma facing components at a few Hz. An electrothermal plasma source can deliver heat fluxes of this magnitude. These sources operate in an ablative arc regime which is driven by a DC capacitive discharge. An electrothermal source was configured in this paper with two pulse lengths and tested under a solenoidal magneticmore » field to determine the resulting impact on liner ablation, plasma parameters, and delivered heat flux. The arc travels through and ablates a boron nitride liner and strikes a tungsten plate. Finally, the tungsten target plate is analyzed for surface damage using a scanning electron microscope.« less

Measurements of ion temperature and flow of pulsed plasmas produced by a magnetized coaxial plasma gun device using an ion Doppler spectrometer

NASA Astrophysics Data System (ADS)

Kitagawa, Y.; Sakuma, I.; Iwamoto, D.; Kikuchi, Y.; Fukumoto, N.; Nagata, M.

2012-10-01

It is important to know surface damage characteristics of plasma-facing component materials during transient heat and particle loads such as type I ELMs. A magnetized coaxial plasma gun (MCPG) device has been used as transient heat and particle source in ELM simulation experiments. Characteristics of pulsed plasmas produced by the MCPG device play an important role for the plasma material interaction. In this study, ion temperature and flow velocity of pulsed He plasmas were measured by an ion Doppler spectrometer (IDS). The IDS system consists of a light collection system including optical fibers, 1m-spectrometer and a 16 channel photomultiplier tube (PMT) detector. The IDS system measures the width and Doppler shift of HeII (468.58 nm) emission line with the time resolution of 1 μs. The Doppler broadened and shifted spectra were measured with 45 and 135 degree angles with respect to the plasmoid traveling direction. The observed emission line profile was represented by sum of two Gaussian components to determine the temperature and flow velocity. The minor component at around the wavelength of zero-velocity was produced by the stationary plasma. As the results, the ion velocity and temperature were 68 km/s and 19 eV, respectively. Thus, the He ion flow energy is 97 eV. The observed flow velocity agrees with that measured by a time of flight technique.

PRELIMINARY PROGRESS IN THE DEVELOPMENT OF DUCTILE-PHASE TOUGHENED TUNGSTEN FOR PLASMA-FACING MATERIALS: DUAL-PHASE FINITE ELEMENT DAMAGE MODELS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Henager, Charles H.; Nguyen, Ba Nghiep; Kurtz, Richard J.

The objective of this study is to develop a finite element continuum damage model suitable for modeling deformation, cracking, and crack bridging for W-Cu, W-Ni-Fe, and other ductile phase toughened W-composites, or more generally, any multi-phase composite structure where two or more phases undergo cooperative deformation in a composite system.

Development of a Laser Induced Fluorescence (LIF) system on the Plasma Material Interaction System (PLAMIS-II) device

NASA Astrophysics Data System (ADS)

Kang, I. J.; Lee, K. Y.; Lee, K. I.; Choi, Y.-S.; Cho, S. G.; Bae, M. K.; Lee, D.-H.; Hong, S. H.; Lho, T.; Chung, K.-S.

2015-12-01

A laser induced fluorescence (LIF) system has been developed for the plasma material interaction system (PLAMIS-II) device, which is equipped with a unique plasma gun composed of a LaB6 cathode and two anodes with electromagnets to generate a focused dense plasma. PLAMIS-II simulates the interactions of plasma with different materials and is to be used for the test of plasma facing components of fusion devices. The LIF system is composed of a seed laser with Littmann/Metcalf cavity and a master oscillator power amplifier to pump 3d4F7/2 metastable argon ion to 4p4D5/2 level at the wavelength of 668.61 nm, which has the following input parameters: laser power = 20 mW, line width < 100 kHz, and a mode-hop free tuning range > 70 GHz. For in-situ measurement of laser wavelength, the wavelength spectrum of an iodine cell was measured by a photo-transistor during LIF measurement. To measure argon ion temperature (Ti) and drift velocity (vd) in PLAMIS-II, the fluorescence light with the wavelength of 442.72 nm, emitted from 4p4D5/2 level to 4s4P3/2 level and passing through 1 nm band-width filter, was collected by the photomultiplier tube combined with a lock-in amplifier and a chopper with frequency of 3 kHz. Initial data of Ti and vd were analysed in terms of gas flow rate and applied power.

Final Progress Report The U.S. Department of Energy Research Grant No. DE-SC0008660 Plasma Surface Interactions: Bridging from the Surface to the Micron Frontier through Leadership Class Computing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krasheninnikov, Sergei; Smirnov, Roman; Guterl, Jerome

The choice of material for the plasma facing components (PFC), in particular, for divertor targets, is one of the main issues for future tokamak reactors. There are two major requirements for the PFC’s material: acceptable level of tritium retention and durability in a harsh environment of fusion grade plasma. Based on these criteria, some years ago it was decided that tungsten is an acceptable material for divertor targets in ITER. However, further experimental studies reveal that the irradiation of tungsten even with low energetic (well below sputtering threshold!) He containing plasma causes significant modification of surface morphology, formation of themore » layer of He nano-bubbles (in the temperature range T<1000 K), “fuzz” (for 1000 K2000 K) (e.g. see Fig. 1). Recall that He, being an ash of D-T fusion reactions, is an inherent impurity in fusion plasma. The goals of the UCSD Applied Plasma Theory Group was: i) investigate the mechanisms of the formation of He nano-bubble layer and fuzz growth under He irradiation, as well as the physics of transport of hydrogen species in tungsten lattice, and ii) develop physics understanding of the models suitable for the incorporation into the Xolotl-PSI code based on the reaction-diffusion approach, which is the flagship of the whole SciDAC project [8], which can guide both numerical simulations and experimental studies. Here we just highlight our major accomplishments.« less

Long-term erosion of plasma-facing materials with different surface roughness in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Hakola, A.; Karhunen, J.; Koivuranta, S.; Likonen, J.; Balden, M.; Herrmann, A.; Mayer, M.; Müller, H. W.; Neu, R.; Rohde, V.; Sugiyama, K.; The ASDEX Upgrade Team

2014-04-01

The effect of surface roughness on the long-term erosion patterns of tungsten coatings was investigated in the outer strike-point region of ASDEX Upgrade during its 2010-11 plasma operations. The net erosion rates of rough coatings (Ra = 5-6 μm) were three to seven times smaller than those of smooth coatings (Ra = 0.4-0.8 μm). This is because rough surfaces are largely modified and damaged in the microscopic scale but the material is re-deposited together with boron, deuterium and carbon on the shadowed sides of the most prominent surface features. In addition, we observed that W coatings were eroded on average at a rate of 0.03 nm s-1, which was three to four times smaller than the value for Cr, simulating here steel.

High pulse number thermal shock tests on tungsten with steady state particle background

NASA Astrophysics Data System (ADS)

Wirtz, M.; Kreter, A.; Linke, J.; Loewenhoff, Th; Pintsuk, G.; Sergienko, G.; Steudel, I.; Unterberg, B.; Wessel, E.

2017-12-01

Thermal fatigue of metallic materials, which will be exposed to severe environmental conditions e.g. plasma facing materials in future fusion reactors, is an important issue in order to predict the life time of complete wall components. Therefore experiments in the linear plasma device PSI-2 were performed to investigate the synergistic effects of high pulse number thermal shock events (L = 0.38 GW m-2, Δt = 0.5 ms) and stationary D/He (6%) plasma particle background on the thermal fatigue behavior of tungsten. Similar to experiments with pure thermal loads, the induced microstructural and surface modifications such as recrystallization and roughening as well as crack formation become more pronounced with increasing number of thermal shock events. However, the amount of damage significantly increases for synergistic loads showing severe surface roughening, plastic deformation and erosion resulting from the degradation of the mechanical properties caused by bombardment and diffusion of D/He to the surface and the bulk of the material. Additionally, D/He induced blistering and bubble formation were observed for all tested samples, which could change the thermal and mechanical properties of near surface regions.

Effect of starting microstructure on helium plasma-materials interaction in tungsten

DOE PAGES

Wang, Kun; Bannister, Mark E.; Meyer, Fred W.; ...

2016-11-24

Here, in a magnetic fusion energy (MFE) device, the plasma-facing materials (PFMs) will be subjected to tremendous fluxes of ions, heat, and neutrons. The response of PFMs to the fusion environment is still not well defined. Tungsten metal is the present candidate of choice for PFM applications such as the divertor in ITER. However, tungsten's microstructure will evolve in service, possibly to include recrystallization. How tungsten's response to plasma exposure evolves with changes in microstructure is presently unknown. In this work, we have exposed hot-worked and recrystallized tungsten to an 80 eV helium ion beam at a temperature of 900more » °C to fluences of 2 × 10 23 or 20 × 10 23 He/m 2. This resulted in a faceted surface structure at the lower fluence or short but well-developed nanofuzz structure at the higher fluence. There was little difference in the hot-rolled or recrystallized material's near-surface (≤50 nm) bubbles at either fluence. At higher fluence and deeper depth, the bubble populations of the hot-rolled and recrystallized were different, the recrystallized being larger and deeper. This may explain previous high-fluence results showing pronounced differences in recrystallized material. The deeper penetration in recrystallized material also implies that grain boundaries are traps, rather than high-diffusivity paths.« less

The EPQ Code System for Simulating the Thermal Response of Plasma-Facing Components to High-Energy Electron Impact

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ward, Robert Cameron; Steiner, Don

2004-06-15

The generation of runaway electrons during a thermal plasma disruption is a concern for the safe and economical operation of a tokamak power system. Runaway electrons have high energy, 10 to 300 MeV, and may potentially cause extensive damage to plasma-facing components (PFCs) through large temperature increases, melting of metallic components, surface erosion, and possible burnout of coolant tubes. The EPQ code system was developed to simulate the thermal response of PFCs to a runaway electron impact. The EPQ code system consists of several parts: UNIX scripts that control the operation of an electron-photon Monte Carlo code to calculate themore » interaction of the runaway electrons with the plasma-facing materials; a finite difference code to calculate the thermal response, melting, and surface erosion of the materials; a code to process, scale, transform, and convert the electron Monte Carlo data to volumetric heating rates for use in the thermal code; and several minor and auxiliary codes for the manipulation and postprocessing of the data. The electron-photon Monte Carlo code used was Electron-Gamma-Shower (EGS), developed and maintained by the National Research Center of Canada. The Quick-Therm-Two-Dimensional-Nonlinear (QTTN) thermal code solves the two-dimensional cylindrical modified heat conduction equation using the Quickest third-order accurate and stable explicit finite difference method and is capable of tracking melting or surface erosion. The EPQ code system is validated using a series of analytical solutions and simulations of experiments. The verification of the QTTN thermal code with analytical solutions shows that the code with the Quickest method is better than 99.9% accurate. The benchmarking of the EPQ code system and QTTN versus experiments showed that QTTN's erosion tracking method is accurate within 30% and that EPQ is able to predict the occurrence of melting within the proper time constraints. QTTN and EPQ are verified and validated as able to calculate the temperature distribution, phase change, and surface erosion successfully.« less

Device for providing high-intensity ion or electron beam

DOEpatents

McClanahan, Edwin D.; Moss, Ronald W.

1977-01-01

A thin film of a low-thermionic-work-function material is maintained on the cathode of a device for producing a high-current, low-pressure gas discharge by means of sputter deposition from an auxiliary electrode. The auxiliary electrode includes a surface with a low-work-function material, such as thorium, uranium, plutonium or one of the rare earth elements, facing the cathode but at a disposition and electrical potential so as to extract ions from the gas discharge and sputter the low-work-function material onto the cathode. By continuously replenishing the cathode film, high thermionic emissions and ion plasmas can be realized and maintained over extended operating periods.

Refractory clad transient internal probe for magnetic field measurements in high temperature plasmas

NASA Astrophysics Data System (ADS)

Kim, Hyundae; Cellamare, Vincent; Jarboe, Thomas R.; Mattick, Arthur T.

2005-05-01

The transient internal probe (TIP) is a diagnostic for local internal field measurements in high temperature plasmas. A verdet material, which rotates the polarization angle of the laser light under magnetic fields, is launched into a plasma at about 1.8km/s. A linearly polarized Ar+ laser illuminates the probe in transit and the light retroreflected from the probe is analyzed to determine the local magnetic field profiles. The TIP has been used for magnetic field measurements on the helicity injected torus where electron temperature Te⩽80eV. In order to apply the TIP in higher temperature plasmas, refractory clad probes have been developed utilizing a sapphire tube, rear disc, and a MgO window on the front. The high melting points of these refractory materials should allow probe operation at plasma electron temperatures up to Te˜300eV. A retroreflecting probe has also been developed using "catseye" optics. The front window is replaced with a plano-convex MgO lens, and the back surface of the probe is aluminized. This approach reduces spurious polarization effects and provides refractory cladding for the probe entrance face. In-flight measurements of a static magnetic field demonstrate the ability of the clad probes to withstand gun-launch acceleration, and provide high accuracy measurements of magnetic field.

High temperature thermo-physical properties of SPS-ed W-Cu functional gradient materials

NASA Astrophysics Data System (ADS)

Galatanu, Magdalena; Enculescu, Monica; Galatanu, Andrei

2018-02-01

The divertor of a fusion reactor like DEMO requires materials able to withstand high heat fluxes and neutron irradiation for several years. For the water cooling concept of this essential part of the reactor, the most likely plasma facing material will be W, while the heatsink material considered is CuCrZr or an improved version of such a Cu-based alloy. To realize W-Cu alloy joints able to withstand thousands of thermal cycles can be difficult due to the difference between the thermal expansion coefficients of these materials. In this work we investigate the possibility to realize such joints by using W-Cu functional gradient materials (FGMs) produced from nanometric and micrometric metallic powders mixtures and consolidated by spark plasma sintering at about 900 °C. Morphological and thermal properties investigations, performed for typical compositions, shows that the best results are obtained using powders with micrometric dimensions. A resulting 1 mm thick, 3 layers W-Cu FGM produced by this simple method shows a remarkable almost constant thermal conductivity value of 200 W m-1 K-1, from room temperature up to 1000 °C.

The radiation asymmetry in MGI rapid shutdown on J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Tong, Ruihai; Chen, Zhongyong; Huang, Duwei; Cheng, Zhifeng; Zhang, Xiaolong; Zhuang, Ge; J-TEXT Team

2017-10-01

Disruptions, the sudden termination of tokamak fusion plasmas by instabilities, have the potential to cause severe material wall damage to large tokamaks like ITER. The mitigation of disruption damage is an essential part of any fusion reactor system. Massive gas injection (MGI) rapid shutdown is a technique in which large amounts of noble gas are injected into the plasma in order to safely radiate the plasma energy evenly over the entire plasma-facing first wall. However, the radiated energy during the thermal quench (TQ) in massive gas injection (MGI) induced disruptions is found toroidal asymmetric, and the degrees of asymmetry correlate with the gas penetration and MGI induced magnetohydrodynamics (MHD) activities. A toroidal and poloidal array of ultraviolet photodiodes (AXUV) has been developed to investigate the radiation asymmetry on J-TEXT tokamak. Together with the upgraded mirnov probe arrays, the relation between MGI triggered MHD activities with radiation asymmetry is studied.

OFF-Stagnation point testing in plasma facility

NASA Astrophysics Data System (ADS)

Viladegut, A.; Chazot, O.

2015-06-01

Reentry space vehicles face extreme conditions of heat flux when interacting with the atmosphere at hypersonic velocities. Stagnation point heat flux is normally used as a reference for Thermal Protection Material (TPS) design; however, many critical phenomena also occur at off-stagnation point. This paper adresses the implementation of an offstagnation point methodology able to duplicate in ground facility the hypersonic boundary layer over a flat plate model. The first analysis using two-dimensional (2D) computational fluid dynamics (CFD) simulations is carried out to understand the limitations of this methodology when applying it in plasma wind tunnel. The results from the testing campaign at VKI Plasmatron are also presented.

A flowing liquid lithium limiter for the Experimental Advanced Superconducting Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ren, J.; Zuo, G. Z.; Hu, J. S.

2015-02-15

A program involving the extensive and systematic use of lithium (Li) as a “first,” or plasma-facing, surface in Tokamak fusion research devices located at Institute of Plasma Physics, Chinese Academy of Sciences, was started in 2009. Many remarkable results have been obtained by the application of Li coatings in Experimental Advanced Superconducting Tokamak (EAST) and liquid Li limiters in the HT-7 Tokamak—both located at the institute. In furtherance of the lithium program, a flowing liquid lithium (FLiLi) limiter system has been designed and manufactured for EAST. The design of the FLiLi limiter is based on the concept of a thinmore » flowing film which was previously tested in HT-7. Exploiting the capabilities of the existing material and plasma evaluation system on EAST, the limiter will be pre-wetted with Li and mechanically translated to the edge of EAST during plasma discharges. The limiter will employ a novel electro-magnetic pump which is designed to drive liquid Li flow from a collector at the bottom of limiter into a distributor at its top, and thus supply a continuously flowing liquid Li film to the wetted plasma-facing surface. This paper focuses on the major design elements of the FLiLi limiter. In addition, a simulation of incoming heat flux has shown that the distribution of heat flux on the limiter surface is acceptable for a future test of power extraction on EAST.« less

A gas-puff-driven theta pinch for plasma-surface interaction studies

NASA Astrophysics Data System (ADS)

Jung, Soonwook; Kesler, Leigh; Yun, Hyun-Ho; Curreli, Davide; Andruczyk, Daniel; Ruzic, David

2012-10-01

DEVeX is a theta pinch device used to investigate fusion-related material interaction such as vapor shielding and ICRF antenna interactions with plasma-pulses in a laboratory setting. The simulator is required to produce high heat-flux plasma enough to induce temperature gradient high enough to study extreme conditions happened in a plasma fusion reactor. In order to achieve it, DEVeX is reconfigured to be combined with gas puff system as gas puffing may reduce heat flux loss resulting from collisions with neutral. A gas puff system as well as a conical gas nozzle is manufactured and several diagnostics including hot wire anemometer and fast ionization gauge are carried out to quantitatively estimate the supersonic flow of gas. Energy deposited on the target for gas puffing and static-filled conditions is measured with thermocouples and its application to TELS, an innovative concept utilizing a thermoelectric-driven liquid metal flow for plasma facing component, is discussed.

Edge and divertor plasma: detachment, stability, and plasma-wall interactions

NASA Astrophysics Data System (ADS)

Krasheninnikov, S. I.; Kukushkin, A. S.; Lee, Wonjae; Phsenov, A. A.; Smirnov, R. D.; Smolyakov, A. I.; Stepanenko, A. A.; Zhang, Yanzeng

2017-10-01

The paper presents an overview of the results of studies on a wide range of the edge plasma related issues. The rollover of the plasma flux to the target during progressing detachment process is shown to be caused by the increase of the impurity radiation loss and volumetric plasma recombination, whereas the ion-neutral friction, although important for establishing the necessary edge plasma conditions, does not contribute per se to the rollover of the plasma flux to the target. The processes limiting the power loss by impurity radiation are discussed and a simple estimate of this limit is obtained. Different mechanisms of meso-scale thermal instabilities driven by impurity radiation and resulting in self-sustained oscillations in the edge plasma are identified. An impact of sheared magnetic field on the dynamics of the blobs and ELM filaments playing an important role in the edge and SOL plasma transport is discussed. Trapping of He, which is an intrinsic impurity for the fusion plasmas, in the plasma-facing tungsten material is considered. A newly developed model, accounting for the generation of additional He traps caused by He bubble growth, fits all the available experimental data on the layer of nano-bubbles observed in W under irradiation by low energy He plasma.

RECENT PROGRESS IN THE FABRICATION AND CHARACTERIZATION OF DUCTILE-PHASE-TOUGHENED TUNGSTEN LAMINATES FOR PLASMA-FACING MATERIALS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cunningham, Kevin; Odette, G Robert; Fields, Kirk A.

2015-09-23

A promising approach to increasing the fracture toughness of W-alloys is ductile-phase toughening (DPT). A ductile phase reinforcement in a brittle matrix increases toughness primarily by crack bridging. A W-Cu laminate was fabricated and the properties of the constituent metals were characterized along with those for the composite. Development of a design model for large-scale crack bridging continued.

Unraveling wall conditioning effects on plasma facing components in NSTX-U with the Materials Analysis Particle Probe (MAPP)

DOE PAGES

Bedoya, F.; Allain, J. P.; Kaita, R.; ...

2016-07-14

A novel PFC diagnostic, the Materials Analysis Particle Probe (MAPP), has been recently commissioned in the National Spherical Torus Experiment Upgrade (NSTX-U). MAPP is currently monitoring the chemical evolution of the PFCs in the NSTX-U lower divertor at 107 cm from the tokamak axis on a day-to-day basis. Here in this work, we summarize the methodology that was adopted to obtain qualitative and quantitative descriptions of the samples chemistry. Using this methodology, we were able to describe all the features in all our spectra to within a standard deviation of ±0.22 eV in position and ±248 s -1 eV inmore » area. Additionally, we provide an example of this methodology with data of boronized ATJ graphite exposed to NSTX-U plasmas.« less

The impact of the fast ion fluxes and thermal plasma loads on the design of the ITER fast ion loss detector

NASA Astrophysics Data System (ADS)

Kocan, M.; Garcia-Munoz, M.; Ayllon-Guerola, J.; Bertalot, L.; Bonnet, Y.; Casal, N.; Galdon, J.; Garcia-Lopez, J.; Giacomin, T.; Gonzalez-Martin, J.; Gunn, J. P.; Rodriguez-Ramos, M.; Reichle, R.; Rivero-Rodriguez, J. F.; Sanchis-Sanchez, L.; Vayakis, G.; Veshchev, E.; Vorpahl, C.; Walsh, M.; Walton, R.

2017-12-01

Thermal plasma loads to the ITER Fast Ion Loss Detector are studied for QDT = 10 burning plasma equilibrium using the 3D field line tracing. The simulations are performed for a FILD insertion 9-13 cm past the port plasma facing surface, optimized for fast ion measurements, and include the worst-case perturbation of the plasma boundary and the error in the magnetic reconstruction. The FILD head is exposed to superimposed time-averaged ELM heat load, static inter-ELM heat flux and plasma radiation. The study includes the estimate of the instantaneous temperature rise due to individual 0.6 MJ controlled ELMs. The maximum time-averaged surface heat load is lesssim 12 MW/m2 and will lead to increase of the FILD surface temperature well below the melting temperature of the materials considered here, for the FILD insertion time of 0.2 s. The worst-case instantaneous temperature rise during controlled 0.6 MJ ELMs is also significantly smaller than the melting temperature of e.g. Tungsten or Molybdenum, foreseen for the FILD housing.

Characterization of boronized graphite in NSTX-U and its effect on plasma performance

NASA Astrophysics Data System (ADS)

Bedoya, Felipe; Allain, Jean Paul; Kaita, Robert; Skinner, Charles; University of Illinois Team; Princeton Plasma Physics Laboratory Collaboration

2017-10-01

Plasma Facing Components (PFC) conditioning can have a crucial influence in plasma performance in tokamak machines. The National Spherical Torus Experiment (NSTX-U) used boronization as the main wall conditioning technique during the FY16 experimental campaign. The Materials Analysis Particle Probe (MAPP), a characterization facility, was used to investigate the surface of ATJ graphite exposed to boronization and plasma in the tokamak using X-ray Photoelectron Spectroscopy (XPS). The measurements showed that plasma induced oxidation plays a critical role in the chemical evolution of the surfaces and as a consequence in plasma performance. Additionally, ex-vessel in-situ laboratory experiments and post-mortem studies of extracted NSTX-U tiles were performed to complement the observations made with MAPP, including controlled D irradiations and XPS depth profiles. These three methodologies show congruent results where D exposures increase the oxygen concentration between 20-30%, highlighting the influence of these two species on the chemistry of the samples. USDOE Contract DE-AC02-09CH11466, USDOE Contract DE-SC0010717 and Award Number DE-SC0012890.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wood, Mitchell; Thompson, Aidan P.

The purpose of this short contribution is to report on the development of a Spectral Neighbor Analysis Potential (SNAP) for tungsten. We have focused on the characterization of elastic and defect properties of the pure material in order to support molecular dynamics simulations of plasma-facing materials in fusion reactors. A parallel genetic algorithm approach was used to efficiently search for fitting parameters optimized against a large number of objective functions. In addition, we have shown that this many-body tungsten potential can be used in conjunction with a simple helium pair potential1 to produce accurate defect formation energies for the W-Hemore » binary system.« less

Some problems of brazing technology for the divertor plate manufacturing

NASA Astrophysics Data System (ADS)

Prokofiev, Yu. G.; Barabash, V. R.; Khorunov, V. F.; Maksimova, S. V.; Gervash, A. A.; Fabritsiev, S. A.; Vinokurov, V. F.

1992-09-01

Among the different design options of the ITER reactor divertor, the joints of the carbon-based materials and molybdenum alloys and joints of tungsten and copper alloys are considered. High-temperature brazing is one of the most promising joining methods for the plasma facing and heat sink materials. The use of brazing for creation of W-Cu and graphite-Mo joints are given here. In addition, the investigation results of microstructure, microhardness and mechanical properties of the joints are presented. For W-Cu samples an influence of the neutron irradiation on the joining strength was studied.

Global gas balance and influence of atomic hydrogen irradiation on the wall inventory in steady-state operation of QUEST tokamak

NASA Astrophysics Data System (ADS)

Kuzmin, A.; Zushi, H.; Takagi, I.; Sharma, S. K.; Rusinov, A.; Inoue, Y.; Hirooka, Y.; Zhou, H.; Kobayashi, M.; Sakamoto, M.; Hanada, K.; Yoshida, N.; Nakamura, K.; Fujisawa, A.; Matsuoka, K.; Idei, H.; Nagashima, Y.; Hasegawa, M.; Onchi, T.; Banerjee, S.; Mishra, K.

2015-08-01

Hydrogen wall pumping is studied in steady state tokamak operation (SSTO) of QUEST with all metal plasma facing materials PFMs at 100 °C. The duration of SSTO is up to 820 s in fully non-inductive plasma. Global gas balance analysis shows that wall pumping at the apparent (retention-release) rate of 1-6 × 1018 H/s is dominant and 70-80% of injected H2 can be retained in PFMs. However, immediately after plasma termination the H2 release rate enhances to ∼1019 H/s. In order to understand a true retention process the direct measurement of retention flux has been carried out by permeation probes. The comparison between the evaluated wall retention and results from global analysis is discussed.

GITR Simulation of Helium Exposed Tungsten Erosion and Redistribution in PISCES-A

NASA Astrophysics Data System (ADS)

Younkin, T. R.; Green, D. L.; Doerner, R. P.; Nishijima, D.; Drobny, J.; Canik, J. M.; Wirth, B. D.

2017-10-01

The extreme heat, charged particle, and neutron flux / fluence to plasma facing materials in magnetically confined fusion devices has motivated research to understand, predict, and mitigate the associated detrimental effects. Of relevance to the ITER divertor is the helium interaction with the tungsten divertor, the resulting erosion and migration of impurities. The linear plasma device PISCES A has performed dedicated experiments for high (4x10-22 m-2s-1) and low (4x10-21 m-2s-1) flux, 250 eV He exposed tungsten targets to assess the net and gross erosion of tungsten and volumetric transport. The temperature of the target was held between 400 and 600 degrees C. We present results of the erosion / migration / re-deposition of W during the experiment from the GITR (Global Impurity Transport) code coupled to materials response models. In particular, the modeled and experimental W I emission spectroscopy data for the 429.4 nm wavelength and net erosion through target and collector mass difference measurements are compared. Overall, the predictions are in good agreement with experiments. This material is supported by the US DOE, Office of Science, Office of Fusion Energy Sciences and Office of Advanced Scientific Computing Research through the SciDAC program on Plasma-Surface Interactions.

Isomer and Fluorination Effects among Fluorine Substituted Hydrocarbon C3/C4 Molecules in Electron Impact Ionization

NASA Astrophysics Data System (ADS)

Patel, U. R.; Joshipura, K. N.

2015-05-01

Electron collision processes are very important in both man-made and natural plasmas, for determining the energy balances and transport properties of electrons. Electron -molecule scattering leading to ionization represents one of the most fundamental processes in collision physics. In the gas phase, the total efficiency of the process is described by the absolute total electron impact ionization cross section. Carbon based materials are some of the widely used materials for a divertor plate and magnetically confined fusion devices. In the ``ITER,'' it is very important for steady state operation to have an estimate of the lifetime of carbon plasma facing components. Apart from fusion plasma relevance, the present theoretical study is very important in modeling and controlling other electron assisted processes in many areas. Hydrocarbons play an important role for plasma diagnostics as impurities in the Tokamak fusion divertor, as seed gases for the production of radicals and ions in low temperature plasma processing. Fluorine substituted hydrocarbons (perfluorocarbons) are important as reactants in plasma assisted fabrication processes. In the present work, we have calculated total ionization cross sections Qion for C3/C4 Hydrocarbon isomers by electron impact, and comparisons are made mutually to observe isomer effect. Comparisons are also made by substituting H atom by F atom and revealing fluorination effect. The present calculations are quite significant owing to the lack of experimental data, with just an isolated previous theoretical work in some cases.

Material Surface Characteristics and Plasma Performance in the Lithium Tokamak Experiment

NASA Astrophysics Data System (ADS)

Lucia, Matthew James

The performance of a tokamak plasma and the characteristics of the surrounding plasma facing component (PFC) material surfaces strongly influence each other. Despite this relationship, tokamak plasma physics has historically been studied more thoroughly than PFC surface physics. The disparity is particularly evident in lithium PFC research: decades of experiments have examined the effect of lithium PFCs on plasma performance, but the understanding of the lithium surface itself is much less complete. This latter information is critical to identifying the mechanisms by which lithium PFCs affect plasma performance. This research focused on such plasma-surface interactions in the Lithium Tokamak Experiment (LTX), a spherical torus designed to accommodate solid or liquid lithium as the primary PFC. Surface analysis was accomplished via the novel Materials Analysis and Particle Probe (MAPP) diagnostic system. In a series of experiments on LTX, the MAPP x-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS) capabilities were used for in vacuo interrogation of PFC samples. This represented the first application of XPS and TDS for in situ surface analysis of tokamak PFCs. Surface analysis indicated that the thin (dLi ˜ 100nm) evaporative lithium PFC coatings in LTX were converted to Li2O due to oxidizing agents in both the residual vacuum and the PFC substrate. Conversion was rapid and nearly independent of PFC temperature, forming a majority Li2O surface within minutes and an entirely Li2O surface within hours. However, Li2O PFCs were still capable of retaining hydrogen and sequestering impurities until the Li2 O was further oxidized to LiOH, a process that took weeks. For hydrogen retention, Li2O PFCs retained H+ from LTX plasma discharges, but no LiH formation was observed. Instead, results implied that H+ was only weakly-bound, such that it almost completely outgassed as H 2 within minutes. For impurity sequestration, LTX plasma performance---ascertained from plasma current and density measurements---progressively improved as plasma carbon and oxygen impurity levels fell. This was true for PFC conditioning by vacuum baking and argon glow discharge cleaning, as well as by lithium evaporation. Some evidence suggested that impurity sequestration was more important than hydrogen retention in enhancing LTX plasma performance. In contrast with expectations for lithium PFCs, heating the Li2 O PFCs in LTX caused increased plasma impurity levels that tended to reduce plasma performance.

Plasma Wall interaction in the IGNITOR machine

NASA Astrophysics Data System (ADS)

Ferro, C.

1998-11-01

One of the critical issues in ignited machines is the management of the heat and particle exhaust without degradation of the plasma quality (pollution and confinement time) and without damage of the material facing the plasma. The IGNITOR machine has been conceived as a ``limiter" device, i.e., with the plasma leaning nearly on the entire surface of the first wall. Peak heat loads can easily be maintained at values lower than 1.35 MW/m^2 even considering displacements of the plasma column^1. This ``limiter" choice is based on the operational performances of high density, high field machines which suggests that intrinsic physics processes in the edge of the plasma are effective in spreading heat loads and maintaining the plasma pollution at a low level. The possibility of these operating scenarios has been demonstrated recently by different machines both in limiter and divertor configurations. The basis for the different physical processes that are expected to influence the IGNITOR edge parameters ^2 are discussed and a comparison with the latest experimental results is given. ^1 C. Ferro, G. Franzoni, R. Zanino, ENEA Internal Report RT/ERG/FUS/94/14. ^2 C. Ferro, R. Zanino, J. Nucl. Mater. 543, 176 (1990).

Temporal and spatial dynamics of optical emission from laser ablation of the first wall materials of fusion device

NASA Astrophysics Data System (ADS)

Dongye, ZHAO; Cong, LI; Yong, WANG; Zhiwei, WANG; Liang, GAO; Zhenhua, HU; Jing, WU; Guang-Nan, LUO; Hongbin, DING

2018-01-01

Laser-induced breakdown spectroscopy (LIBS) has been developed to in situ diagnose the chemical compositions of the first wall in the EAST tokamak. However, the dynamics of optical emission of the key plasma-facing materials, such as tungsten, molybdenum and graphite have not been investigated in a laser produced plasma (LPP) under vacuum. In this work, the temporal and spatial dynamics of optical emission were investigated using the spectrometer with ICCD. Plasma was produced by an Nd:YAG laser (1064 nm) with pulse duration of 6 ns. The results showed that the typical lifetime of LPP is less than 1.4 μs, and the lifetime of ions is shorter than atoms at ˜10-6 mbar. Temporal features of optical emission showed that the optimized delay times for collecting spectra are from 100 to 400 ns which depended on the corresponding species. For spatial distribution, the maximum LIBS spectral intensity in plasma plume is obtained in the region from 1.5 to 3.0 mm above the sample surface. Moreover, the plasma expansion velocity involving the different species in a multicomponent system was measured for obtaining the proper timing (gate delay time and gate width) of the maximum emission intensity and for understanding the plasma expansion mechanism. The order of expansion velocities for various species is {V}{{{C}}+}> {V}{{H}}> {V}{{{Si}}+}> {V}{{Li}}> {V}{{Mo}}> {V}{{W}}. These results could be attributed to the plasma sheath acceleration and mass effect. In addition, an optimum signal-to-background ratio was investigated by varying both delay time and detecting position.

Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U

DOE PAGES

Jaworski, M. A.; Brooks, A.; Kaita, R.; ...

2016-08-08

Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physicsmore » and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. As a result, two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.« less

Developing the science and technology for the Material Plasma Exposure eXperiment

NASA Astrophysics Data System (ADS)

Rapp, J.; Biewer, T. M.; Bigelow, T. S.; Caneses, J. F.; Caughman, J. B. O.; Diem, S. J.; Goulding, R. H.; Isler, R. C.; Lumsdaine, A.; Beers, C. J.; Bjorholm, T.; Bradley, C.; Canik, J. M.; Donovan, D.; Duckworth, R. C.; Ellis, R. J.; Graves, V.; Giuliano, D.; Green, D. L.; Hillis, D. L.; Howard, R. H.; Kafle, N.; Katoh, Y.; Lasa, A.; Lessard, T.; Martin, E. H.; Meitner, S. J.; Luo, G.-N.; McGinnis, W. D.; Owen, L. W.; Ray, H. B.; Shaw, G. C.; Showers, M.; Varma, V.; the MPEX Team

2017-11-01

Linear plasma generators are cost effective facilities to simulate divertor plasma conditions of present and future fusion reactors. They are used to address important R&D gaps in the science of plasma material interactions and towards viable plasma facing components for fusion reactors. Next generation plasma generators have to be able to access the plasma conditions expected on the divertor targets in ITER and future devices. The steady-state linear plasma device MPEX will address this regime with electron temperatures of 1-10 eV and electron densities of 1021{\\text{}}-1020 m-3 . The resulting heat fluxes are about 10 MW m-2 . MPEX is designed to deliver those plasma conditions with a novel Radio Frequency plasma source able to produce high density plasmas and heat electron and ions separately with electron Bernstein wave (EBW) heating and ion cyclotron resonance heating with a total installed power of 800 kW. The linear device Proto-MPEX, forerunner of MPEX consisting of 12 water-cooled copper coils, has been operational since May 2014. Its helicon antenna (100 kW, 13.56 MHz) and EC heating systems (200 kW, 28 GHz) have been commissioned and 14 MW m-2 was delivered on target. Furthermore, electron temperatures of about 20 eV have been achieved in combined helicon and ECH heating schemes at low electron densities. Overdense heating with EBW was achieved at low heating powers. The operational space of the density production by the helicon antenna was pushed up to 1.1 × 1020 m-3 at high magnetic fields of 1.0 T at the target. The experimental results from Proto-MPEX will be used for code validation to enable predictions of the source and heating performance for MPEX. MPEX, in its last phase, will be capable to expose neutron-irradiated samples. In this concept, targets will be irradiated in ORNL’s High Flux Isotope Reactor and then subsequently exposed to fusion reactor relevant plasmas in MPEX.

Effects of ELMs on ITER divertor armour materials

NASA Astrophysics Data System (ADS)

Zhitlukhin, A.; Klimov, N.; Landman, I.; Linke, J.; Loarte, A.; Merola, M.; Podkovyrov, V.; Federici, G.; Bazylev, B.; Pestchanyi, S.; Safronov, V.; Hirai, T.; Maynashev, V.; Levashov, V.; Muzichenko, A.

2007-06-01

This paper is concerned with investigation of an erosion of the ITER-like divertor plasma facing components under plasma heat loads expected during the Type I ELMs in ITER. These experiments were carried out on plasma accelerator QSPA at the SRC RF TRINITI under EU/RF collaboration. Targets were exposed by series repeated plasma pulses with heat loads in a range of 0.5-1.5 MJ/m2 and pulse duration 0.5 ms. Erosion of CFC macrobrushes was determined mainly by sublimation of PAN-fibres that was less than 2.5 μm per pulse. The CFC erosion was negligible at the energy density less than 0.5 MJ/m2 and was increased to the average value 0.3 μm per pulse at 1.5 MJ/m2. The pure tungsten macrobrushes erosion was small in the energy range of 0.5-1.3 MJ/m2. The sharp growth of tungsten erosion and the intense droplet ejection were observed at the energy density of 1.5 MJ/m2.

Development of GEM detector for plasma diagnostics application: simulations addressing optimization of its performance

NASA Astrophysics Data System (ADS)

Chernyshova, M.; Malinowski, K.; Kowalska-Strzęciwilk, E.; Czarski, T.; Linczuk, P.; Wojeński, A.; Krawczyk, R. D.

2017-12-01

The advanced Soft X-ray (SXR) diagnostics setup devoted to studies of the SXR plasma emissivity is at the moment a highly relevant and important for ITER/DEMO application. Especially focusing on the energy range of tungsten emission lines, as plasma contamination by W and its transport in the plasma must be understood and monitored for W plasma-facing material. The Gas Electron Multiplier, with a spatial and energy-resolved photon detecting chamber, based SXR radiation detection system under development by our group may become such a diagnostic setup considering and solving many physical, technical and technological aspects. This work presents the results of simulations aimed to optimize a design of the detector's internal chamber and its performance. The study of the effect of electrodes alignment allowed choosing the gap distances which maximizes electron transmission and choosing the optimal magnitudes of the applied electric fields. Finally, the optimal readout structure design was identified suitable to collect a total formed charge effectively, basing on the range of the simulated electron cloud at the readout plane which was in the order of ~ 2 mm.

Overview of decade-long development of plasma-facing components at ASIPP

NASA Astrophysics Data System (ADS)

Luo, G.-N.; Liu, G. H.; Li, Q.; Qin, S. G.; Wang, W. J.; Shi, Y. L.; Xie, C. Y.; Chen, Z. M.; Missirlian, M.; Guilhem, D.; Richou, M.; Hirai, T.; Escourbiac, F.; Yao, D. M.; Chen, J. L.; Wang, T. J.; Bucalossi, J.; Merola, M.; Li, J. G.; EAST Team

2017-06-01

The first EAST (Experimental Advanced Superconducting Tokamak) plasma ignited in 2006 with non-actively cooled steel plates as the plasma-facing materials and components (PFMCs) which were then upgraded into full graphite tiles bolted onto water-cooled copper heat sinks in 2008. The first wall was changed further into molybdenum alloy in 2012, while keeping the graphite for both the upper and lower divertors. With the rapid increase in heating and current driving power in EAST, the W/Cu divertor project was launched around the end of 2012, aiming at achieving actively cooled full W/Cu-PFCs for the upper divertor, with heat removal capability up to 10 MW m-2. The W/Cu upper divertor was finished in the spring of 2014, consisting of 80 cassette bodies toroidally assembled. Commissioning of the EAST upper W/Cu divertor in 2014 was unsatisfactory and then several practical measures were implemented to improve the design, welding quality and reliability, which helped us achieve successful commissioning in the 2015 Spring Campaign. In collaboration with the IO and CEA teams, we have demonstrated our technological capability to remove heat loads of 5000 cycles at 10 MW m-2 and 1000 cycles at 20 MW m-2 for the small scale monoblock mockups, and surprisingly over 300 cycles at 20 MW m-2 for the flat-tile ones. The experience and lessons we learned from batch production and commissioning are undoubtedly valuable for ITER (International Thermonuclear Experimental Reactor) engineering validation and tungsten-related plasma physics.

Impact of combined hydrogen plasma and transient heat loads on the performance of tungsten as plasma facing material

NASA Astrophysics Data System (ADS)

Wirtz, M.; Bardin, S.; Huber, A.; Kreter, A.; Linke, J.; Morgan, T. W.; Pintsuk, G.; Reinhart, M.; Sergienko, G.; Steudel, I.; De Temmerman, G.; Unterberg, B.

2015-11-01

Experiments were performed in three different facilities in order to investigate the impact of combined steady state deuterium plasma exposure and ELM-like thermal shock events on the performance of ultra high purity tungsten. The electron beam facility JUDITH 1 was used to simulate pure thermal loads. In addition the linear plasma devices PSI-2 and Pilot-PSI have been used for successive as well as simultaneous exposure where the transient heat loads were applied by a high energy laser and the pulsed plasma operation, respectively. The results show that the damage behaviour strongly depends on the loading conditions and the sequence of the particle and heat flux exposure. This is due to hydrogen embrittlement and/or a higher defect concentration in the tungsten near surface region due to supersaturation of hydrogen. The different results in terms of damage formation from both linear plasma devices indicate that also the plasma parameters such as particle energy, flux and fluence, plasma impurities and the pulse shape have a strong influence on the damage performance. In addition, the different loading methods such as the scanning with the electron beam in contrast to the homogeneous exposure by the laser leads to an faster increase of the surface roughness due to plastic deformation.

Direct depth distribution measurement of deuterium in bulk tungsten exposed to high-flux plasma

DOE PAGES

Taylor, Chase N.; Shimada, M.

2017-05-08

Understanding tritium retention and permeation in plasma-facing components is critical for fusion safety and fuel cycle control. Glow discharge optical emission spectroscopy (GD-OES) is shown to be an effective tool to reveal the depth profile of deuterium in tungsten. Results confirm the detection of deuterium. Furthermore, a ~46 µm depth profile revealed that the deuterium content decreased precipitously in the first 7 µm, and detectable amounts were observed to depths in excess of 20 µm. The large probing depth of GD-OES (up to 100s of µm) enables studies not previously accessible to the more conventional techniques for investigating deuterium retention.more » Of particular applicability is the use of GD-OES to measure the depth profile for experiments where high diffusion is expected: deuterium retention in neutron irradiated materials, and ultra-high deuterium fluences in burning plasma environment.« less

Melt damage simulation of W-macrobrush and divertor gaps after multiple transient events in ITER

NASA Astrophysics Data System (ADS)

Bazylev, B. N.; Janeschitz, G.; Landman, I. S.; Loarte, A.; Pestchanyi, S. E.

2007-06-01

Tungsten in the form of macrobrush structure is foreseen as one of two candidate materials for the ITER divertor and dome. In ITER, even for moderate and weak ELMs when a thin shielding layer does not protect the armour surface from the dumped plasma, the main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. The melt erosion of W-macrobrush targets with different geometry of brush surface under the heat loads caused by weak ELMs is numerically investigated using the modified code MEMOS. The optimal angle of brush surface inclination that provides a minimum of surface roughness is estimated for given inclination angles of impacting plasma stream and given parameters of the macrobrush target. For multiple disruptions the damage of the dome gaps and the gaps between divertor cassettes caused by the radiation impact is estimated.

Tungsten nitride coatings obtained by HiPIMS as plasma facing materials for fusion applications

NASA Astrophysics Data System (ADS)

Tiron, Vasile; Velicu, Ioana-Laura; Porosnicu, Corneliu; Burducea, Ion; Dinca, Paul; Malinský, Petr

2017-09-01

In this work, tungsten nitride coatings with nitrogen content in the range of 19-50 at% were prepared by reactive multi-pulse high power impulse magnetron sputtering as a function of the argon and nitrogen mixture and further exposed to a deuterium plasma jet. The elemental composition, morphological properties and physical structure of the samples were investigated by Rutherford backscattering spectrometry, atomic force microscopy and X-ray diffraction. Deuterium implantation was performed using a deuterium plasma jet and its retention in nitrogen containing tungsten films was investigated using thermal desorption spectrometry. Deuterium retention and release behaviour strongly depend on the nitrogen content in the coatings and the films microstructure. All nitride coatings have a polycrystalline structure and retain a lower deuterium level than the pure tungsten sample. Nitrogen content in the films acts as a diffusion barrier for deuterium and leads to a higher desorption temperature, therefore to a higher binding energy.

Blistering behavior and deuterium retention in tungsten vanadium alloys exposed to deuterium plasma in the linear plasma device STEP

NASA Astrophysics Data System (ADS)

Wang, Jun; Cheng, Long; Yuan, Yue; Qin, Shao-Yang; Arshad, Kameel; Guo, Wang-Guo; Wang, Zheng; Zhou, Zhang-Jian; Lu, Guang-Hong

2018-03-01

The behavior of tungsten-vanadium (W-V) alloys fabricated by powder metallurgy as a plasma facing material has been studied. W-V alloys with different vanadium concentrations (5 and 10 wt %) manufactured by hot pressing (HP) were exposed to deuterium plasma (flux ∼4.6 × 1021 m-2s-1, fluence ∼5.6 × 1025 m-2, ion energy ∼60 eV, target temperature ∼450 K) in the linear plasma device STEP at Beihang University. Three typical grains are observed on HP sintered W-V alloys and exhibit a significant effect on its performance under deuterium plasma irradiation. Surface blistering only occurs at W-enriched grains and is significantly mitigated in W-V alloys, especially in W-10 V, blistering is completely suppressed. On the other hand, deuterium retention dramatically increases in the W-V alloys due to vanadium addition. The deuterium retention in W-5 wt. % V is about 6.2 times more than that in rolled pure W, and this factor further increases to 6.9 when the V concentration rises to 10 wt %. We ascribe these phenomena to the changes of microstructures and components caused by vanadium addition.

Probe manipulators for Wendelstein 7-X and their interaction with the magnetic topology

NASA Astrophysics Data System (ADS)

M, RACK; D, HÖSCHEN; D, REITER; B, UNTERBERG; J, W. COENEN; S, BREZINSEK; O, NEUBAUER; S, BOZHENKOV; G, CZYMEK; Y, LIANG; M, HUBENY; Ch, LINSMEIER; the Wendelstein 7-X Team

2018-05-01

Probe manipulators are a versatile addition to typical plasma edge diagnostics. Equipped with material samples they allow for detailed investigation of plasma–wall interaction processes, such as material erosion, deposition or impurity transport pathways. When combined with electrical probes, a study of scrape-off layer and plasma edge density, temperature and flow profiles as well as magnetic topologies is possible. A mid-plane manipulator is already in operation on Wendelstein 7-X. A system in the divertor region is currently under development. In the present paper we discuss the critical issue of heat and power loads, power redistribution and experimental access to the complex magnetic topology of Wendelstein 7-X. All the aforementioned aspects are of relevance for the design and operation of a probe manipulator in a device like Wendelstein 7-X. A focus is put on the topological region that is accessible for the different coil current configurations at Wendelstein 7-X and the power load on the manipulator with respect to the resulting different magnetic configurations. Qualitative analysis of power loads on plasma-facing components is performed using a numerical tracer particle diffusion tool provided via the Wendelstein 7-X Webservices.

Design of a Microwave Assisted Discharge Inductive Plasma Accelerator

NASA Technical Reports Server (NTRS)

Hallock, Ashley K.; Polzin, Kurt A.

2010-01-01

A new plasma accelerator concept that employs electrodeless plasma preionization and pulsed inductive acceleration is presented. Preionization is achieved through an electron cyclotron resonance discharge that produces a weakly-ionized plasma at the face of a conical theta pinch-shaped inductive coil. The presence of the preionized plasma allows for current sheet formation at lower discharge voltages than those found in other pulsed inductive accelerators. The location of an electron cyclotron resonance discharge can be controlled through the design of the applied magnetic field in the thruster. A finite-element model of the magnetic field was used as a design tool, allowing for the implementation of an arrangement of permanent magnets that yields a small volume of preionized propellant at the coil face. This allows for current sheet formation at the face of the inductive coil, minimizing the initial inductance of the pulse circuit and maximizing the potential efficiency of the new accelerator.

A dynamic monitoring approach for the surface morphology evolution measurement of plasma facing components by means of speckle interferometry

NASA Astrophysics Data System (ADS)

Wang, Hongbei; Cui, Xiaoqian; Feng, Chunlei; Li, Yuanbo; Zhao, Mengge; Luo, Guangnan; Ding, Hongbin

2017-11-01

Plasma Facing Components (PFCs) in a magnetically confined fusion plasma device will be exposed to high heat load and particle fluxes, and it would cause PFCs' surface morphology to change due to material erosion and redeposition from plasma wall interactions. The state of PFCs' surface condition will seriously affect the performance of long-pulse or steady state plasma discharge in a tokamak; it will even constitute an enormous threat to the operation and the safety of fusion plasma devices. The PFCs' surface morphology evolution measurement could provide important information about PFCs' real-time status or damage situation and it would help to a better understanding of the plasma wall interaction process and mechanism. Meanwhile through monitoring the distribution of dust deposition in a tokamak and providing an upper limit on the amount of loose dust, the PFCs' surface morphology measurement could indirectly contribute to keep fusion operational limits and fusion device safety. Aiming at in situ dynamic monitoring PFCs' surface morphology evolution, a laboratory experimental platform DUT-SIEP (Dalian University of Technology-speckle interferometry experimental platform) based on the speckle interferometry technique has been constructed at Dalian University of Technology (DUT) in China. With directional specific designing and focusing on the real detection condition of EAST (Experimental Advanced Superconducting Tokamak), the DUT-SIEP could realize a variable measurement range, widely increased from 0.1 μm to 300 μm, with high spatial resolution (<1 mm) and ultra-high time resolution (<2 s for EAST measuring conditions). Three main components of the DUT-SIEP are all integrated and synchronized by a time schedule control and data acquisition terminal and coupled with a three-dimensional phase unwrapping algorithm, the surface morphology information of target samples can be obtained and reconstructed in real-time. A local surface morphology of the real divertor tiles adopted from EAST has been measured, and the feasibility and reliability of this new experimental platform have been demonstrated.

Cryogenic Considerations for Superconducting Magnet Design for the Material Plasma Exposure eXperiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Duckworth, Robert C; Demko, Dr. Jonathan A; Lumsdaine, Arnold

2015-01-01

In order to determine long term performance of plasma facing components such as diverters and first walls for fusion devices, next generation plasma generators are needed. A Material Plasma Exposure eXperiment (MPEX) has been proposed to address this need through the generation of plasmas in front of the target with electron temperatures of 1-15 eV and electron densities of 1020 to 1021 m-3. Heat fluxes on target diverters could reach 20 MW/m2. In order generate this plasma, a unique radio frequency helicon source and heating of electrons and ions through Electron Bernstein Wave (EBW) and Ion Cyclotron Resonance Heating (ICRH)more » has been proposed. MPEX requires a series of magnets with non-uniform central fields up to 2 T over a 5m length in the heating and transport region and 1 T uniform central field over a 1-m length on a diameter of 1.3 m. Given the field requirements, superconducting magnets are under consideration for MPEX. In order to determine the best construction method for the magnets, the cryogenic refrigeration has been analyzed with respect to cooldown and operational performance criteria for open-cycle and closed-cycle systems, capital and operating costs of these system, and maturity of supporting technology such as cryocoolers. These systems will be compared within the context of commercially available magnet constructions to determine the most economical method for MPEX operation. The current state of the MPEX magnet design including details on possible superconducting magnet configurations will be presented.« less

Investigation of high power impulse magnetron sputtering (HIPIMS) discharge using fast ICCD camera

NASA Astrophysics Data System (ADS)

Hecimovic, Ante

2012-10-01

High power impulse magnetron sputtering (HIPIMS) combines impulse glow discharges at power levels up to the MW range with conventional magnetron cathodes to achieve a highly ionised sputtered flux. The dynamics of the HIPIMS discharge was investigated using fast Intensified Charge Coupled Device (ICCD) camera. In the first experiment the HIPIMS plasma was recorded from the side with goal to analyse the plasma intensity using Abel inversion to obtain the emissivity maps of the plasma species. Resulting emissivity maps provide the information on the spatial distribution of Ar and sputtered material and evolution of the plasma chemistry above the cathode. In the second experiment the plasma emission was recorded with camera facing the target. The images show that the HIPIMS plasma develops drift wave type instabilities characterized by well defined regions of high and low plasma emissivity along the racetrack of the magnetron. The instabilities cause periodic shifts in the floating potential. The structures rotate in ExB direction at velocities of 10 kms-1 and frequencies up to 200 kHz. The high emissivity regions comprise Ar and metal ion emission with strong Ar and metal neutral emission depletion. A detailed analysis of the temporal evolution of the saturated instabilities using four consequently triggered fast ICCD cameras is presented. Furthermore working gas pressure and discharge current variation showed that the shape and the speed of the instability strongly depend on the working gas and target material combination. In order to better understand the mechanism of the instability, different optical interference band pass filters (of metal and gas atom, and ion lines) were used to observe the spatial distribution of each species within the instability.

Digital Holography for in Situ Real-Time Measurement of Plasma-Facing-Component Erosion

DOE Office of Scientific and Technical Information (OSTI.GOV)

ThomasJr., C. E.; Granstedt, E. M.; Biewer, Theodore M

2014-01-01

In situ, real time measurement of net plasma-facing-component (PFC) erosion/deposition in a real plasma device is challenging due to the need for good spatial and temporal resolution, sufficient sensitivity, and immunity to fringe-jump errors. Design of a high-sensitivity, potentially high-speed, dual-wavelength CO2 laser digital holography system (nominally immune to fringe jumps) for PFC erosion measurement is discussed.

Plasma treatment of fiber facets for increased (de)mating endurance in physical contact fiber connectors

NASA Astrophysics Data System (ADS)

Van Erps, Jürgen; Voss, Kevin; De Witte, Martijn; Radulescu, Radu; Beri, Stefano; Watté, Jan; Thienpont, Hugo

2016-04-01

It is known that cleaving an optical fiber introduces a number of irregularities and defects to the fiber's end-face, such as hackles and shockwaves. These defects can act as failure initiators when stress is applied to the end-face. Given the fiber's small diameter of 125 ffm, a large amount of mechanical stress can be expected to be applied on its end-face during the mating-demating cycle. In addition, a connector in a fiber-to-the-home (FTTH) network can be expected to be mated and demated more than 30 times during its lifetime for purposes such as testing, churning, or provisioning. For this reason, the performance of a connector that displays low optical loss when first installed can dramatically degrade after few mating-demating cycles and catastrophic connector failure due to end-face breakage is likely. We present plasma discharge shaping of cleaved fiber tips to strongly improve the endurance of the fibers to repeated mating-demating cycles. We quantify the dependency of the plasma-induced surface curvature of the fiber tip on the plasma duration and on the position of the fiber tip within the plasma cloud. Finally we present data showing the improved endurance of fibers that are exposed to plasma compared to conventional as-cleaved fibers.

Sputtering effects on mirrors made of different tungsten grades

NASA Astrophysics Data System (ADS)

Voitsenya, V. S.; Ogorodnikova, O. V.; Bardamid, A. F.; Bondarenko, V. N.; Konovalov, V. G.; Lytvyn, P. M.; Marot, L.; Ryzhkov, I. V.; Shtan', A. F.; Skoryk, O. O.; Solodovchenko, S. I.

2018-03-01

Because tungsten (W) is used in present fusion devices and it is a reference material for ITER divertor and possible plasma-facing material for DEMO, we strive to understand the response of different W grades to ion bombardment. In this study, we investigated the behavior of mirrors made of four polycrystalline W grades under long-term ion sputtering. Argon (Ar) and deuterium (D) ions extracted from a plasma were used to investigate the effect of projectile mass on surface modification. Depending on the ion fluence, the reflectance measured at normal incidence was very different for different W grades. The lowest degradation rate of the reflectance was measured for the mirror made of recrystallized W. The highest degradation rate was found for one of the ITER-grade W samples. Pre-irradiation of a mirror with 20-MeV W6+ ions, as simulation of neutron irradiation in ITER, had no noticeable influence on reflectance degradation under sputtering with either Ar or D ions.

Trends in Dielectric Etch for Microelectronics Processing

NASA Astrophysics Data System (ADS)

Hudson, Eric A.

2003-10-01

Dielectric etch technology faces many challenges to meet the requirements for leading-edge microelectronics processing. The move to sub 100-nm device design rules increases the aspect ratios of certain features, imposes tighter restrictions on etched features' critical dimensions, and increases the density of closely packed arrays of features. Changes in photolithography are driving transitions to new photoresist materials and novel multilayer resist methods. The increasing use of copper metallization and low-k interlayer dielectric materials has introduced dual-damascene integration methods, with specialized dielectric etch applications. A common need is the selective removal of multiple layers which have very different compositions, while maintaining close control of the etched features' profiles. To increase productivity, there is a growing trend toward in-situ processing, which allows several films to be successively etched during a single pass through the process module. Dielectric etch systems mainly utilize capacitively coupled etch reactors, operating with medium-density plasmas and low gas residence time. Commercial technology development increasingly relies upon plasma diagnostics and modeling to reduce development cycle time and maximize performance.

RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Part I: Theory and description of model capabilities

NASA Astrophysics Data System (ADS)

Raffray, A. René; Federici, Gianfranco

1997-04-01

RACLETTE (Rate Analysis Code for pLasma Energy Transfer Transient Evaluation), a comprehensive but relatively simple and versatile model, was developed to help in the design analysis of plasma facing components (PFCs) under 'slow' high power transients, such as those associated with plasma vertical displacement events. The model includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the PFC block thermal response and the coolant behaviour. This paper represents part I of two sister and complementary papers. It covers the model description, calibration and validation, and presents a number of parametric analyses shedding light on and identifying trends in the PFC armour block response to high plasma energy deposition transients. Parameters investigated include the plasma energy density and deposition time, the armour thickness and the presence of vapour shielding effects. Part II of the paper focuses on specific design analyses of ITER plasma facing components (divertor, limiter, primary first wall and baffle), including improvements in the thermal-hydraulic modeling required for better understanding the consequences of high energy deposition transients in particular for the ITER limiter case.

Ring-Opening Polymerization of Cyclic Hemiacetal Esters for the Preparation of Hydrolytically and Thermally Degradable Polymers

NASA Astrophysics Data System (ADS)

Neitzel, Angelika Susanne Elisabeth

During the course of tokamak operation, material is routinely eroded from plasma facing components and transported to other regions of the machine. This net-reshaping process will lead to many challenges in a high duty cycle magnetic fusion reactor, and is also highly relevant to the wall conditioning process in current experiments. Proper modeling of this mechanism requires a global treatment of the entire tokamak, and integration of tightly coupled plasma and surface processes. This thesis focuses on extending and applying the WallDYN mixed-material migration code [1] [2], which couples local erosion and deposition processes with plasma impurity transport in a non-iterative, self-consistent manner that maintains overall material balance. NSTX-U operated in 2016 with carbon PFCs, periodically conditioned with boron-containing films to suppress oxygen impurities. However, oxygen levels tended to return to a pre-conditioned state following repeated plasma exposure, and this occurred on a faster time scale when conditioning with less boron. This C/B/O migration is interpretively modeled with WallDYN, which successfully reproduces observed trends in oxygen evolution. A new model for spatially inhomogenous mixed material films has been developed for WallDYN, which allows for the differentiation between conditioning films of varying thicknesses. A boron coverage model for the NSTX-U glow discharge boronization process is also developed. These new capabilities improve WallDYN agreement with observed NSTX-U spectroscopic data by at least a factor of 2. As part of the integrated model, plasma backgrounds representing NSTX-U H-modes and L-modes are calculated using OSM-EIRENE, constrained by a combination of NSTX-U data and NSTX SOLPS calculations. The effect of modifying the assumed parallel SOL profile is examined, with the result that inner divertor-directed flows turn the outer divertor from a region of net boron deposition to one of net boron erosion. Plasma impurity transport calculations are carried out with DIVIMP, and mixed-material sputtering calculations are carried out for a range of possible surfaces with SDTRIMSP. WallDYN modeling of C/Li/O migration in NSTX is presented, utilizing OSM-EIRENE calculations of lithiated NSTX plasmas. An adatom model of temperature-enhanced sputtering has been added to WallDYN, and the effect of various surface temperature scenarios is examined. A sensitivity study of surface binding energies used in WallDYN sputtering calculations is carried out, finding that mixed material effects become dominant when the system contains both tightly- and weakly- bound elements (such as C and Li).

DOE Office of Scientific and Technical Information (OSTI.GOV)

Webster, A. J.; Morris, J.; Todd, T. N.

A unique sequence of 120 almost identical plasmas in the Joint European Torus (JET) recently provided two orders of magnitude more statistically equivalent data than ever previously available. The purpose was to study movement of eroded plasma-facing material from JET's new Beryllium wall, but it has allowed the statistical detection of otherwise unobservable phenomenon. This includes a sequence of resonant-like waiting times between edge-localised plasma instabilities (ELMs), instabilities that must be mitigated or avoided in large magnetically confined plasmas such as those planned for ITER. Here, we investigate the cause of this phenomenon, using the unprecedented quantity of data tomore » produce a detailed picture of the plasma's behaviour. After combining the data, oscillations are clearly observable in the plasma's vertical position, in edge losses of ions, and in Beryllium II (527 nm) light emissions. The oscillations are unexpected, are not obvious in data from a single pulse alone, and are all clearly correlated with each other. They are likely to be caused by a small vertical oscillation that the plasma control system is not reacting to prevent, but a more complex explanation is possible. The clearly observable but unexpected link between small changes in the plasma's position and changes to edge-plasma transport and stability suggest that these characteristics cannot always be studied in isolation. It also suggests new opportunities for ELM mitigation and control that may exist.« less

Micro/nano composited tungsten material and its high thermal loading behavior

NASA Astrophysics Data System (ADS)

Fan, Jinglian; Han, Yong; Li, Pengfei; Sun, Zhiyu; Zhou, Qiang

2014-12-01

Tungsten (W) is considered as promising candidate material for plasma facing components (PFCs) in future fusion reactors attributing to its many excellent properties. Current commercial pure tungsten material in accordance with the ITER specification can well fulfil the performance requirements, however, it has defects such as coarse grains, high ductile-brittle transition temperature (DBTT) and relatively low recrystallization temperature compared with its using temperature, which cannot meet the harsh wall loading requirement of future fusion reactor. Grain refinement has been reported to be effective in improving the thermophysical and mechanical properties of W. In this work, rare earth oxide (Y2O3/La2O3) and carbides (TiC/ZrC) were used as dispersion phases to refine W grains, and micro/nano composite technology with a process of "sol gel - heterogeneous precipitation - spray drying - hydrogen reduction - ordinary consolidation sintering" was invented to introduce these second-phase particles uniformly dispersed into W grains and grain-boundaries. Via this technology, fine-grain W materials with near-full density and relatively high mechanical properties compared with traditional pure W material were manufactured. Preliminary transient high-heat flux tests were performed to evaluate the thermal response under plasma disruption conditions, and the results show that the W materials prepared by micro/nano composite technology can endure high-heat flux of 200 MW/m2 (5 ms).

Material Surface Characteristics and Plasma Performance in the Lithium Tokamak Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lucia, Matthew James

The performance of a tokamak plasma and the characteristics of the surrounding plasma facing component (PFC) material surfaces strongly influence each other. Despite this relationship, tokamak plasma physics has historically been studied more thoroughly than PFC surface physics. The disparity is particularly evident in lithium PFC research: decades of experiments have examined the effect of lithium PFCs on plasma performance, but the understanding of the lithium surface itself is much less complete. This latter information is critical to identifying the mechanisms by which lithium PFCs affect plasma performance. This research focused on such plasma-surface interactions in the Lithium Tokamak Experimentmore » (LTX), a spherical torus designed to accommodate solid or liquid lithium as the primary PFC. Surface analysis was accomplished via the novel Materials Analysis and Particle Probe (MAPP) diagnostic system. In a series of experiments on LTX, the MAPP x-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS) capabilities were used for in vacuo interrogation of PFC samples. This represented the first application of XPS and TDS for in situ surface analysis of tokamak PFCs. Surface analysis indicated that the thin (d ~ 100nm) evaporative lithium PFC coatings in LTX were converted to Li2O due to oxidizing agents in both the residual vacuum and the PFC substrate. Conversion was rapid and nearly independent of PFC temperature, forming a majority Li2O surface within minutes and an entirely Li2O surface within hours. However, Li2O PFCs were still capable of retaining hydrogen and sequestering impurities until the Li2O was further oxidized to LiOH, a process that took weeks. For hydrogen retention, Li2O PFCs retained H+ from LTX plasma discharges, but no LiH formation was observed. Instead, results implied that H+ was only weakly-bound, such that it almost completely outgassed as H2 within minutes. For impurity sequestration, LTX plasma performance—ascertained from plasma current and density measurements—progressively improved as plasma carbon and oxygen impurity levels fell. This was true for PFC conditioning by vacuum baking and argon glow discharge cleaning, as well as by lithium evaporation. Some evidence suggested that impurity sequestration was more important than hydrogen retention in enhancing LTX plasma performance. In contrast with expectations for lithium PFCs, heating the Li2O PFCs in LTX caused increased plasma impurity levels that tended to reduce plasma performance.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bourham, Mohamed A.; Gilligan, John G.

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing componentsmore » safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.« less

Overview of the JET results with the ITER-like wall

NASA Astrophysics Data System (ADS)

Romanelli, F.; EFDA Contributors, JET

2013-10-01

Following the completion in May 2011 of the shutdown for the installation of the beryllium wall and the tungsten divertor, the first set of JET campaigns have addressed the investigation of the retention properties and the development of operational scenarios with the new plasma-facing materials. The large reduction in the carbon content (more than a factor ten) led to a much lower Zeff (1.2-1.4) during L- and H-mode plasmas, and radiation during the burn-through phase of the plasma initiation with the consequence that breakdown failures are almost absent. Gas balance experiments have shown that the fuel retention rate with the new wall is substantially reduced with respect to the C wall. The re-establishment of the baseline H-mode and hybrid scenarios compatible with the new wall has required an optimization of the control of metallic impurity sources and heat loads. Stable type-I ELMy H-mode regimes with H98,y2 close to 1 and βN ˜ 1.6 have been achieved using gas injection. ELM frequency is a key factor for the control of the metallic impurity accumulation. Pedestal temperatures tend to be lower with the new wall, leading to reduced confinement, but nitrogen seeding restores high pedestal temperatures and confinement. Compared with the carbon wall, major disruptions with the new wall show a lower radiated power and a slower current quench. The higher heat loads on Be wall plasma-facing components due to lower radiation made the routine use of massive gas injection for disruption mitigation essential.

Evidence of formation of lithium compounds on FTU tiles and dust

NASA Astrophysics Data System (ADS)

Ghezzi, F.; Laguardia, L.; Apicella, M. L.; Bressan, C.; Caniello, R.; Cippo, E. Perelli; Conti, C.; De Angeli, M.; Maddaluno, G.; Mazzitelli, G.

2018-01-01

Since 2006 lithium as an advanced plasma facing material has been tested on the Frascati Tokamak Upgrade (FTU). Lithium in the liquid phase acts both as plasma facing component, i.e. limiter, and plays also a role in plasma operation because by depositing a lithium film on the walls (lithization) oxygen is gettered. As in all deposition processes, even for the lithization, the presence of impurities in plasma phase strongly affects the properties of the deposited film. During the 2008 campaigns of FTU it was observed a strong release of carbon dioxide (during disruptions), resulting in successive serious difficulty of operation. In order to find the possible reactions occurred, we have analyzed the surface of two tiles of the toroidal limiter close to the Liquid Lithium Limiter (LLL). The presence of molybdenum oxides and carbides suggested that the surface temperatures could have exceeded 1000 K, likely during disruptions. lithium oxides and hydroxides have been found on the tiles and in the dust collected in the vessel, confirming the presence of LiO and LiOH and a not negligible concentration of Li2CO3 especially at the LLL location. On the basis of the above results, we propose here a simple rationale, based on a two reactions mechanism, which can explain the formation of Li2CO3 and its subsequent decomposition during disruption with release of CO2 in the vessel. Admitting surface temperatures above 1000 K during a disruption, relatively high partial pressures of CO2 are also predicted by the equilibrium constant for Li2CO3 decomposition.

First-principles study of solvent-solute mixed dumbbells in body-centered-cubic tungsten crystals

NASA Astrophysics Data System (ADS)

Suzudo, Tomoaki; Tsuru, Tomohito; Hasegawa, Akira

2018-07-01

Tungsten (W) is considered as a promising candidate for plasma-facing materials for future nuclear fusion devices, and selecting optimal alloying constituents is a critical issue to improve radiation resistance of the W alloys as well as to improve their mechanical properties. We conducted in the current study a series of first-principles calculations for investigating solvent-solute mixed dumbbells in W crystals. The results suggested that titanium (Ti), vanadium (V), and chromium (Cr) are favorable as solutes for W alloys from irradiation-effect perspectives because these elements are expected to promote vacancy-interstitial recombination without causing radiation-induced precipitation that reduces ductility of irradiated materials.

Material properties and their influence on the behaviour of tungsten as plasma facing material

NASA Astrophysics Data System (ADS)

Wirtz, M.; Uytdenhouwen, I.; Barabash, V.; Escourbiac, F.; Hirai, T.; Linke, J.; Loewenhoff, Th.; Panayotis, S.; Pintsuk, G.

2017-06-01

With the aim of a possible improvement of the material specification for tungsten, five different tungsten products by different companies and by different production technologies (forging and rolling) are subject to a materials characterization program. Tungsten produced by forging results in an uniaxial elongated grain shape while rolled products have a plate like grain shape which has an influence on the mechanical properties of the material. The materials were investigated with respect to the following parameters: hardness measurements, microstructural investigations, tensile tests and recrystallisation sensitivity tests at 3 different temperatures. The obtained results show that different production processes have an influence on the resulting anisotropic microstructure and the related material properties of tungsten in the as-received state. Additionally, the recrystallization sensitivity varies between the different products, what could be a result of the different production processes. Additionally, two tungsten products were exposed to thermal shocks. The obtained results show that the improved recrystallisation behaviour has no major impact on the thermal shock performance.

Secondary electron emission from plasma-generated nanostructured tungsten fuzz

DOE PAGES

Patino, M.; Raitses, Y.; Wirz, R.

2016-11-14

Recently, several researchers (e.g., Q. Yang, Y.-W. You, L. Liu, H. Fan, W. Ni, D. Liu, C. S. Liu, G. Benstetter, and Y. Wang, Scientific Reports 5, 10959 (2015)) have shown that tungsten fuzz can grow on a hot tungsten surface under bombardment by energetic helium ions in different plasma discharges and applications, including magnetic fusion devices with plasma facing tungsten components. This work reports direct measurements of the total effective secondary electron emission (SEE) from tungsten fuzz. Using dedicated material surface diagnostics and in-situ characterization, we find two important results: (1) SEE values for tungsten fuzz are 40-63% lowermore » than for smooth tungsten and (2) the SEE values for tungsten fuzz are independent of the angle of the incident electron. The reduction in SEE from tungsten fuzz is most pronounced at high incident angles, which has important implications for many plasma devices since in a negative-going sheath the potential structure leads to relatively high incident angles for the electrons at the plasma confining walls. Overall, low SEE will create a relatively higher sheath potential difference that reduces plasma electron energy loss to the confining wall. Thus the presence or self-generation in a plasma of a low SEE surface such as tungsten fuzz can be desirable for improved performance of many plasma devices.:7px« less

Evaluating optical hazards from plasma arc cutting.

PubMed

Glassford, Eric; Burr, Gregory

2018-01-01

The Health Hazard Evaluation Program of the National Institute for Occupational Safety and Health evaluated a steel building materials manufacturer. The employer requested the evaluation because of concerns about optical radiation hazards from a plasma arc cutting system and the need to clarify eye protection requirements for plasma operators, other employees, and visitors. The strength of the ultraviolet radiation, visible radiation (light), and infrared radiation generated by the plasma arc cutter was measured at various distances from the source and at different operating amperages. Investigators also observed employees performing the plasma arc cutting. Optical radiation above safe levels for the unprotected eyes in the ultraviolet-C, ultraviolet-B, and visible light ranges were found during plasma arc cutting. In contrast, infrared and ultraviolet-A radiation levels during plasma arc cutting were similar to background levels. The highest non-ionizing radiation exposures occurred when no welding curtains were used. A plasma arc welding curtain in place did not eliminate optical radiation hazards to the plasma arc operator or to nearby employees. In most instances, the measured intensities for visible light, UV-C, and UV-B resulted in welding shade lens numbers that were lower than those stipulated in the OSHA Filter Lenses for Protection Against Radiant Energy table in 29 CFR 1910.133(a)(5). [1] Investigators recommended using a welding curtain that enclosed the plasma arc, posting optical radiation warning signs in the plasma arc cutter area, installing audible or visual warning cues when the plasma arc cutter was operating, and using welding shades that covered the plasma arc cutter operator's face to protect skin from ultraviolet radiation hazards.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Kun; Bannister, Mark E.; Meyer, Fred W.

Here, in a magnetic fusion energy (MFE) device, the plasma-facing materials (PFMs) will be subjected to tremendous fluxes of ions, heat, and neutrons. The response of PFMs to the fusion environment is still not well defined. Tungsten metal is the present candidate of choice for PFM applications such as the divertor in ITER. However, tungsten's microstructure will evolve in service, possibly to include recrystallization. How tungsten's response to plasma exposure evolves with changes in microstructure is presently unknown. In this work, we have exposed hot-worked and recrystallized tungsten to an 80 eV helium ion beam at a temperature of 900more » °C to fluences of 2 × 10 23 or 20 × 10 23 He/m 2. This resulted in a faceted surface structure at the lower fluence or short but well-developed nanofuzz structure at the higher fluence. There was little difference in the hot-rolled or recrystallized material's near-surface (≤50 nm) bubbles at either fluence. At higher fluence and deeper depth, the bubble populations of the hot-rolled and recrystallized were different, the recrystallized being larger and deeper. This may explain previous high-fluence results showing pronounced differences in recrystallized material. The deeper penetration in recrystallized material also implies that grain boundaries are traps, rather than high-diffusivity paths.« less

RF H-minus ion source development in China spallation neutron source

NASA Astrophysics Data System (ADS)

Chen, W.; Ouyang, H.; Xiao, Y.; Liu, S.; Lü, Y.; Cao, X.; Huang, T.; Xue, K.

2017-08-01

China Spallation Neutron Source (CSNS) phase-I project currently uses a Penning surface plasma H- ion source, which has a life time of several weeks with occasional sparks between high voltage electrodes. To extend the life time of the ion source and prepare for the CSNS phase-II, we are trying to develop a RF negative hydrogen ion source with external antenna. The configuration of the source is similar to the DESY external antenna ion source and SNS ion source. However several changes are made to improve the stability and the life time. Firstly, Si3N4 ceramic with high thermal shock resistance, and high thermal conductivity is used for plasma chamber, which can endure an average power of 2000W. Secondly, the water-cooled antenna is brazed on the chamber to improve the energy efficiency. Thirdly, cesium is injected directly to the plasma chamber if necessary, to simplify the design of the converter and the extraction. Area of stainless steel exposed to plasma is minimized to reduce the sputtering and degassing. Instead Mo, Ta, and Pt coated materials are used to face the plasma, which makes the self-cleaning of the source possible.

Redefinition of the self-bias voltage in a dielectrically shielded thin sheath RF discharge

NASA Astrophysics Data System (ADS)

Ho, Teck Seng; Charles, Christine; Boswell, Rod

2018-05-01

In a geometrically asymmetric capacitively coupled discharge where the powered electrode is shielded from the plasma by a layer of dielectric material, the self-bias manifests as a nonuniform negative charging in the dielectric rather than on the blocking capacitor. In the thin sheath regime where the ion transit time across the powered sheath is on the order of or less than the Radiofrequency (RF) period, the plasma potential is observed to respond asymmetrically to extraneous impedances in the RF circuit. Consequently, the RF waveform on the plasma-facing surface of the dielectric is unknown, and the behaviour of the powered sheath is not easily predictable. Sheath circuit models become inadequate for describing this class of discharges, and a comprehensive fluid, electrical, and plasma numerical model is employed to accurately quantify this behaviour. The traditional definition of the self-bias voltage as the mean of the RF waveform is shown to be erroneous in this regime. Instead, using the maxima of the RF waveform provides a more rigorous definition given its correlation with the ion dynamics in the powered sheath. This is supported by a RF circuit model derived from the computational fluid dynamics and plasma simulations.

Avoiding Tokamak Disruptions by Applying Static Magnetic Fields That Align Locked Modes with Stabilizing Wave-Driven Currents [Avoiding Tokamak Disruptions by Magnetically Aligning Locked Modes with Stabilizing Wave-Driven Currents

DOE PAGES

Volpe, F. A.; Hyatt, Alan; La Haye, Robert J.; ...

2015-10-19

The international ITER tokamak has the objective of demonstrating the scientific feasibility of magnetic confinement fusion as a source of energy. A concern towards the achievement of this goal is represented by major disruptions: complete losses of confinement often initiated by a non-rotating ('locked') magnetic island created by magnetic reconnection. During disruptions, energy and particles accumulated in the plasma volume over many seconds are lost in a few milliseconds and released on the plasma-facing materials. In addition, multi-MA level currents flowing in the tokamak plasma for its sustainment and confinement are lost, also in milliseconds, thus terminating the plasma dischargemore » and causing electromagnetic stresses that, if unmitigated, could lead to excessive device wear. Moreover it is shown that magnetic perturbations can be used to avoid disruptions by "guiding" the magnetic island to lock in a position where it is accessible to millimetre wave beams that fully stabilize it.« less

Overview of experimental preparation for the ITER-Like Wall at JET

NASA Astrophysics Data System (ADS)

Jet Efda Contributors Brezinsek, S.; Fundamenski, W.; Eich, T.; Coad, J. P.; Giroud, C.; Huber, A.; Jachmich, S.; Joffrin, E.; Krieger, K.; McCormick, K.; Lehnen, M.; Loarer, T.; de La Luna, E.; Maddison, G.; Matthews, G. F.; Mertens, Ph.; Nunes, I.; Philipps, V.; Riccardo, V.; Rubel, M.; Stamp, M. F.; Tsalas, M.

2011-08-01

Experiments in JET with carbon-based plasma-facing components have been carried out in preparation of the ITER-Like Wall with beryllium main chamber and full tungsten divertor. The preparatory work was twofold: (i) development of techniques, which ensure safe operation with the new wall and (ii) provision of reference plasmas, which allow a comparison of operation with carbon and metallic wall. (i) Compatibility with the W divertor with respect to energy loads could be achieved in N2 seeded plasmas at high densities and low temperatures, finally approaching partial detachment, with only moderate confinement reduction of 10%. Strike-point sweeping increases the operational space further by re-distributing the load over several components. (ii) Be and C migration to the divertor has been documented with spectroscopy and QMBs under different plasma conditions providing a database which will allow a comparison of the material transport to remote areas with metallic walls. Fuel retention rates of 1.0-2.0 × 1021 D s-1 were obtained as references in accompanied gas balance studies.

Behavior of W-SiC/SiC dual layer tiles under LHD plasma exposure

NASA Astrophysics Data System (ADS)

Mohrez, Waleed A.; Kishimoto, Hirotatsu; Kohno, Yutaka; Hirotaki, S.; Kohyama, Akira

2013-11-01

Towards the early realization of fusion power reactors, high performance first wall and plasma facing components (PFCs) are essentially required. As one of the biggest challenges for this, high heat flux component (HHFC) design and R & D has been emphasized. This report provides the high performance HHFC materials R & D status and the first plasma exposure test result from large helical device (LHD). W-SiC/SiC dual layer tiles (hereafter, W-SiC/SiC) were developed by applied NITE process. This is the realistic concept of tungsten armor with ceramic composite substrates for fusion power reactors. The dual layer tiles were fabricated and tested their survival under the LHD divertor plasma exposure (Nominally 10 MW/m2 maximum heat load for 6 s operation cycle). The microstructure evolution, including crack and pore formation, was analyzed, besides the behavior of bonding layer between tungsten and SiC/SiC was evaluated by C-scanning images of ultrasonic method and Electron probe Micro-analyzer (EPMA). Thermal analysis was conducted by finite element method, where ANSYS code release 13.0 was used.

Electromagnetic effects on dynamics of high-beta filamentary structures

DOE PAGES

Lee, Wonjae; Umansky, Maxim V.; Angus, J. R.; ...

2015-01-12

The impacts of the electromagnetic effects on blob dynamics are considered. Electromagnetic BOUT++ simulations on seeded high-beta blobs demonstrate that inhomogeneity of magnetic curvature or plasma pressure along the filament leads to bending of the blob filaments and the magnetic field lines due to increased propagation time of plasma current (Alfvén time). The bending motion can enhance heat exchange between the plasma facing materials and the inner SOL region. The effects of sheath boundary conditions on the part of the blob away from the boundary are also diminished by the increased Alfvén time. Using linear analysis and the BOUT++ simulation,more » it is found that electromagnetic effects in high temperature and high density plasmas reduce the growth rate of resistive drift wave turbulence when resistivity drops below some certain value. Lastly, in the course of blobs motion in the SOL its temperature is reduced, which leads to enhancement of resistive effects, so the blob can switch from electromagnetic to electrostatic regime, where resistive drift wave turbulence become important.« less

Development of nanostructures on plasma facing components

NASA Astrophysics Data System (ADS)

Ruzic, David; Fiflis, Peter; Kalathiparambil, Kishor Kumar

2015-11-01

Exposure to low temperature helium plasma, with parameters similar to tokamak edge plasmas, have been found to induce the growth of nanostructures on tungsten. These nanostructures results in an increase in the effective surface area, and will alter the physical properties of the components. Although this has several potential applications in the industrial scenario, it is an undesired effect for fusion reactor components, and is hence necessary to understand their growth mechanisms in order to figure out suitable remedial schemes. Work done using a high density, low temperature helicon discharge plasma source with a resistively heated tungsten wire immersed in the discharge as the substrate have demonstrated the well-defined stages of the growth as a function of total fluence. The required fluence was attained by extending the exposure time. Extensive research work has also shown that a variety of other materials are also prone to develop such structures under similar conditions. In the present work, the effect of the experimental conditions on the various stages of structure development will be presented and a comparison between the structures developed on different types of substrates will be shown.

Developing the science and technology for the Material Plasma Exposure eXperiment

DOE PAGES

Rapp, J.; Biewer, T. M.; Bigelow, T. S.; ...

2017-07-27

Linear plasma generators are cost effective facilities to simulate divertor plasma conditions of present and future fusion reactors. They are used to address important R&D gaps in the science of plasma material interactions and towards viable plasma facing components for fusion reactors. Next generation plasma generators have to be able to access the plasma conditions expected on the divertor targets in ITER and future devices. The steady-state linear plasma device MPEX will address this regime with electron temperatures of 1–10 eV and electron densities ofmore » $$10^{21}{\\text{}}\\!-\\!10^{20}$$ $${\\rm m}^{-3}$$. The resulting heat fluxes are about 10 MW $${\\rm m}^{-2}$$ . MPEX is designed to deliver those plasma conditions with a novel Radio Frequency plasma source able to produce high density plasmas and heat electron and ions separately with electron Bernstein wave (EBW) heating and ion cyclotron resonance heating with a total installed power of 800 kW. The linear device Proto-MPEX, forerunner of MPEX consisting of 12 water-cooled copper coils, has been operational since May 2014. Its helicon antenna (100 kW, 13.56 MHz) and EC heating systems (200 kW, 28 GHz) have been commissioned and 14 MW $${\\rm m}^{-2}$$ was delivered on target. Furthermore, electron temperatures of about 20 eV have been achieved in combined helicon and ECH heating schemes at low electron densities. Overdense heating with EBW was achieved at low heating powers. The operational space of the density production by the helicon antenna was pushed up to $$1.1 \\times 10^{20}$$ $${\\rm m}^{-3}$$ at high magnetic fields of 1.0 T at the target. Finally, the experimental results from Proto-MPEX will be used for code validation to enable predictions of the source and heating performance for MPEX. MPEX, in its last phase, will be capable to expose neutron-irradiated samples. In this concept, targets will be irradiated in ORNL's High Flux Isotope Reactor and then subsequently exposed to fusion reactor relevant plasmas in MPEX.« less

Developing the science and technology for the Material Plasma Exposure eXperiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rapp, J.; Biewer, T. M.; Bigelow, T. S.

Linear plasma generators are cost effective facilities to simulate divertor plasma conditions of present and future fusion reactors. They are used to address important R&D gaps in the science of plasma material interactions and towards viable plasma facing components for fusion reactors. Next generation plasma generators have to be able to access the plasma conditions expected on the divertor targets in ITER and future devices. The steady-state linear plasma device MPEX will address this regime with electron temperatures of 1–10 eV and electron densities ofmore » $$10^{21}{\\text{}}\\!-\\!10^{20}$$ $${\\rm m}^{-3}$$. The resulting heat fluxes are about 10 MW $${\\rm m}^{-2}$$ . MPEX is designed to deliver those plasma conditions with a novel Radio Frequency plasma source able to produce high density plasmas and heat electron and ions separately with electron Bernstein wave (EBW) heating and ion cyclotron resonance heating with a total installed power of 800 kW. The linear device Proto-MPEX, forerunner of MPEX consisting of 12 water-cooled copper coils, has been operational since May 2014. Its helicon antenna (100 kW, 13.56 MHz) and EC heating systems (200 kW, 28 GHz) have been commissioned and 14 MW $${\\rm m}^{-2}$$ was delivered on target. Furthermore, electron temperatures of about 20 eV have been achieved in combined helicon and ECH heating schemes at low electron densities. Overdense heating with EBW was achieved at low heating powers. The operational space of the density production by the helicon antenna was pushed up to $$1.1 \\times 10^{20}$$ $${\\rm m}^{-3}$$ at high magnetic fields of 1.0 T at the target. Finally, the experimental results from Proto-MPEX will be used for code validation to enable predictions of the source and heating performance for MPEX. MPEX, in its last phase, will be capable to expose neutron-irradiated samples. In this concept, targets will be irradiated in ORNL's High Flux Isotope Reactor and then subsequently exposed to fusion reactor relevant plasmas in MPEX.« less

Upgraded PMI diagnostic capabilities using Accelerator-based In-situ Materials Surveillance (AIMS) on Alcator C-Mod

NASA Astrophysics Data System (ADS)

Kesler, Leigh; Barnard, Harold; Hartwig, Zachary; Sorbom, Brandon; Lanza, Richard; Terry, David; Vieira, Rui; Whyte, Dennis

2014-10-01

The AIMS diagnostic was developed to rapidly and non-invasively characterize in-situ plasma material interactions (PMI) in a tokamak. Recent improvements are described which significantly expand this measurement capability on Alcator C-Mod. The detection time at each wall location is reduced from about 10 min to 30 s, via improved hardware and detection geometry. Detectors are in an augmented re-entrant tube to maximize the solid angle between detectors and diagnostic locations. Spatial range is expanded by using beam dynamics simulation to design upgraded B-field power supplies to provide maximal poloidal access, including a ~20° toroidal range in the divertor. Measurement accuracy is improved with angular and energy resolved cross section measurements obtained using a separate 0.9 MeV deuteron ion accelerator. Future improvements include the installation of recessed scintillator tiles as beam targets for calibration of the diagnostic. Additionally, implanted depth marker tiles will enable AIMS to observe the in-situ erosion and deposition of high-Z plasma-facing materials. This work is supported by U.S. DOE Grant No. DE-FG02-94ER54235 and Cooperative Agreement No. DE-FC02-99ER54512.

Lithium As Plasma Facing Component for Magnetic Fusion Research

DOE Office of Scientific and Technical Information (OSTI.GOV)

Masayuki Ono

The use of lithium in magnetic fusion confinement experiments started in the 1990's in order to improve tokamak plasma performance as a low-recycling plasma-facing component (PFC). Lithium is the lightest alkali metal and it is highly chemically reactive with relevant ion species in fusion plasmas including hydrogen, deuterium, tritium, carbon, and oxygen. Because of the reactive properties, lithium can provide strong pumping for those ions. It was indeed a spectacular success in TFTR where a very small amount (~ 0.02 gram) of lithium coating of the PFCs resulted in the fusion power output to improve by nearly a factor ofmore » two. The plasma confinement also improved by a factor of two. This success was attributed to the reduced recycling of cold gas surrounding the fusion plasma due to highly reactive lithium on the wall. The plasma confinement and performance improvements have since been confirmed in a large number of fusion devices with various magnetic configurations including CDX-U/LTX (US), CPD (Japan), HT-7 (China), EAST (China), FTU (Italy), NSTX (US), T-10, T-11M (Russia), TJ-II (Spain), and RFX (Italy). Additionally, lithium was shown to broaden the plasma pressure profile in NSTX, which is advantageous in achieving high performance H-mode operation for tokamak reactors. It is also noted that even with significant applications (up to 1,000 grams in NSTX) of lithium on PFCs, very little contamination (< 0.1%) of lithium fraction in main fusion plasma core was observed even during high confinement modes. The lithium therefore appears to be a highly desirable material to be used as a plasma PFC material from the magnetic fusion plasma performance and operational point of view. An exciting development in recent years is the growing realization of lithium as a potential solution to solve the exceptionally challenging need to handle the fusion reactor divertor heat flux, which could reach 60 MW/m2 . By placing the liquid lithium (LL) surface in the path of the main divertor heat flux (divertor strike point), the lithium is evaporated from the surface. The evaporated lithium is quickly ionized by the plasma and the ionized lithium ions can provide a strongly radiative layer of plasma ("radiative mantle"), thus could significantly reduce the heat flux to the divertor strike point surfaces, thus protecting the divertor surface. The protective effects of LL have been observed in many experiments and test stands. As a possible reactor divertor candidate, a closed LL divertor system is described. Finally, it is noted that the lithium applications as a PFC can be quite flexible and broad. The lithium application should be quite compatible with various divertor configurations, and it can be also applied to protecting the presently envisioned tungsten based solid PFC surfaces such as the ones for ITER. Lithium based PFCs therefore have the exciting prospect of providing a cost effective flexible means to improve the fusion reactor performance, while providing a practical solution to the highly challenging divertor heat handling issue confronting the steadystate magnetic fusion reactors.« less

Cellulose microfibrils: visualization of biosynthetic and orienting complexes in association with the plasma membrane.

PubMed

Brown, R M; Montezinos, D

1976-01-01

Cellulose microfibril biosynthesis, assembly, and orientation in the unicellular green alga, Oocystis, is visualized in association with a linear enzyme complex embedded in the B face of the plasma membrane. Granule bands of the A face and complementary ridges of the B face are postulated to assist in the orientation of recently synthesized microfibrils. A model for microfibril synthesis and orientation is proposed and correlated with current hypotheses regarding cellulose biosynthesis in higher plants.

The Effect of Ion Energy and Substrate Temperature on Deuterium Trapping in Tungsten

NASA Astrophysics Data System (ADS)

Roszell, John Patrick Town

Tungsten is a candidate plasma facing material for next generation magnetic fusion devices such as ITER and there are major operational and safety issues associated with hydrogen (tritium) retention in plasma facing components. An ion gun was used to simulate plasma-material interactions under various conditions in order to study hydrogen retention characteristics of tungsten thus enabling better predictions of hydrogen retention in ITER. Thermal Desorption Spectroscopy (TDS) was used to measure deuterium retention from ion irradiation while modelling of TDS spectra with the Tritium Migration Analysis Program (TMAP) was used to provide information about the trapping mechanisms involved in deuterium retention in tungsten. X-ray Photoelectron Spectroscopy (XPS) and Secondary Ion Mass Spectrometry (SIMS) were used to determine the depth resolved composition of specimens used for irradiation experiments. Carbon and oxygen atoms will be among the most common contaminants within ITER. C and O contamination in polycrystalline tungsten (PCW) specimens even at low levels (˜0.1%) was shown to reduce deuterium retention by preventing diffusion of deuterium into the bulk of the specimen. This diffusion barrier was also responsible for the inhibition of blister formation during irradiations at 500 K. These observations may provide possible mitigation techniques for problems associated with tritium retention and mechanical damage to plasma facing components caused by hydrogen implantation. Deuterium trapping in PCW and single crystal tungsten (SCW) was studied as a function of ion energy and substrate temperature. Deuterium retention was shown to decrease with decreasing ion energy below 100 eV/D+. Irradiation of tungsten specimens with 10 eV/D+ ions was shown to retain up to an order of magnitude less deuterium than irradiation with 500 eV/D+ ions. Furthermore, the retention mechanism for deuterium was shown to be consistent across the entire energy range studied (10-500 eV) with the shallow penetration depth of low energy ions being the major factor in the reduction in retention. A change in retention mechanism was observed as tungsten temperature during irradiation was increased from 300 to 500 K. Modelling of deuterium retention in 300 and 500 K SCW specimens revealed that two traps, 1.0 and 1.3 eV, are involved in retention for irradiations performed at 300K while a single 2.1 eV trap is present for 500 K irradiations. Experiments suggest that the 2.1 eV trap is created during irradiation of tungsten at 500 K and this process also involves the annihilation of the 1.3 and 1.0 eV traps.

Interface reactions between silicon carbide and interlayers in silicon carbide copper metal matrix composites

NASA Astrophysics Data System (ADS)

Köck, T.; Brendel, A.; Bolt, H.

2007-05-01

Novel copper matrix composites reinforced with silicon carbide fibres are considered as a new generation of heat sink materials for the divertor of future fusion reactors. The divertor is exposed to intense particle bombardment and heat loads of up to 15 MW m-2. This component consists of the plasma-facing material which is bonded to the actively cooled heat sink. Due to its high thermal conductivity of about 400 W m-1 K-1 copper is a promising material for the heat sink. To increase the mechanical properties of copper at working temperature (823 K), silicon carbide fibres with a diameter of 140 μm are used to reinforce the interface area between the plasma-facing material and the heat sink. Push-out tests show that the adhesion between SiC fibre and Cu matrix without any interlayer is very low. To increase the fibre-matrix bonding the fibres are coated with Cr and W with a thickness of 300-400 nm before Cu deposition by magnetron sputtering. Push-out tests on these modified fibres show a significant increase in adhesion compared to the fibres without interlayer. XRD investigations after a heat treatment at 923 K show a chromium carbide (Cr23C6, Cr3C2) formation and the absence of chromium silicides. In the case of a W interlayer a W2C formation is detected and also no tungsten silicides. Single-fibre tensile tests were performed to investigate the influence of the reaction zone on the ultimate tensile strength of the fibres. The ultimate tensile strength for fibres without interlayer remains constant at about 2200 MPa after annealing at 923 K. The fibres with chromium and tungsten interlayers, respectively, show a decrease of about 30% of the ultimate tensile strength after the heat treatment at 923 K.

Establishing Physical and Engineering Science Base to Bridge from ITER to Demo

NASA Astrophysics Data System (ADS)

Peng, Y.-K. Martin; Abdou, M.; Gates, D.; Hegna, C.; Hill, D.; Najmabadi, F.; Navratil, G.; Parker, R.

2007-11-01

A Nuclear Component Testing (NCT) Discussion Group emerged recently to clarify how ``a lowered-risk, reduced-cost approach can provide a progressive fusion environment beyond the ITER level to explore, discover, and help establish the remaining, critically needed physical and engineering sciences knowledge base for Demo.'' The group, assuming success of ITER and other contemporary projects, identified critical ``gap-filling'' investigations: plasma startup, tritium self-sufficiency, plasma facing surface performance and maintainability, first wall/blanket/divertor materials defect control and lifetime management, and remote handling. Only standard or spherical tokamak plasma conditions below the advanced regime are assumed to lower the anticipated physics risk to continuous operation (˜2 weeks). Modular designs and remote handling capabilities are included to mitigate the risk of component failure and ease replacement. Aspect ratio should be varied to lower the cost, accounting for the contending physics risks and the near-term R&D. Cost and time-effective staging from H-H, D-D, to D-T will also be considered. *Work supported by USDOE.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Perez, Danny, E-mail: danny-perez@lanl.gov; Sandoval, Luis; Voter, Arthur F.

Due to its enviable properties, tungsten is a leading candidate plasma facing material in nuclear fusion reactors. However, like many other metals, tungsten is known to be affected by the high doses of helium atoms incoming from the plasma. Indeed, the implanted interstitial helium atoms cluster together and, upon reaching a critical cluster size, convert into substitutional nanoscale He bubbles. These bubbles then grow by absorbing further interstitial clusters from the matrix. This process can lead to deleterious changes in microstructure, degradation of mechanical properties, and contamination of the plasma. In order to better understand the growth process, we usemore » traditional and accelerated molecular dynamics simulations to investigate the interactions between interstitial He clusters and pre-existing bubbles. These interactions are characterized in terms of thermodynamics and kinetics. We show that the proximity of the bubble leads to an enhancement of the trap mutation rate and, consequently, to the nucleation of satellite bubbles in the neighborhood of existing ones. We also uncover a number of mechanisms that can lead to the subsequent annihilation of such satellite nanobubbles.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Perez, Danny; Sandoval, Luis; Uberuaga, Blas P.

Due to its enviable properties, tungsten is a leading candidate plasma facing material in nuclear fusion reactors. But, like many other metals, tungsten is known to be affected by the high doses of helium atoms incoming from the plasma. Indeed, the implanted interstitial helium atoms cluster together and, upon reaching a critical cluster size, convert into substitutional nanoscale He bubbles. These bubbles then grow by absorbing further interstitial clusters from the matrix. This process can lead to deleterious changes in microstructure, degradation of mechanical properties, and contamination of the plasma. In order to better understand the growth process, we usemore » traditional and accelerated molecular dynamics simulations to investigate the interactions between interstitial He clusters and pre-existing bubbles. These interactions are characterized in terms of thermodynamics and kinetics. We also show that the proximity of the bubble leads to an enhancement of the trap mutation rate and, consequently, to the nucleation of satellite bubbles in the neighborhood of existing ones. Finally, we uncover a number of mechanisms that can lead to the subsequent annihilation of such satellite nanobubbles.« less

Effects of ELMs and disruptions on ITER divertor armour materials

NASA Astrophysics Data System (ADS)

Federici, G.; Zhitlukhin, A.; Arkhipov, N.; Giniyatulin, R.; Klimov, N.; Landman, I.; Podkovyrov, V.; Safronov, V.; Loarte, A.; Merola, M.

2005-03-01

This paper describes the response of plasma facing components manufactured with tungsten (macro-brush) and CFC to energy loads characteristic of Type I ELMs and disruptions in ITER, in experiments conducted (under an EU/RF collaboration) in two plasma guns (QSPA and MK-200UG) at the TRINITI institute. Targets were exposed to a series of repetitive pulses in QSPA with heat loads in a range of 1-2 MJ/m 2 lasting 0.5 ms. Moderate tungsten erosion, of less than 0.2 μm per pulse, was found for loads of ˜1.5 MJ/m 2, consistent with ELM erosion being determined by tungsten evaporation and not by melt layer displacement. At energy densities of ˜1.8 MJ/m 2 a sharp growth of tungsten erosion was measured together with intense droplet ejection. MK-200UG experiments were focused on studying mainly vapor plasma production and impurity transport during ELMs. The conditions for removal of thin metal deposits from a carbon substrate were characterized.

Plasma Inter-Particle and Particle-Wall Interactions

NASA Astrophysics Data System (ADS)

Patino, Marlene Idy

An improved understanding of plasma inter-particle and particle-wall interactions is critical to the advancement of plasma devices used for space electric propulsion, fusion, high-power communications, and next-generation energy systems. Two interactions of particular importance are (1) ion-atom collisions in the plasma bulk and (2) secondary electron emission from plasma-facing materials. For ion-atom collisions, interactions between fast ions and slow atoms are commonly dominated by charge-exchange and momentum-exchange collisions that are important to understanding the performance and behavior of many plasma devices. To investigate this behavior, this work developed a simple, well-characterized experiment that accurately measures the effects of high energy xenon ions incident on a background of xenon neutral atoms. By comparing these results to both analytical and computational models of ion-atom interactions, we discovered the importance of (1) accurately treating the differential cross-sections for momentum-exchange and charge-exchange collisions over all neutral background pressures, and (2) commonly overlooked interactions, including ion-induced electron emission and neutral-neutral ionization collisions, at high pressures. Data provide vital information on the angular scattering distributions of charge-exchange and momentum-exchange ions at 1.5 keV relevant for ion thrusters, and serve as canonical data for validation of plasma models. This work also investigates electron-induced secondary electron emission behavior relevant to materials commonly considered for plasma thrusters, fusion systems, and many other plasma devices. For such applications, secondary electron emission can alter the sheath potential, which can significantly affect device performance and life. Secondary electron emission properties were measured for materials that are critical to the efficient operation of many plasma devices, including: graphite (for tokamaks, ion thrusters, and traveling wave tubes), lithium (for tokamak walls), tungsten (the most promising material for future tokamaks such as ITER), and nickel (for plasma-enhanced chemistry). Measurements were made for incident electron energies up to 1.5 keV and angles between 0 and 78°. The most significant results from these measurements are as follows: (1) first-ever measurements of naturally-forming tungsten fuzz show a more than 40% reduction in secondary electron emission and an independence on incidence angle; (2) original measurements of lithium oxide show a 2x and 6x increase in secondary electron emission for 17% and 100% oxidation; and (3) unique measurements of Ni(110) single crystal show extrema in secondary electron emission when incidence angle is varied and an up to 36% increase at 0° over polycrystalline nickel. Each of these results are important discoveries for improving plasma devices. For example, from (1), the growth of tungsten fuzz in tokamaks is desirable for minimizing adverse secondary electron emission effects. From (2), the opposite is true for tokamaks with lithium coatings which are oxidized by typical residual gases. From (3), secondary electron emission from Ni(110) catalysts in plasma-enhanced chemistry may facilitate further reactions.

Numerical study of slip system activity and crystal lattice rotation under wedge nanoindents in tungsten single crystals

NASA Astrophysics Data System (ADS)

Volz, T.; Schwaiger, R.; Wang, J.; Weygand, S. M.

2018-05-01

Tungsten is a promising material for plasma facing components in future nuclear fusion reactors. In the present work, we numerically investigate the deformation behavior of unirradiated tungsten (a body-centered cubic (bcc) single crystal) underneath nanoindents. A finite element (FE) model is presented to simulate wedge indentation. Crystal plasticity finite element (CPFE) simulations were performed for face-centered and body-centered single crystals accounting for the slip system family {110} <111> in the bcc crystal system and the {111} <110> slip family in the fcc system. The 90° wedge indenter was aligned parallel to the [1 ¯01 ]-direction and indented the crystal in the [0 1 ¯0 ]-direction up to a maximum indentation depth of 2 µm. In both, the fcc and bcc single crystals, the activity of slip systems was investigated and compared. Good agreement with the results from former investigations on fcc single crystals was observed. Furthermore, the in-plane lattice rotation in the material underneath an indent was determined and compared for the fcc and bcc single crystals.

Impact on the deuterium retention of simultaneous exposure of tungsten to a steady state plasma and transient heat cycling loads

NASA Astrophysics Data System (ADS)

Huber, A.; Sergienko, G.; Wirtz, M.; Steudel, I.; Arakcheev, A.; Brezinsek, S.; Burdakov, A.; Dittmar, T.; Esser, H. G.; Kreter, A.; Linke, J.; Linsmeier, Ch; Mertens, Ph; Möller, S.; Philipps, V.; Pintsuk, G.; Reinhart, M.; Schweer, B.; Shoshin, A.; Terra, A.; Unterberg, B.

2016-02-01

The impact on the deuterium retention of simultaneous exposure of tungsten to a steady-state plasma and transient cyclic heat loads has been studied in the linear PSI-2 facility with the main objective of qualifying tungsten (W) as plasma-facing material. The transient heat loads were applied by a high-energy laser, a Nd:YAG laser (λ = 1064 nm) with an energy per pulse of up to 32 J and a duration of 1 ms. A pronounced increase in the D retention by a factor of 13 has been observed during the simultaneous transient heat loads and plasma exposure. These data indicate that the hydrogen clustering is enhanced by the thermal shock exposures, as seen on the increased blister size due to mobilization and thermal production of defects during transients. In addition, the significant increase of the D retention during the simultaneous loads could be explained by an increased diffusion of D atoms into the W material due to strong temperature gradients during the laser pulse exposure and to an increased mobility of D atoms along the shock-induced cracks. Only 24% of the retained deuterium is located inside the near-surface layer (d<4 μm). Enhanced blister formation has been observed under combined loading conditions at power densities close to the threshold for damaging. Blisters are not mainly responsible for the pronounced increase of the D retention.

Non-thermal plasma conversion of hydrocarbons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strohm, James J.; Skoptsov, George L.; Musselman, Evan T.

A non-thermal plasma is generated to selectively convert a precursor to a product. More specifically, plasma forming material and a precursor material are provided to a reaction zone of a vessel. The reaction zone is exposed to microwave radiation, including exposing the plasma forming material and the precursor material to the microwave radiation. The exposure of the plasma forming material to the microwave radiation selectively converts the plasma forming material to a non-thermal plasma including formation of one or more streamers. The precursor material is mixed with the plasma forming material and the precursor material is exposed to the non-thermalmore » plasma including exposing the precursor material to the one or more streamers. The exposure of the precursor material to the streamers and the microwave radiation selectively converts the precursor material to a product.« less

Linear facing target sputtering of the epitaxial Ga-doped ZnO transparent contact layer on GaN-based light-emitting diodes

NASA Astrophysics Data System (ADS)

Shin, Hyun-Su; Lee, Ju-Hyun; Kwak, Joon-Seop; Lee, Hyun Hwi; Kim, Han-Ki

2013-10-01

In this study, we reported on the plasma damage-free sputtering of epitaxial Ga-doped ZnO (GZO) films on the p-GaN layer for use as a transparent contact layer (TCL) for GaN-based light-emitting diodes (LEDs) using linear facing target sputtering (LFTS). Effective confinement of high-density plasma between faced GZO targets and the substrate position located outside of the plasma region led to the deposition of the epitaxial GZO TCL with a low sheet resistance of 25.7 Ω/s and a high transmittance of 84.6% on a p-GaN layer without severe plasma damage, which was found using the conventional dc sputtering process. The low turn-on voltage of the GaN-based LEDs with an LFTS-grown GZO TCL layer that was grown at a longer target-to-substrate distance (TSD) indicates that the plasma damage of the GaN-LED could be effectively reduced by adjusting the TSD during the LFTS process.

Edge-localized-modes in tokamaks

DOE PAGES

Leonard, Anthony W.

2014-09-11

Edge-localized-modes (ELMs) are a ubiquitous feature of H-mode in tokamaks. When gradients in the H-mode transport barrier grow to exceed the MHD stability limit the ELM instability grows explosively rapidly transporting energy and particles onto open field lines and material surfaces. Though ELMs provide additional particle and impurity transport through the H-mode transport barrier, enabling steady operation, the resulting heat flux transients to plasma facing surfaces project to large amplitude in future low collisionality burning plasma tokamaks. Measurements of the ELM heat flux deposition onto material surfaces in the divertor and main chamber indicate significant broadening compared to inter-ELM heatmore » flux, with a timescale for energy deposition that is consistent with sonic ion flow and numerical simulation. Comprehensive ELM simulation is highlighting the important physics processes of ELM transport including parallel transport due to magnetic reconnection and turbulence resulting from collapse of the H-mode transport barrier. As a result, encouraging prospects for ELM control and/or suppression in future tokamaks include intrinsic modes of ELM free operation, ELM triggering with frequent small pellet injection and the application of 3D magnetic fields.« less

Edge-localized-modes in tokamaksa)

NASA Astrophysics Data System (ADS)

Leonard, A. W.

2014-09-01

Edge-localized-modes (ELMs) are a ubiquitous feature of H-mode in tokamaks. When gradients in the H-mode transport barrier grow to exceed the MHD stability limit the ELM instability grows explosively, rapidly transporting energy and particles onto open field lines and material surfaces. Though ELMs provide additional particle and impurity transport through the H-mode transport barrier, enabling steady operation, the resulting heat flux transients to plasma facing surfaces project to large amplitude in future low collisionality burning plasma tokamaks. Measurements of the ELM heat flux deposition onto material surfaces in the divertor and main chamber indicate significant broadening compared to inter-ELM heat flux, with a timescale for energy deposition that is consistent with sonic ion flow and numerical simulation. Comprehensive ELM simulation is highlighting the important physics processes of ELM transport including parallel transport due to magnetic reconnection and turbulence resulting from collapse of the H-mode transport barrier. Encouraging prospects for ELM control and/or suppression in future tokamaks include intrinsic modes of ELM free operation, ELM triggering with frequent small pellet injection and the application of 3D magnetic fields.

Powder Metallurgy Processing of a WxTaTiVCr High-Entropy Alloy and Its Derivative Alloys for Fusion Material Applications.

PubMed

Waseem, Owais Ahmed; Ryu, Ho Jin

2017-05-16

The W x TaTiVCr high-entropy alloy with 32at.% of tungsten (W) and its derivative alloys with 42 to 90at.% of W with in-situ TiC were prepared via the mixing of elemental W, Ta, Ti, V and Cr powders followed by spark plasma sintering for the development of reduced-activation alloys for fusion plasma-facing materials. Characterization of the sintered samples revealed a BCC lattice and a multi-phase structure. The selected-area diffraction patterns confirmed the formation of TiC in the high-entropy alloy and its derivative alloys. It revealed the development of C15 (cubic) Laves phases as well in alloys with 71 to 90at.% W. A mechanical examination of the samples revealed a more than twofold improvement in the hardness and strength due to solid-solution strengthening and dispersion strengthening. This study explored the potential of powder metallurgy processing for the fabrication of a high-entropy alloy and other derived compositions with enhanced hardness and strength.

Design requirements for plasma facing materials in ITER

NASA Astrophysics Data System (ADS)

Matera, R.; Federici, G.; ITER Joint Central Team

1996-10-01

After the official approval of the Interim Design Report, the ITER project enters the final phase of the EDA. With the definition of the design requirements of the high heat flux components, the structural and armor materials' working domain is better specified, allowing to focus the R & D program on the most critical issues and to orient the design of divertor and first wall components towards those concepts which potentially have a better chance to withstand normal and off-normal operating conditions. Among the latter, slow, high-power, high recycling transient are at present driving the design of high heat flux components. Examples of possible design solution under experimental validation in the R & D program are presented and discussed in this paper.

Molecular Dynamics Simulation of Hydrogen Trapping on Sigma 5 Tungsten Grain Boundaries

NASA Astrophysics Data System (ADS)

Al-Shalash, Aws Mohammed Taha

Tungsten as a plasma facing material is the predominant contender for future Tokamak reactor environments. The interaction between the plasma particles and tungsten is crucial to be studied for successful usage and design of tungsten in the plasma facing components ensuring the reliability and longevity of the fusion reactors. The bombardment of the sigma 5 polycrystalline tungsten was modeled using the molecular dynamics simulation through the large-scale atomic/molecular massively parallel simulator (LAMMPS) code and Tersoff type interatomic potential. By simulating the operational conditions of the Tokamak reactors, the hydrogen trapping rate, implantation distribution, and bubble formation was investigated at various temperatures (300-1200 K) and various hydrogen incident energy (20-100 eV). The substrate's temperature increases the deflected H atoms, and increases the penetration depth for the ones that go through. As well, the lower temperature tungsten substrates retain more H atoms. Increasing the bombarded hydrogen's energy increases the trapping and retention rate and the depth of penetration. Another experiments were conducted to determine whether the Sigma5 grain boundary's (GB) location affects the trapping profiles in H. The findings are ranges from small effect on deflection rates at low H energies to no effect at high H energies. However, there is a considerable effect on shifting the trapping depth profile upward toward the surface when raising the GB closer to the surface. Hydrogen atoms are highly mobile on tungsten substrate, yet no bubble formation was witnessed.

Crystal orientation effects on helium ion depth distributions and adatom formation processes in plasma-facing tungsten

DOE PAGES

Hammond, Karl D.; Wirth, Brian D.

2014-10-09

Here, we present atomistic simulations that show the effect of surface orientation on helium depth distributions and surface feature formation as a result of low-energy helium plasma exposure. We find a pronounced effect of surface orientation on the initial depth of implanted helium ions, as well as a difference in reflection and helium retention across different surface orientations. Our results indicate that single helium interstitials are sufficient to induce the formation of adatom/substitutional helium pairs under certain highly corrugated tungsten surfaces, such as {1 1 1}-orientations, leading to the formation of a relatively concentrated layer of immobile helium immediately belowmore » the surface. The energies involved for helium-induced adatom formation on {1 1 1} and {2 1 1} surfaces are exoergic for even a single adatom very close to the surface, while {0 0 1} and {0 1 1} surfaces require two or even three helium atoms in a cluster before a substitutional helium cluster and adatom will form with reasonable probability. This phenomenon results in much higher initial helium retention during helium plasma exposure to {1 1 1} and {2 1 1} tungsten surfaces than is observed for {0 0 1} or {0 1 1} surfaces and is much higher than can be attributed to differences in the initial depth distributions alone. Lastly, the layer thus formed may serve as nucleation sites for further bubble formation and growth or as a source of material embrittlement or fatigue, which may have implications for the formation of tungsten “fuzz” in plasma-facing divertors for magnetic-confinement nuclear fusion reactors and/or the lifetime of such divertors.« less

Plasma-surface interaction in the context of ITER.

PubMed

Kleyn, A W; Lopes Cardozo, N J; Samm, U

2006-04-21

The decreasing availability of energy and concern about climate change necessitate the development of novel sustainable energy sources. Fusion energy is such a source. Although it will take several decades to develop it into routinely operated power sources, the ultimate potential of fusion energy is very high and badly needed. A major step forward in the development of fusion energy is the decision to construct the experimental test reactor ITER. ITER will stimulate research in many areas of science. This article serves as an introduction to some of those areas. In particular, we discuss research opportunities in the context of plasma-surface interactions. The fusion plasma, with a typical temperature of 10 keV, has to be brought into contact with a physical wall in order to remove the helium produced and drain the excess energy in the fusion plasma. The fusion plasma is far too hot to be brought into direct contact with a physical wall. It would degrade the wall and the debris from the wall would extinguish the plasma. Therefore, schemes are developed to cool down the plasma locally before it impacts on a physical surface. The resulting plasma-surface interaction in ITER is facing several challenges including surface erosion, material redeposition and tritium retention. In this article we introduce how the plasma-surface interaction relevant for ITER can be studied in small scale experiments. The various requirements for such experiments are introduced and examples of present and future experiments will be given. The emphasis in this article will be on the experimental studies of plasma-surface interactions.

Design, R&D and commissioning of EAST tungsten divertor

NASA Astrophysics Data System (ADS)

Yao, D. M.; Luo, G. N.; Zhou, Z. B.; Cao, L.; Li, Q.; Wang, W. J.; Li, L.; Qin, S. G.; Shi, Y. L.; Liu, G. H.; Li, J. G.

2016-02-01

After commissioning in 2005, the EAST superconducting tokamak had been operated with its water cooled divertors for eight campaigns up to 2012, employing graphite as plasma facing material. With increase in heating power over 20 MW in recent years, the heat flux going to the divertors rises rapidly over 10 MW m-2 for steady state operation. To accommodate the rapid increasing heat load in EAST, the bolting graphite tile divertor must be upgraded. An ITER-like tungsten (W) divertor has been designed and developed; and firstly used for the upper divertor of EAST. The EAST upper W divertor is modular structure with 80 modules in total. Eighty sets of W/Cu plasma-facing components (PFC) with each set consisting of an outer vertical target (OVT), an inner vertical target (IVT) and a DOME, are attached to 80 stainless steel cassette bodies (CB) by pins. The monoblock W/Cu-PFCs have been developed for the strike points of both OVT and IVT, and the flat type W/Cu-PFCs for the DOME and the baffle parts of both OVT and IVT, employing so-called hot isostatic pressing (HIP) technology for tungsten to CuCrZr heat sink bonding, and electron beam welding for CuCrZr to CuCrZr and CuCrZr to other material bonding. Both monoblock and flat type PFC mockups passed high heat flux (HHF) testing by means of electron beam facilities. The 80 divertor modules were installed in EAST in 2014 and results of the first commissioning are presented in this paper.

Examining the temperature behavior of stainless steel surfaces exposed to hydrogen plasmas in the Lithium Tokamak eXperiment (LTX)

NASA Astrophysics Data System (ADS)

Bedoya, Felipe; Allain, Jean Paul; Kaita, Robert; Lucia, Matthew; St-Onge, Denis; Ellis, Robert; Majeski, Richard

2014-10-01

The Materials Analysis Particle Probe (MAPP) is an in-situ diagnostic designed to characterize plasma-facing components (PFCs) in tokamak devices. MAPP is installed in LTX at Princeton Plasma Physics Laboratory. MAPP's capabilities include remotely operated XPS acquisition and temperature control of four samples. The recent addition of a focused ion beam allows XPS depth profiling analysis. Recent published results show an apparent correlation between hydrogen retention and temperature of Li coated stainless steel (SS) PFCs exposed to plasmas like those of LTX. According to XPS data, the retention of hydrogen by the coated surfaces decreases at above 180 °C. In the present study MAPP will be used to study the oxidation of Li coatings as a function of time and temperature of the walls when Li coatings are applied. Experiments in the ion-surface interaction experiment (IIAX) varying the hydrogen fluence on the SS samples will be also performed. Conclusions resulting from this study will be key to explain the PFC temperature-dependent variation of plasma performance observed in LTX. This work was supported by U.S. DOE Contracts DE-AC02-09CH11466, DE-AC52-07NA27344 and DE-SC0010717.

Edge properties with the liquid lithium limiter in FTU—experiment and transport modelling

NASA Astrophysics Data System (ADS)

Pericoli-Ridolfini, V.; Apicella, M. L.; Mazzitelli, G.; Tudisco, O.; Zagórski, R.; FTU Team

2007-07-01

Liquid lithium as a plasma-facing material was tested for the first time on a high field medium size tokamak, FTU. A liquid Li reservoir supplies a mesh of capillaries that is movable from shot to shot in the scrape-off layer (SOL) plasma to act as a secondary limiter. An almost complete lithization of the vacuum vessel walls is obtained in about three discharges. Plasmas cleaner than boronization and titanization, with lower radiation losses and smaller impurity content are produced. The SOL electron temperature increases, ΔTe ~ 10 eV, while density (ne) is less affected. The 2D multifluid code TECXY explains this only if a strong reduction of plasma recycling on the walls and main limiter occurs, consistent with the high Li hydrogen pumping capability. This property also permits a much tighter control of the plasma density. With the Li limiter inserted inside the vessel poloidal asymmetries develop in the SOL that TECXY explains with a local increase of radiation, caused by enhanced evaporation/sputtering of Li. New regimes can be produced in such conditions with a clear increase in |∇ne/ne| and of the peaking factor ne0/

Optimization of Indium Bump Morphology for Improved Flip Chip Devices

NASA Technical Reports Server (NTRS)

Jones, Todd J.; Nikzad, Shouleh; Cunningham, Thomas J.; Blazejewski, Edward; Dickie, Matthew R.; Hoenk, Michael E.; Greer, Harold F.

2011-01-01

Flip-chip hybridization, also known as bump bonding, is a packaging technique for microelectronic devices that directly connects an active element or detector to a substrate readout face-to-face, eliminating the need for wire bonding. In order to make conductive links between the two parts, a solder material is used between the bond pads on each side. Solder bumps, composed of indium metal, are typically deposited by thermal evaporation onto the active regions of the device and substrate. While indium bump technology has been a part of the electronic interconnect process field for many years and has been extensively employed in the infrared imager industry, obtaining a reliable, high-yield process for high-density patterns of bumps can be quite difficult. Under the right conditions, a moderate hydrogen plasma exposure can raise the temperature of the indium bump to the point where it can flow. This flow can result in a desirable shape where indium will efficiently wet the metal contact pad to provide good electrical contact to the underlying readout or imager circuit. However, it is extremely important to carefully control this process as the intensity of the hydrogen plasma treatment dramatically affects the indium bump morphology. To ensure the fine-tuning of this reflow process, it is necessary to have realtime feedback on the status of the bumps. With an appropriately placed viewport in a plasma chamber, one can image a small field (a square of approximately 5 millimeters on each side) of the bumps (10-20 microns in size) during the hydrogen plasma reflow process. By monitoring the shape of the bumps in real time using a video camera mounted to a telescoping 12 magnifying zoom lens and associated optical elements, an engineer can precisely determine when the reflow of the bumps has occurred, and can shut off the plasma before evaporation or de-wetting takes place.

Erosion and deuterium retention of CLF-1 steel exposed to deuterium plasma

NASA Astrophysics Data System (ADS)

Qiao, L.; Wang, P.; Hu, M.; Gao, L.; Jacob, W.; Fu, E. G.; Luo, G. N.

2017-12-01

In recent years reduced activation ferritic martensitic steel has been proposed as the plasma-facing material in remote regions of the first wall. This study reports the erosion and deuterium retention behaviours in CLF-1 steel exposed to deuterium (D) plasma in a linear experimental plasma system as function of incident ion energy and fluence. The incident D ion energy ranges from 30 to 180 eV at a flux of 4 × 1021 D m-2 s-1 up to a fluence of 1025 D m-2. SEM images revealed a clear change of the surface morphology as functions of incident fluence and impinging energy. The mass loss results showed a decrease of the total sputtering yield of CLF-1 steel with increasing incident fluence by up to one order of magnitude. The total sputtering yield of CLF-1 steel after 7.2 × 1024 D m-2 deuterium plasma exposure reduced by a factor of 4 compared with that of pure iron, which can be attributed to the enrichment of W at the surface due to preferential sputtering of iron and chromium. After D plasma exposure, the total deuterium retention in CLF-1 steel samples measured by TDS decreased with increasing incident fluence and energy, and a clear saturation tendency as function of incident fluence or energy was also observed.

Methods of chemically converting first materials to second materials utilizing hybrid-plasma systems

DOEpatents

Kong, Peter C.; Grandy, Jon D.

2002-01-01

In one aspect, the invention encompasses a method of chemically converting a first material to a second material. A first plasma and a second plasma are formed, and the first plasma is in fluid communication with the second plasma. The second plasma comprises activated hydrogen and oxygen, and is formed from a water vapor. A first material is flowed into the first plasma to at least partially ionize at least a portion of the first material. The at least partially ionized first material is flowed into the second plasma to react at least some components of the first material with at least one of the activated hydrogen and activated oxygen. Such converts at least some of the first material to a second material. In another aspect, the invention encompasses a method of forming a synthetic gas by flowing a hydrocarbon-containing material into a hybrid-plasma system. In yet another aspect, the invention encompasses a method of degrading a hydrocarbon-containing material by flowing such material into a hybrid-plasma system. In yet another aspect, the invention encompasses a method of releasing an inorganic component of a complex comprising the inorganic component and an other component, wherein the complex is flowed through a hybrid-plasma system.

Recrystallization kinetics of warm-rolled tungsten in the temperature range 1150-1350 °C

NASA Astrophysics Data System (ADS)

Alfonso, A.; Juul Jensen, D.; Luo, G.-N.; Pantleon, W.

2014-12-01

Pure tungsten is a potential candidate material for the plasma-facing first wall and the divertor of fusion reactors. Both parts have to withstand high temperatures during service. This will alter the microstructure of the material by recovery, recrystallization and grain growth and will cause degradation in material properties as a loss in mechanical strength and embrittlement. The thermal stability of a pure tungsten plate warm-rolled to 67% thickness reduction was investigated by long-term isothermal annealing in the temperature range between 1150 °C and 1350 °C up to 2200 h. Changes in the mechanical properties during annealing are quantified by Vickers hardness measurements. They are described concisely by classical kinetic models for recovery and recrystallization. The observed time spans for recrystallization and the obtained value for the activation energy of the recrystallization process indicate a sufficient thermal stability of the tungsten plate during operation below 1075 °C.

Numerical study of the effects of physical parameters on the dynamic fuel retention in tungsten materials

NASA Astrophysics Data System (ADS)

Sang, Chaofeng; Sun, Jizhong; Bonnin, Xavier; Dai, Shuyu; Hu, Wanpeng; Wang, Dezhen

2014-12-01

Effects of different possible values of physical parameters on the fuel retention in tungsten (W) materials are studied in this work since W is considered as the primary plasma-facing surface material and fuel retention is a critical issue for next-step fusion devices. The upgraded Hydrogen Isotope Inventory Processes Code is used to conduct the study. First, the inventories of hydrogen isotopes (HI) inside W with different possible values of diffusivities and recombination rate coefficients are studied; then the influences of uncertainties in diffusivity, trap concentration, and recombination rate on the effective diffusion are also analyzed. Finally, an illustration of effective diffusion on the permeation and inventory is given. The enhancements of HI permeation flux and inventory in bulk W due to the presence of a carbide WxC layer on the PFS are explained.

Assessment and selection of materials for ITER in-vessel components

NASA Astrophysics Data System (ADS)

Kalinin, G.; Barabash, V.; Cardella, A.; Dietz, J.; Ioki, K.; Matera, R.; Santoro, R. T.; Tivey, R.; ITER Home Teams

2000-12-01

During the international thermonuclear experimental reactor (ITER) engineering design activities (EDA) significant progress has been made in the selection of materials for the in-vessel components of the reactor. This progress is a result of the worldwide collaboration of material scientists and industries which focused their effort on the optimisation of material and component manufacturing and on the investigation of the most critical material properties. Austenitic stainless steels 316L(N)-IG and 316L, nickel-based alloys Inconel 718 and Inconel 625, Ti-6Al-4V alloy and two copper alloys, CuCrZr-IG and CuAl25-IG, have been proposed as reference structural materials, and ferritic steel 430, and austenitic steel 304B7 with the addition of boron have been selected for some specific parts of the ITER in-vessel components. Beryllium, tungsten and carbon fibre composites are considered as plasma facing armour materials. The data base on the properties of all these materials is critically assessed and briefly reviewed in this paper together with the justification of the material selection (e.g., effect of neutron irradiation on the mechanical properties of materials, effect of manufacturing cycle, etc.).

Investigation of plasma-induced erosion of multilayer condenser optics

NASA Astrophysics Data System (ADS)

Anderson, Richard J.; Buchenauer, Dean A.; Williams, K. A.; Clift, W. M.; Klebanoff, L. E.; Edwards, N. V.; Wood, O. R., II; Wurm, S.

2005-05-01

Experiments are presented that investigate the mechanistic cause of multilayer erosion observed from condenser optics exposed to EUV laser-produced plasma (LPP) sources. Using a Xe filament jet source excited with Nd-YAG laser radiation (300 mJ/pulse), measurements were made of material erosion from Au, Mo, Si and C using coated quartz microbalances located 127 mm from the plasma. The observed erosion rates were as follows: Au=99nm/106 shots, Mo= 26nm/106 shots, Si=19nm/106 shots, and C=6nm/106 shots. The relative ratio Au:Mo:Si:C of erosion rates observed experimentally, 16:4:3:1 compares favorably with that predicted from an atomic sputtering model assuming 20 kV Xe ions, 16:6:4:1. The relative agreement indicates that Xe-substrate sputtering is largely responsible for the erosion of Mo/Si multilayers on condenser optics that directly face the plasma. Time-of-flight Faraday cup measurements reveal the emission of high energy Xe ions from the Xe-filament jet plasma. The erosion rate does not depend on the repetition rate of the laser, suggesting a thermal mechanism is not operative. The Xe-filament jet erosion is ~20x that observed from a Xe spray jet. Since the long-lived (millisecond time scale) plasma emanating from these two sources are the same to within ~30%, sputtering from this long-lived plasma can be ruled out as an erosion agent.

Suppressed gross erosion of high-temperature lithium via rapid deuterium implantation

DOE PAGES

Abrams, T.; Jaworski, M. A.; Chen, M.; ...

2015-12-17

Lithium-coated high-Z substrates are planned for use in the NSTX-U divertor and are a candidate plasma facing component (PFC) for reactors, but it remains necessary to characterize the gross Li erosion rate under high plasma fluxes (>10 23 m -2 s -1), typical for the divertor region. In this work, a realistic model for the compositional evolution of a Li/D layer is developed that incorporates first principles molecular dynamics (MD) simulations of D diffusion in liquid Li. Predictions of Li erosion from a mixed Li/D material are also developed that include formation of lithium deuteride (LiD). The erosion rate ofmore » Li from LiD is predicted to be significantly lower than from pure Li. This prediction is tested in the Magnum-PSI linear plasma device at ion fluxes of 10 23-10 24 m -2 s -1 and Li surface temperatures. ≤800 °C. Li/LiD coatings ranging in thickness from 0.2 to 500 μm are studied. The dynamic D/Li concentrations are inferred via diffusion simulations. The pure Li erosion rate remains greater than Langmuir Law evaporation, as expected. For mixed-material Li/LiD surfaces, the erosion rates are reduced, in good agreement with modelling in almost all cases. Lastly, these results imply that the temperature limit for a Li-coated PFC may be significantly higher than previously imagined.« less

Development and Testing of Dispersion-Strengthened Tungsten Alloys via Spark Plasma Sinterin

NASA Astrophysics Data System (ADS)

Lang, Eric; Madden, Nathan; Smith, Charles; Krogstad, Jessica; Allain, Jean Paul

2017-10-01

Tungsten (W) is a common plasma-facing component (PFC) material in the divertor region of tokamak fusion devices due to its high melting point and high sputter threshold. However, W is intrinsically brittle and is further embrittled under neutron irradiation, and the low recrystallization temperature pose complications in fusion environments. More ductile W alloys, such as dispersion-strengthened tungsten are being developed. In this work, W samples are processed via spark plasma sintering (SPS) with TiC, ZrC, and TaC dispersoids alloyed from 0.5 to 10 weight %. SPS is a powder compaction technique that provides high pressure and heating rates via electrical current, allowing for a lower final temperature and hold time for compaction. Initial testing of material properties, smicrostructure, and composition of specimens will be presented. Deuterium and helium irradiations have been performed in IGNIS, a multi-functional, in-situ irradiation and characterization facility at the University of Illinois. High-flux, low-energy exposures at the Magnum-PSI facility at DIFFER exposed samples to a D fluence of 1×1026 cm-2 and He fluence of 1x1025-1x1026 cm-2 at temperatures of 300-1000 C. In-situ chemistry changes via XPS and ex-situ morphology changes via SEM will be studied. Work supported by US DOE Contract DE-SC0014267.

Manufacturing and High Heat Flux Testing of Brazed Flat-Type W/CuCrZr Plasma Facing Components

NASA Astrophysics Data System (ADS)

Lian, Youyun; Liu, Xiang; Feng, Fan; Chen, Lei; Cheng, Zhengkui; Wang, Jin; Chen, Jiming

2016-02-01

Water-cooled flat-type W/CuCrZr plasma facing components with an interlayer of oxygen-free copper (OFC) have been developed by using vacuum brazing route. The OFC layer for the accommodation of thermal stresses was cast onto the surface of W at a temperature range of 1150 °C-1200 °C in a vacuum furnace. The W/OFC cast tiles were vacuum brazed to a CuCrZr heat sink at 940 °C using the silver-free filler material CuMnSiCr. The microstructure, bonding strength, and high heat flux properties of the brazed W/CuCrZr joint samples were investigated. The W/Cu joint exhibits an average tensile strength of 134 MPa, which is about the same strength as pure annealed copper. High heat flux tests were performed in the electron beam facility EMS-60. Experimental results indicated that the brazed W/CuCrZr mock-up experienced screening tests of up to 15 MW/m2 and cyclic tests of 9 MW/m2 for 1000 cycles without visible damage. supported by National Natural Science Foundation of China (No. 11205049) and the National Magnetic Confinement Fusion Science Program of China (No. 2011GB110004)

In situ investigation of helium fuzz growth on tungsten in relation to ion flux, fluence, surface temperature and ion energy using infrared imaging in PSI-2

NASA Astrophysics Data System (ADS)

Möller, S.; Kachko, O.; Rasinski, M.; Kreter, A.; Linsmeier, Ch

2017-12-01

Tungsten is a candidate material for plasma-facing components in nuclear fusion reactors. In operation it will face temperatures >800 K together with an influx of helium ions. Previously, the evolution of special surface nanostructures called fuzz was found under these conditions in a limited window of surface temperature, ion flux and ion energy. Fuzz potentially leads to lower heat load tolerances, enhanced erosion and dust formation, hence should be avoided in a fusion reactor. Here the fuzz growth is reinvestigated in situ during its growth by considering its impact on the surfaces infrared emissivity at 4 μm wavelength with an infrared camera in the linear plasma device PSI-2. A hole in the surface serves as an emissivity reference to calibrate fuzz thickness versus infrared emissivity. Among new data on the above mentioned relations, a lower fuzz growth threshold of 815 ± 24 K is found. Fuzz is seen to grow on rough and polished surfaces and even on the hole’s side walls alike. Literature scalings for thickness, flux and time relations of the fuzz growth rate could not be reproduced, but for the temperature scaling a good agreement to the Arrhenius equation was found.

Conceptual design of divertor and first wall for DEMO-FNS

NASA Astrophysics Data System (ADS)

Sergeev, V. Yu.; Kuteev, B. V.; Bykov, A. S.; Gervash, A. A.; Glazunov, D. A.; Goncharov, P. R.; Dnestrovskij, A. Yu.; Khayrutdinov, R. R.; Klishchenko, A. V.; Lukash, V. E.; Mazul, I. V.; Molchanov, P. A.; Petrov, V. S.; Rozhansky, V. A.; Shpanskiy, Yu. S.; Sivak, A. B.; Skokov, V. G.; Spitsyn, A. V.

2015-11-01

Key issues of design of the divertor and the first wall of DEMO-FNS are presented. A double null closed magnetic configuration was chosen with long external legs and V-shaped corners. The divertor employs a cassette design similar to that of ITER. Water-cooled first wall of the tokamak is made of Be tiles and CuCrZr-stainless steel shells. Lithium injection and circulation technologies are foreseen for protection of plasma facing components. Simulations of thermal loads onto the first wall and divertor plates suggest a possibility to distribute heat loads making them less than 10 MW m-2. Evaluations of sputtering and evaporation of plasma-facing materials suggest that lithium may protect the first wall. To prevent Be erosion at the outer divertor plates either the full detached divertor operation or arrangement of the renewal lithium flow on targets should be implemented. Test bed experiments on the Tsefey-M facility with the first wall mockup coated by Ве tiles and cooled by water are presented. The temperature of the surface of tiles reached 280-300 °С at 5 MW m-2 and 600-650 °С at 10.5 MW m-2. The mockup successfully withstood 1000 cycles with the lower thermal loading and 100 cycles with higher thermal loading.

Plasma-assisted microwave processing of materials

NASA Technical Reports Server (NTRS)

Barmatz, Martin (Inventor); Jackson, Henry (Inventor); Ylin, Tzu-yuan (Inventor)

1998-01-01

A microwave plasma assisted method and system for heating and joining materials. The invention uses a microwave induced plasma to controllably preheat workpiece materials that are poorly microwave absorbing. The plasma preheats the workpiece to a temperature that improves the materials' ability to absorb microwave energy. The plasma is extinguished and microwave energy is able to volumetrically heat the workpiece. Localized heating of good microwave absorbing materials is done by shielding certain parts of the workpiece and igniting the plasma in the areas not shielded. Microwave induced plasma is also used to induce self-propagating high temperature synthesis (SHS) process for the joining of materials. Preferably, a microwave induced plasma preheats the material and then microwave energy ignites the center of the material, thereby causing a high temperature spherical wave front from the center outward.

Arcing and its role in PFC erosion and dust production in DIII-D

NASA Astrophysics Data System (ADS)

Rudakov, D. L.; Chrobak, C. P.; Doerner, R. P.; Krasheninnikov, S. I.; Moyer, R. A.; Umstadter, K. R.; Wampler, W. R.; Wong, C. P. C.

2013-07-01

Two types of arc tracks are observed on the plasma-facing components (PFCs) in DIII-D. "Unmagnetized" random walk tracks are produced during glow discharges; they are rare and have no importance for PFC erosion but may degrade diagnostic mirrors. "Magnetized" scratch-like type II tracks are produced by unipolar arcs during plasma operations; they are formed by "retrograde BxJ" motion of the cathode spot and are roughly perpendicular to the local magnetic field. Type II arcs cause measurable erosion of graphite, but based on the evidence available they are relatively small contributors to the total erosion of carbon in DIII-D compared to other mechanisms such as physical and chemical sputtering and ablation from leading edges. Erosion by arcing of tungsten films deposited on graphite samples was observed in Divertor Material Evaluation System (DiMES) experiments. New DiMES experiments aimed at time-resolved arc measurements are proposed.

Friction surfaced Stellite6 coatings

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rao, K. Prasad; Damodaram, R.; Rafi, H. Khalid, E-mail: khalidrafi@gmail.com

2012-08-15

Solid state Stellite6 coatings were deposited on steel substrate by friction surfacing and compared with Stellite6 cast rod and coatings deposited by gas tungsten arc and plasma transferred arc welding processes. Friction surfaced coatings exhibited finer and uniformly distributed carbides and were characterized by the absence of solidification structure and compositional homogeneity compared to cast rod, gas tungsten arc and plasma transferred coatings. Friction surfaced coating showed relatively higher hardness. X-ray diffraction of samples showed only face centered cubic Co peaks while cold worked coating showed hexagonally close packed Co also. - Highlights: Black-Right-Pointing-Pointer Stellite6 used as coating material formore » friction surfacing. Black-Right-Pointing-Pointer Friction surfaced (FS) coatings compared with casting, GTA and PTA processes. Black-Right-Pointing-Pointer Finer and uniformly distributed carbides in friction surfaced coatings. Black-Right-Pointing-Pointer Absence of melting results compositional homogeneity in FS Stellite6 coatings.« less

The Influence of Microstructure on Deuterium Retention in Polycrystalline Tungsten

DOE PAGES

Garrison, Lauren M.; Meyer, Fred W.; Bannister, Mark E.

2017-09-18

The retention of hydrogen isotopes in the plasma-facing materials of a fusion reactor is dependent on the density of trapping sites in the material. One factor that can influence the trapping defects is the surface state of the material before exposure. Mechanically polished, electropolished, and recrystallized tungsten samples were compared by exposing them to 350 eV D + beams with peak fluences of ~1 × 10 24 D +/m 2 at 500 and 740 K at the Multicharged Ion Research Facility (MIRF). At the exposure temperature of 740 K, no significant retention was detected. For material exposed at 500 K,more » significant differences in retention were observed, and the order of increasing retention was recrystallized, electropolished, and mechanically polished. Lastly, the other variable besides surface treatment was the time delay between ion exposure and thermal desorption spectroscopy which also may have impacted the retention measurements if there was out-gassing of the D while samples were in storage before thermal desorption spectroscopy (TDS).« less

An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility

NASA Astrophysics Data System (ADS)

Calderoni, P.; Sharpe, J.; Shimada, M.; Denny, B.; Pawelko, B.; Schuetz, S.; Longhurst, G.; Hatano, Y.; Hara, M.; Oya, Y.; Otsuka, T.; Katayama, K.; Konishi, S.; Noborio, K.; Yamamoto, Y.

2011-10-01

The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.

The Influence of Microstructure on Deuterium Retention in Polycrystalline Tungsten

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garrison, Lauren M.; Meyer, Fred W.; Bannister, Mark E.

The retention of hydrogen isotopes in the plasma-facing materials of a fusion reactor is dependent on the density of trapping sites in the material. One factor that can influence the trapping defects is the surface state of the material before exposure. Mechanically polished, electropolished, and recrystallized tungsten samples were compared by exposing them to 350 eV D + beams with peak fluences of ~1 × 10 24 D +/m 2 at 500 and 740 K at the Multicharged Ion Research Facility (MIRF). At the exposure temperature of 740 K, no significant retention was detected. For material exposed at 500 K,more » significant differences in retention were observed, and the order of increasing retention was recrystallized, electropolished, and mechanically polished. Lastly, the other variable besides surface treatment was the time delay between ion exposure and thermal desorption spectroscopy which also may have impacted the retention measurements if there was out-gassing of the D while samples were in storage before thermal desorption spectroscopy (TDS).« less

Scaling mechanisms of vapour/plasma shielding from laser-produced plasmas to magnetic fusion regimes

NASA Astrophysics Data System (ADS)

Sizyuk, Tatyana; Hassanein, Ahmed

2014-02-01

The plasma shielding effect is a well-known mechanism in laser-produced plasmas (LPPs) reducing laser photon transmission to the target and, as a result, significantly reducing target heating and erosion. The shielding effect is less pronounced at low laser intensities, when low evaporation rate together with vapour/plasma expansion processes prevent establishment of a dense plasma layer above the surface. Plasma shielding also loses its effectiveness at high laser intensities when the formed hot dense plasma plume causes extensive target erosion due to radiation fluxes back to the surface. The magnitude of emitted radiation fluxes from such a plasma is similar to or slightly higher than the laser photon flux in the low shielding regime. Thus, shielding efficiency in LPPs has a peak that depends on the laser beam parameters and the target material. A similar tendency is also expected in other plasma-operating devices such as tokamaks of magnetic fusion energy (MFE) reactors during transient plasma operation and disruptions on chamber walls when deposition of the high-energy transient plasma can cause severe erosion and damage to the plasma-facing and nearby components. A detailed analysis of these abnormal events and their consequences in future power reactors is limited in current tokamak reactors. Predictions for high-power future tokamaks are possible only through comprehensive, time-consuming and rigorous modelling. We developed scaling mechanisms, based on modelling of LPP devices with their typical temporal and spatial scales, to simulate tokamak abnormal operating regimes to study wall erosion, plasma shielding and radiation under MFE reactor conditions. We found an analogy in regimes and results of carbon and tungsten erosion of the divertor surface in ITER-like reactors with erosion due to laser irradiation. Such an approach will allow utilizing validated modelling combined with well-designed and well-diagnosed LPP experimental studies for predicting consequences of plasma instabilities in complex fusion environment, which are of serious concern for successful energy production.

The thermodynamic and kinetic interactions of He interstitial clusters with bubbles in W

DOE PAGES

Perez, Danny; Sandoval, Luis; Uberuaga, Blas P.; ...

2016-05-26

Due to its enviable properties, tungsten is a leading candidate plasma facing material in nuclear fusion reactors. But, like many other metals, tungsten is known to be affected by the high doses of helium atoms incoming from the plasma. Indeed, the implanted interstitial helium atoms cluster together and, upon reaching a critical cluster size, convert into substitutional nanoscale He bubbles. These bubbles then grow by absorbing further interstitial clusters from the matrix. This process can lead to deleterious changes in microstructure, degradation of mechanical properties, and contamination of the plasma. In order to better understand the growth process, we usemore » traditional and accelerated molecular dynamics simulations to investigate the interactions between interstitial He clusters and pre-existing bubbles. These interactions are characterized in terms of thermodynamics and kinetics. We also show that the proximity of the bubble leads to an enhancement of the trap mutation rate and, consequently, to the nucleation of satellite bubbles in the neighborhood of existing ones. Finally, we uncover a number of mechanisms that can lead to the subsequent annihilation of such satellite nanobubbles.« less

The thermodynamic and kinetic interactions of He interstitial clusters with bubbles in W

NASA Astrophysics Data System (ADS)

Perez, Danny; Sandoval, Luis; Uberuaga, Blas P.; Voter, Arthur F.

2016-05-01

Due to its enviable properties, tungsten is a leading candidate plasma facing material in nuclear fusion reactors. However, like many other metals, tungsten is known to be affected by the high doses of helium atoms incoming from the plasma. Indeed, the implanted interstitial helium atoms cluster together and, upon reaching a critical cluster size, convert into substitutional nanoscale He bubbles. These bubbles then grow by absorbing further interstitial clusters from the matrix. This process can lead to deleterious changes in microstructure, degradation of mechanical properties, and contamination of the plasma. In order to better understand the growth process, we use traditional and accelerated molecular dynamics simulations to investigate the interactions between interstitial He clusters and pre-existing bubbles. These interactions are characterized in terms of thermodynamics and kinetics. We show that the proximity of the bubble leads to an enhancement of the trap mutation rate and, consequently, to the nucleation of satellite bubbles in the neighborhood of existing ones. We also uncover a number of mechanisms that can lead to the subsequent annihilation of such satellite nanobubbles.

Integration of uncooled scraper elements and its diagnostics into Wendelstein 7-X

DOE PAGES

Fellinger, Joris; Loesser, Doug; Neilson, Hutch; ...

2017-08-08

The modular stellarator Wendelstein 7-X in Greifswald (Germany) successfully started operation in 2015 with short pulse limiter plasmas. In 2017, the next operation phase (OP) OP1.2 will start once 10 uncooled test divertor units (TDU) with graphite armor will be installed. The TDUs allow for plasma pulses of 10 s with 8 MW heating. OP2, allowing for steady state operation, is planned for 2020 after the TDUs will be replaced by 10 water cooled CFC armored divertors. Due to the development of plasma currents like bootstrap currents in long pulse plasmas in OP2, the plasma could hit the edge ofmore » the divertor targets which has a reduced cooling capacity compared to the central part of the target tiles. To prevent overloading of these edges, a so-called scraper element can be positioned in front of the divertor, intersecting those strike lines that would otherwise hit the divertor edges. As a result, these edges are protected but as a drawback the pumping efficiency of neutrals is also reduced. As a test an uncooled scraper element with graphite tiles will be placed in two out of ten half modules in OP1.2. A decision to install ten water cooled scraper elements for OP2 is pending on the results of this test in OP1.2. To monitor the impact of the scraper element on the plasma, Langmuir probes are integrated in the plasma facing surface, and a neutral gas manometer measures the neutral density directly behind the plasma facing surface. Moreover, IR and VIS cameras observe the plasma facing surface and thermocouples monitor the temperatures of the graphite tiles and underlying support structure. This paper describes the integration of the scraper element and its diagnostics in Wendelstein 7-X.« less

Plasma facing components: a conceptual design strategy for the first wall in FAST tokamak

NASA Astrophysics Data System (ADS)

Labate, C.; Di Gironimo, G.; Renno, F.

2015-09-01

Satellite tokamaks are conceived with the main purpose of developing new or alternative ITER- and DEMO-relevant technologies, able to contribute in resolving the pending issues about plasma operation. In particular, a high criticality needs to be associated to the design of plasma facing components, i.e. first wall (FW) and divertor, due to physical, topological and thermo-structural reasons. In such a context, the design of the FW in FAST fusion plant, whose operational range is close to ITER’s one, takes place. According to the mission of experimental satellites, the FW design strategy, which is presented in this paper relies on a series of innovative design choices and proposals with a particular attention to the typical key points of plasma facing components design. Such an approach, taking into account a series of involved physical constraints and functional requirements to be fulfilled, marks a clear borderline with the FW solution adopted in ITER, in terms of basic ideas, manufacturing aspects, remote maintenance procedure, manifolds management, cooling cycle and support system configuration.

Design of a Microwave Assisted Discharge Inductive Plasma Accelerator

NASA Technical Reports Server (NTRS)

Hallock, Ashley K.; Polzin, Kurt A.

2010-01-01

The design and construction of a thruster that employs electrodeless plasma preionization and pulsed inductive acceleration is described. Preionization is achieved through an electron cyclotron resonance discharge that produces a weakly-ionized plasma at the face of a conical theta pinch-shaped inductive coil. The presence of the preionized plasma allows for current sheet formation at lower discharge voltages than those employed in other pulsed inductive accelerators that do not employ preionization. The location of the electron cyclotron resonance discharge is controlled through the design of the applied magnetic field in the thruster. Finite element analysis shows that there is an arrangement of permanent magnets that yields a small volume of resonant magnetic field at the coil face. Preionization in the resonant zone leads to current sheet formation at the coil face, which minimizes the initial inductance of the pulse circuit and maximizes the potential electrical efficiency of the accelerator. A magnet assembly was constructed around an inductive coil to provide structural support to the selected arrangement of neodymium magnets. Measured values of the resulting magnetic field compare favorably with the finite element model.

High Speed Photographic Analysis Of Railgun Plasmas

NASA Astrophysics Data System (ADS)

Macintyre, I. B.

1985-02-01

Various experiments are underway at the Materials Research Laboratories, Australian Department of Defence, to develop a theory for the behaviour and propulsion action of plasmas in rail guns. Optical recording and imaging devices, with their low vulnerability to the effects of magnetic and electric fields present in the vicinity of electromagnetic launchers, have proven useful as diagnostic tools. This paper describes photoinstrumentation systems developed to provide visual qualitative assessment of the behaviour of plasma travelling along the bore of railgun launchers. In addition, a quantitative system is incorporated providing continuous data (on a microsecond time scale) of (a) Length of plasma during flight along the launcher bore. (b) Velocity of plasma. (c) Distribution of plasma with respect to time after creation. (d) Plasma intensity profile as it travels along the launcher bore. The evolution of the techniques used is discussed. Two systems were employed. The first utilized a modified high speed streak camera to record the light emitted from the plasma, through specially prepared fibre optic cables. The fibre faces external to the bore were then imaged onto moving film. The technique involved the insertion of fibres through the launcher body to enable the plasma to be viewed at discrete positions as it travelled along the launcher bore. Camera configuration, fibre optic preparation and experimental results are outlined. The second system utilized high speed streak and framing photography in conjunction with accurate sensitometric control procedures on the recording film. The two cameras recorded the plasma travelling along the bore of a specially designed transparent launcher. The streak camera, fitted with a precise slit size, recorded a streak image of the upper brightness range of the plasma as it travelled along the launcher's bore. The framing camera recorded an overall view of the launcher and the plasma path, to the maximum possible, governed by the film's ability to reproduce the plasma's brightness range. The instrumentation configuration, calibration, and film measurement using microdensitometer scanning techniques to evaluate inbore plasma behaviour, are also presented.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kaita, Robert; Boyle, Dennis; Gray, Timothy

Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The Lithium Tokamak Experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the Current Drive Experiment-Upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in Ohmically-heated tokamak plasmas was achieved with a toroidal liquid lithium limiter. The LTX extends this liquid lithium PFC by using a conducting conformal shell that almost completely surrounds the plasma. By heating themore » shell, a lithium coating on the plasma-facing side can be kept liquefied. A consequence of the low-recycling conditions from liquid lithium walls is the need for efficient plasma fueling. For this purpose, a molecular cluster injector is being developed. Future plans include the installation of a neutral beam for core plasma fueling, and also ion temperature measurements using charge-exchange recombination spectroscopy. Low edge recycling is also predicted to reduce temperature gradients that drive drift wave turbulence. Gyrokinetic simulations are in progress to calculate fluctuation levels and transport for LTX plasmas, and new fluctuation diagnostics are under development to test these predictions. __________________________________________________« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ruzic, David

The Thermoelectric-Driven Liquid-Metal Plasma-Facing Structures (TELS) project was able to establish the experimental conditions necessary for flowing liquid metal surfaces in order to be utilized as surfaces facing fusion relevant energetic plasma flux. The work has also addressed additional developments along with progressing along the timeline detailed in the proposal. A no-cost extension was requested to conduct other relevant experiment- specifically regarding the characterization droplet ejection during energetic plasma flux impact. A specially designed trench module, which could accommodate trenches with different aspect ratios was fabricated and installed in the TELS setup and plasma gun experiments were performed. Droplet ejectionmore » was characterized using high speed image acquisition and also surface mounted probes were used to characterize the plasma. The Gantt chart below had been provided with the original proposal, indicating the tasks to be performed in the third year of funding. These tasks are listed above in the progress report outline, and their progress status is detailed below.« less

Liquid surfaces for fusion plasma facing components—A critical review. Part I: Physics and PSI

DOE PAGES

Nygren, R. E.; Tabares, F. L.

2016-12-01

This review of the potential of robust plasma facing components (PFCs) with liquid surfaces for applications in future D/T fusion device summarizes the critical issues for liquid surfaces and research being done worldwide in confinement facilities, and supporting R&D in plasma surface interactions. In the paper are a set of questions and related criteria by which we will judge the progress and readiness of liquid surface PFCs. Part-II (separate paper) will cover R&D on the technology-oriented aspects of liquid surfaces including the liquid surfaces as integrated first walls in tritium breeding blankets, tritium retention and recovery, and safety.

PLASMA DEVICE

DOEpatents

Baker, W.R.; Brathenahl, A.; Furth, H.P.

1962-04-10

A device for producing a confined high temperature plasma is described. In the device the concave inner surface of an outer annular electrode is disposed concentrically about and facing the convex outer face of an inner annular electrode across which electrodes a high potential is applied to produce an electric field there between. Means is provided to create a magnetic field perpendicular to the electric field and a gas is supplied at reduced pressure in the area therebetween. Upon application of the high potential, the gas between the electrodes is ionized, heated, and under the influence of the electric and magnetic fields there is produced a rotating annular plasma disk. The ionized plasma has high dielectric constant properties. The device is useful as a fast discharge rate capacitor, in controlled thermonuclear research, and other high temperature gas applications. (AEC)

Innovative design and material solutions of thermal contact layers for high heat flux applications in fusion devices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Federici, G.; Matera, R.; Chiocchio, S.

1994-11-01

One difficulty associated with the design and development of sacrificial plasma facing components that have to handle the high heat and particle fluxes in ITER is achieving the necessary contact conductance between the plasma protection material and the high-conductivity substrate in contact with the coolant. This paper presents a novel bond idea which is proposed as one of the options for the sacrificial energy dump targets located at the bottom of the divertor legs. The bonded joint in this design concept provides thermal and electrical contact between the armour and the cooled sub-structure while promoting remote, in-situ maintenance repair andmore » an easy replaceability of the armour part without disturbing the cooling pipes or rewelding neutron irradiated materials. To provide reliable and demountable adhesion, the bond consists of a metal alloy, treated in the semi-solid phase so that it leads to a fine dispersion of a globular solid phase into a liquid matrix (rheocast process). This thermal bond layer would normally operate in the solid state but could be brought reversibly to the semi-solid state during the armour replacement simply by heating it slightly above its solidus temperature. Material and design options are discussed in this paper. Possible methods of installation and removal are described, and lifetime considerations are addressed. In order to validate this concept within the ITER time-frame, a R&D programme must be rapidly implemented.« less

Collection strategy, inner morphology, and size distribution of dust particles in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Balden, M.; Endstrasser, N.; Humrickhouse, P. W.; Rohde, V.; Rasinski, M.; von Toussaint, U.; Elgeti, S.; Neu, R.; the ASDEX Upgrade Team

2014-07-01

The dust collection and analysis strategy in ASDEX Upgrade (AUG) is described. During five consecutive operation campaigns (2007-2011), Si collectors were installed, which were supported by filtered vacuum sampling and collection with adhesive tapes in 2009. The outer and inner morphology (e.g. shape) and elemental composition of the collected particles were analysed by scanning electron microscopy. The majority of the ˜50 000 analysed particles on the Si collectors of campaign 2009 contain tungsten—the plasma-facing material in AUG—and show basically two different types of outer appearance: spheroids and irregularly shaped particles. By far most of the W-dominated spheroids consist of a solid W core, i.e. solidified W droplets. A part of these particles is coated with a low-Z material; a process that seems to happen presumably in the far scrape-off layer plasma. In addition, some conglomerates of B, C and W appear as spherical particles after their contact with plasma. By far most of the particles classified as B-, C- and W-dominated irregularly shaped particles consist of the same conglomerate with varying fraction of embedded W in the B-C matrix and some porosity, which can exceed 50%. The fragile structures of many conglomerates confirm the absence of intensive plasma contact. Both the ablation and mobilization of conglomerate material and the production of W droplets are proposed to be triggered by arcing. The size distribution of each dust particle class is best described by a log-normal distribution allowing an extrapolation of the dust volume and surface area. The maximum in this distribution is observed above the resolution limit of 0.28 µm only for the W-dominated spheroids, at around 1 µm. The amount of W-containing dust is extrapolated to be less than 300 mg on the horizontal areas of AUG.

New steady-state quiescent high-confinement plasma in an experimental advanced superconducting tokamak.

PubMed

Hu, J S; Sun, Z; Guo, H Y; Li, J G; Wan, B N; Wang, H Q; Ding, S Y; Xu, G S; Liang, Y F; Mansfield, D K; Maingi, R; Zou, X L; Wang, L; Ren, J; Zuo, G Z; Zhang, L; Duan, Y M; Shi, T H; Hu, L Q

2015-02-06

A critical challenge facing the basic long-pulse high-confinement operation scenario (H mode) for ITER is to control a magnetohydrodynamic (MHD) instability, known as the edge localized mode (ELM), which leads to cyclical high peak heat and particle fluxes at the plasma facing components. A breakthrough is made in the Experimental Advanced Superconducting Tokamak in achieving a new steady-state H mode without the presence of ELMs for a duration exceeding hundreds of energy confinement times, by using a novel technique of continuous real-time injection of a lithium (Li) aerosol into the edge plasma. The steady-state ELM-free H mode is accompanied by a strong edge coherent MHD mode (ECM) at a frequency of 35-40 kHz with a poloidal wavelength of 10.2 cm in the ion diamagnetic drift direction, providing continuous heat and particle exhaust, thus preventing the transient heat deposition on plasma facing components and impurity accumulation in the confined plasma. It is truly remarkable that Li injection appears to promote the growth of the ECM, owing to the increase in Li concentration and hence collisionality at the edge, as predicted by GYRO simulations. This new steady-state ELM-free H-mode regime, enabled by real-time Li injection, may open a new avenue for next-step fusion development.

Preparation of erosion and deposition investigations on plasma facing components in Wendelstein 7-X

NASA Astrophysics Data System (ADS)

Dhard, C. P.; Balden, M.; Braeuer, T.; Brezinsek, S.; Coenen, J. W.; Dudek, A.; Ehrke, G.; Hathiramani, D.; Klose, S.; König, R.; Laux, M.; Linsmeier, Ch; Manhard, A.; Masuzaki, S.; Mayer, M.; Motojima, G.; Naujoks, D.; Neu, R.; Neubauer, O.; Rack, M.; Ruset, C.; Schwarz-Selinger, T.; Pedersen, T. Sunn; Tokitani, M.; Unterberg, B.; Yajima, M.; W7-X Team1, The

2017-12-01

In the Wendelstein 7-X stellarator with its twisted magnetic geometry the investigation of plasma wall interaction processes in 3D plasma configurations is an important research subject. For the upcoming operation phase i.e. OP1.2, three different types of material probes have been installed within the plasma vessel for the erosion/deposition investigations in selected areas with largely different expected heat load levels, namely, ≤10 MW m-2 at the test divertor units (TDU), ≤500 kW m-2 at the baffles, heat shields and toroidal closures and ≤100 kW m-2 at the stainless steel wall panels. These include 18 exchangeable target elements at TDU, about 30 000 screw heads at graphite tiles and 44 wafer probes on wall panels, coated with marker layers. The layer thicknesses, surface morphologies and the impurity contents were pre-characterized by different techniques and subjected to various qualification tests. The positions of these probes were fixed based on the strike line locations on the divertor predicted by field line diffusion and EMC3/EIRENE modeling calculations for the OP1.2 plasma configurations and availability of locations on panels in direct view of the plasma. After the first half of the operation phase i.e. OP1.2a the probes will be removed to determine the erosion/deposition pattern by post-mortem analysis and replaced by a new set for the second half of the operation phase, OP1.2b.

Effects of the plasma-facing materials on the negative ion H ‑ density in an ECR (2.45 GHz) plasma

NASA Astrophysics Data System (ADS)

Bentounes, J.; Béchu, S.; Biggins, F.; Michau, A.; Gavilan, L.; Menu, J.; Bonny, L.; Fombaron, D.; Bès, A.; Lebedev, Yu A.; Shakhatov, V. A.; Svarnas, P.; Hassaine, T.; Lemaire, J. L.; Lacoste, A.

2018-05-01

Within the framework of fundamental research, the present work focuses on the role of surface material in the production of H ‑ negative ion, with a potential application of designing cesium-free H ‑ negative ion sources oriented to fusion application. It is widely accepted that the main reaction leading to H ‑ production, in the plasma volume, is the dissociative attachment of low-energy electrons (T e ≤ 1 eV) on highly ro-vibrationally excited hydrogen molecules. In parallel with other mechanisms, the density of these excited molecules may be enhanced by means of the recombinative desorption, i.e. the interaction between surface absorbed atoms with other atoms (surface adsorbed or not) through the path {H}{{ads}}+{H}{{gas}/{{ads}}}\\to {H}2{(v,J)}{{gas}}+{{Δ }}E. Accordingly, a systematic study on the role played by the surface in this reaction, with respect to the production of H ‑ ion in the plasma volume, is here performed. Thus, tantalum and tungsten (already known as H ‑ enhancers) and quartz (inert surface) materials are employed as inner surfaces of a test bench chamber. The plasma inside the chamber is produced by electron cyclotron resonance (ECR) driving and it is characterized with conventional electrostatic probes, laser photodetachment, and emission and absorption spectroscopy. Two different positions (close to and away from the ECR driving zone) are investigated under various conditions of pressure and power. The experimental results are supported by numerical data generated by a 1D model. The latter couples continuity and electron energy balance equations in the presence of magnetic field, and incorporates vibrational kinetics, H2 molecular reactions, H electronically excited states and ground-state species kinetics. In the light of this study, recombinative desorption has been evidenced as the most probable mechanism, among others, responsible for an enhancement by a factor of about 3.4, at 1.6 Pa and 175 W of microwave power, in the case of tantalum.

A successful experience of the Iranian blood transfusion organization in improving accessibility and affordability of plasma derived medicine.

PubMed

Chegini, Azita; Torab, Seyed Ardeshir; Pourfatollah, Ali Akbar

2017-02-01

Plasma is the liquid part of blood. It is estimated 21.6 million liters of plasma collect from Whole blood annually. From these plasma, 4.2 million liters transfuse, 8.1 million liters fractionate, 9.3 million liters waste. Nowadays, blood products and PDM (plasma derived medicine) consider as essential medicine in modern health care and transfusion medicine. Iranian blood transfusion organization as a non-profit organization was established in 1974 in order to centralize all blood transfusion activities from donor recruitment to distribution of blood components to hospitals. Iran is the only country in EMR region with the rate of 20-29.9 blood donations per 1000 population and reached 100% voluntary non-remunerated blood donation in 2007. RBCs and platelets demand are much more than FFPs so the IBTO was faced the surplus plasma that could cause surplus plasma wastage. Simultaneously, hospitals need more plasma derived medicine especially albumin, IVIG, factor VIII, factor IX. IBTO was faced the challenges such as Fractionators selection, Plasma volume shipment, Contract duration, Product profile, Multiple External audits, Cold chain maintenance, Transporting plasma across international borders, NAT test. To overcome plasma wastage and storage of PDM. IBTO involved toll manufacturing in 2005 and not only prevents plasma wastage but also save MOH (ministry of health) budget. Copyright © 2016. Published by Elsevier Ltd.

FUSION ENERGY SCIENCES WORKSHOP ON PLASMA MATERIALS INTERACTIONS: Report on Science Challenges and Research Opportunities in Plasma Materials Interactions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maingi, Rajesh; Zinkle, Steven J.; Foster, Mark S.

2015-05-01

The realization of controlled thermonuclear fusion as an energy source would transform society, providing a nearly limitless energy source with renewable fuel. Under the auspices of the U.S. Department of Energy, the Fusion Energy Sciences (FES) program management recently launched a series of technical workshops to “seek community engagement and input for future program planning activities” in the targeted areas of (1) Integrated Simulation for Magnetic Fusion Energy Sciences, (2) Control of Transients, (3) Plasma Science Frontiers, and (4) Plasma-Materials Interactions aka Plasma-Materials Interface (PMI). Over the past decade, a number of strategic planning activities1-6 have highlighted PMI and plasmamore » facing components as a major knowledge gap, which should be a priority for fusion research towards ITER and future demonstration fusion energy systems. There is a strong international consensus that new PMI solutions are required in order for fusion to advance beyond ITER. The goal of the 2015 PMI community workshop was to review recent innovations and improvements in understanding the challenging PMI issues, identify high-priority scientific challenges in PMI, and to discuss potential options to address those challenges. The community response to the PMI research assessment was enthusiastic, with over 80 participants involved in the open workshop held at Princeton Plasma Physics Laboratory on May 4-7, 2015. The workshop provided a useful forum for the scientific community to review progress in scientific understanding achieved during the past decade, and to openly discuss high-priority unresolved research questions. One of the key outcomes of the workshop was a focused set of community-initiated Priority Research Directions (PRDs) for PMI. Five PRDs were identified, labeled A-E, which represent community consensus on the most urgent near-term PMI scientific issues. For each PRD, an assessment was made of the scientific challenges, as well as a set of actions to address those challenges. No prioritization was attempted amongst these five PRDs. We note that ITER, an international collaborative project to substantially extend fusion science and technology, is implicitly a driver and beneficiary of the research described in these PRDs; specific ITER issues are discussed in the background and PRD chapters. For succinctness, we describe these PRDs directly below; a brief introduction to magnetic fusion and the workshop process/timeline is given in Chapter I, and panelists are listed in the Appendix.« less

Material Transport in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Rohde, V.; Dux, R.; Mayer, M.; Neu, R.; PA~ 1/4 tterich, T.; Schneider, W.; ASDEX Upgrade-Team,

Today carbon is the most common first wall material in fusion experiments, whereas the first wall of the next step device will consist of a mixture of elements. Especially tungsten has been shown to be an alternative to low-Z materials. However, even with 40% of tungsten coated plasma facing components, carbon is still the dominant impurity at ASDEX Upgrade. A consistent picture of the carbon migration in ASDEX Upgrade has been achieved. Primary carbon sources are the protection limiters at the low field side of the main chamber. Eroded carbon is distributed all over the main chamber. So, the initially tungsten coated central column acts as the main carbon source during discharges, even though a considerable amount of tungsten surfaces persists. Carbon coverage of the central column can significantly change on a shot to shot basis. The divertor target plates act as a strong carbon sink. Deposits are found at the inner and outer divertor, which may be re-eroded forming precursors for layer production at remote areas. In ASDEX Upgrade, deposits on the subdivertor structure are formed by hydro-carbons with a high effective sticking coefficient. A parasitic plasma at these locations may enhance the surface loss probability by surface activation. At more remote areas, such as the pump ducts, a very small deposition is found. Non sticking hydro-carbons are effectively pumped by the cryopump and turbo molecular pumps.

Evaluation of Cooling Conditions for a High Heat Flux Testing Facility Based on Plasma-Arc Lamps

DOE PAGES

Charry, Carlos H.; Abdel-khalik, Said I.; Yoda, Minami; ...

2015-07-31

The new Irradiated Material Target Station (IMTS) facility for fusion materials at Oak Ridge National Laboratory (ORNL) uses an infrared plasma-arc lamp (PAL) to deliver incident heat fluxes as high as 27 MW/m 2. The facility is being used to test irradiated plasma-facing component materials as part of the joint US-Japan PHENIX program. The irradiated samples are to be mounted on molybdenum sample holders attached to a water-cooled copper rod. Depending on the size and geometry of samples, several sample holders and copper rod configurations have been fabricated and tested. As a part of the effort to design sample holdersmore » compatible with the high heat flux (HHF) testing to be conducted at the IMTS facility, numerical simulations have been performed for two different water-cooled sample holder designs using the ANSYS FLUENT 14.0 commercial computational fluid dynamics (CFD) software package. The primary objective of this work is to evaluate the cooling capability of different sample holder designs, i.e. to estimate their maximum allowable incident heat flux values. 2D axisymmetric numerical simulations are performed using the realizable k-ε turbulence model and the RPI nucleate boiling model within ANSYS FLUENT 14.0. The results of the numerical model were compared against the experimental data for two sample holder designs tested in the IMTS facility. The model has been used to parametrically evaluate the effect of various operational parameters on the predicted temperature distributions. The results were used to identify the limiting parameter for safe operation of the two sample holders and the associated peak heat flux limits. The results of this investigation will help guide the development of new sample holder designs.« less

Developing structural, high-heat flux and plasma facing materials for a near-term DEMO fusion power plant: The EU assessment

NASA Astrophysics Data System (ADS)

Stork, D.; Agostini, P.; Boutard, J. L.; Buckthorpe, D.; Diegele, E.; Dudarev, S. L.; English, C.; Federici, G.; Gilbert, M. R.; Gonzalez, S.; Ibarra, A.; Linsmeier, Ch.; Li Puma, A.; Marbach, G.; Morris, P. F.; Packer, L. W.; Raj, B.; Rieth, M.; Tran, M. Q.; Ward, D. J.; Zinkle, S. J.

2014-12-01

The findings of the EU 'Materials Assessment Group' (MAG), within the 2012 EU Fusion Roadmap exercise, are discussed. MAG analysed the technological readiness of structural, plasma facing and high heat flux materials for a DEMO concept to be constructed in the early 2030s, proposing a coherent strategy for R&D up to a DEMO construction decision. A DEMO phase I with a 'Starter Blanket' and 'Starter Divertor' is foreseen: the blanket being capable of withstanding ⩾2 MW yr m-2 fusion neutron fluence (∼20 dpa in the front-wall steel). A second phase ensues for DEMO with ⩾5 MW yr m-2 first wall neutron fluence. Technical consequences for the materials required and the development, testing and modelling programmes, are analysed using: a systems engineering approach, considering reactor operational cycles, efficient maintenance and inspection requirements, and interaction with functional materials/coolants; and a project-based risk analysis, with R&D to mitigate risks from material shortcomings including development of specific risk mitigation materials. The DEMO balance of plant constrains the blanket and divertor coolants to remain unchanged between the two phases. The blanket coolant choices (He gas or pressurised water) put technical constraints on the blanket steels, either to have high strength at higher temperatures than current baseline variants (above 650 °C for high thermodynamic efficiency from He-gas coolant), or superior radiation-embrittlement properties at lower temperatures (∼290-320 °C), for construction of water-cooled blankets. Risk mitigation proposed would develop these options in parallel, and computational and modelling techniques to shorten the cycle-time of new steel development will be important to achieve tight R&D timescales. The superior power handling of a water-cooled divertor target suggests a substructure temperature operating window (∼200-350 °C) that could be realised, as a baseline-concept, using tungsten on a copper-alloy substructure. The difficulty of establishing design codes for brittle tungsten puts great urgency on the development of a range of advanced ductile or strengthened tungsten and copper compounds. Lessons learned from Fission reactor material development have been included, especially in safety and licensing, fabrication/joining techniques and designing for in-vessel inspection. The technical basis of using the ITER licensing experience to refine the issues in nuclear testing of materials is discussed. Testing with 14 MeV neutrons is essential to Fusion Materials development, and the Roadmap requires acquisition of ⩾30 dpa (steels) 14 MeV test data by 2026. The value and limits of pre-screening testing with fission neutrons on isotopically- or chemically-doped steels and with ion-beams are evaluated to help determine the minimum14 MeV testing programme requirements.

Vapor shielding models and the energy absorbed by divertor targets during transient events

DOE Office of Scientific and Technical Information (OSTI.GOV)

Skovorodin, D. I., E-mail: dskovorodin@gmail.com; Arakcheev, A. S.; Pshenov, A. A.

2016-02-15

The erosion of divertor targets caused by high heat fluxes during transients is a serious threat to ITER operation, as it is going to be the main factor determining the divertor lifetime. Under the influence of extreme heat fluxes, the surface temperature of plasma facing components can reach some certain threshold, leading to an onset of intense material evaporation. The latter results in formation of cold dense vapor and secondary plasma cloud. This layer effectively absorbs the energy of the incident plasma flow, turning it into its own kinetic and internal energy and radiating it. This so called vapor shieldingmore » is a phenomenon that may help mitigating the erosion during transient events. In particular, the vapor shielding results in saturation of energy (per unit surface area) accumulated by the target during single pulse of heat load at some level E{sub max}. Matching this value is one of the possible tests to verify complicated numerical codes, developed to calculate the erosion rate during abnormal events in tokamaks. The paper presents three very different models of vapor shielding, demonstrating that E{sub max} depends strongly on the heat pulse duration, thermodynamic properties, and evaporation energy of the irradiated target material. While its dependence on the other shielding details such as radiation capabilities of material and dynamics of the vapor cloud is logarithmically weak. The reason for this is a strong (exponential) dependence of the target material evaporation rate, and therefore the “strength” of vapor shield on the target surface temperature. As a result, the influence of the vapor shielding phenomena details, such as radiation transport in the vapor cloud and evaporated material dynamics, on the E{sub max} is virtually completely masked by the strong dependence of the evaporation rate on the target surface temperature. However, the very same details define the amount of evaporated particles, needed to provide an effective shielding to the target, and, therefore, strongly influence resulting erosion rate. Thus, E{sub max} cannot be used for validation of shielding models and codes, aimed at the target material erosion calculations.« less

Measurement of differential cross section of D(3He,p)4He from 0.8 MeV to 3.6 MeV

NASA Astrophysics Data System (ADS)

Zhu, J. P.; Xiao, X.; Yan, S.; Gao, Y.; Xue, J. M.; Wang, Y. G.

2017-12-01

Precise knowledge of the nuclear reaction cross-section is crucial for nuclear reaction analysis methods and its applications. In order to apply nuclear reaction analysis methods to Plasma Facing Materials studies on 4.5 MV electrostatic accelerator at Peking University, differential cross-section for d(3He,p) α at several backward angles was measured with a relative error about ± 6.2 % , gives detailed information at the laboratory angle of 135° from 800 keV to 3600 keV, as well as a rough angular distribution from 130° to 160°.

Erosion and re-deposition of lithium and boron coatings under high-flux plasma bombardment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Abrams, Tyler Wayne

2015-01-01

Lithium and boron coatings are applied to the walls of many tokamaks to enhance performance and protect the underlying substrates. Li and B-coated high-Z substrates are planned for use in NSTX-U and are a candidate plasma-facing component (PFC) for DEMO. However, previous measurements of Li evaporation and thermal sputtering on low-flux devices indicate that the Li temperature permitted on such devices may be unacceptably low. Thus it is crucial to characterize gross and net Li erosion rates under high-flux plasma bombardment. Additionally, no quantitative measurements have been performed of the erosion rate of a boron-coated PFC during plasma bombardment. Amore » realistic model for the compositional evolution of a Li layer under D bombardment was developed that incorporates adsorption, implantation, and diffusion. A model was developed for temperature-dependent mixed-material Li-D erosion that includes evaporation, physical sputtering, chemical sputtering, preferential sputtering, and thermal sputtering. The re-deposition fraction of a Li coating intersecting a linear plasma column was predicted using atomic physics information and by solving the Li continuity equation. These models were tested in the Magnum-PSI linear plasma device at ion fluxes of 10^23-10^24 m^-2 s^-1 and Li surface temperatures less than 800 degrees C. Li erosion was measured during bombardment with a neon plasma that will not chemically react with Li and the results agreed well with the erosion model. Next the ratio of the total D fluence to the areal density of the Li coating was varied to quantify differences in Li erosion under D plasma bombardment as a function of the D concentration. The ratio of D/Li atoms was calculated using the results of MD simulations and good agreement is observed between measurements and the predictions of the mixed-material erosion model. Li coatings are observed to disappear from graphite much faster than from TZM Mo, indicating that fast Li diffusion into the bulk graphite substrate occurred, as predicted. Li re-deposition fractions very close to unity are observed in Magnum-PSI, as predicted by modeling. Finally, predictions of Li coating lifetimes in the NSTX-U divertor are calculated. The gross erosion rate of boron coatings was also measured for the first time in a high-flux plasma device.« less

Impurity re-distribution in the corner regions of the JET divertor

NASA Astrophysics Data System (ADS)

Widdowson, A.; Coad, J. P.; Alves, E.; Baron-Wiechec, A.; Barradas, N. P.; Catarino, N.; Corregidor, V.; Heinola, K.; Krat, S.; Likonen, J.; Matthews, G. F.; Mayer, M.; Petersson, P.; Rubel, M.; Contributors, JET

2017-12-01

The International Thermonuclear Experimental Reactor (ITER) will use a mixture of deuterium (D) and tritium (T) as the fuel to generate power. Since T is both radioactive and expensive the Joint European Torus (JET) has been at the forefront of research to discover how much T is used and where it may be retained within the main reaction chamber. Until the year 2010 the JET plasma facing components were constructed of carbon fibre composites. During the JET carbon (C) phases impurities accumulated at the corners of the divertor located towards the bottom of the chamber in regions shadowed from the plasma where they are very difficult to reach and remove. This build-up of C and the associated H-isotope (including T) retention were of particular concern for future fusion reactors therefore, in 2010 JET changed the wall protection to (mainly) Be and the divertor to tungsten (W)—the JET ITER-like wall (ILW)—the choice of materials for ITER. This paper reveals that with the JET ILW impurities are still accumulating in the shadowed regions, with Be being the majority element, though the overall quantities are very much reduced from those in the C phases. Material will be transported into the shadowed regions principally when the plasma strike points are on the corner tiles, but particles typically have about a 75% probability of reflection from line-of sight surfaces, and multiple reflection/scattering results in deposition over all surfaces.

Flexible Al-doped ZnO films grown on PET substrates using linear facing target sputtering for flexible OLEDs

NASA Astrophysics Data System (ADS)

Jeong, Jin-A.; Shin, Hyun-Su; Choi, Kwang-Hyuk; Kim, Han-Ki

2010-11-01

We report the characteristics of flexible Al-doped zinc oxide (AZO) films prepared by a plasma damage-free linear facing target sputtering (LFTS) system on PET substrates for use as a flexible transparent conducting electrode in flexible organic light-emitting diodes (OLEDs). The electrical, optical and structural properties of LFTS-grown flexible AZO electrodes were investigated as a function of dc power. We obtained a flexible AZO film with a sheet resistance of 39 Ω/squ and an average transmittance of 84.86% in the visible range although it was sputtered at room temperature without activation of the Al dopant. Due to the effective confinement of the high-density plasma between the facing AZO targets, the AZO film was deposited on the PET substrate without plasma damage and substrate heating caused by bombardment of energy particles. Moreover, the flexible OLED fabricated on the AZO/PET substrate showed performance similar to the OLED fabricated on a ITO/PET substrate in spite of a lower work function. This indicates that LFTS is a promising plasma damage-free and low-temperature sputtering technique for deposition of flexible and indium-free AZO electrodes for use in cost-efficient flexible OLEDs.

Development of non-destructive examination techniques for CFC-metal joints in annular geometry and their application to the manufacturing of plasma-facing components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Di Pietro, E.; Visca, E.; Orsini, A.

1995-12-31

The design of plasma-facing components for ITER, as for any of the envisaged next-step machines, relies heavily on the use of brazed junctions to couple armour materials to the heat sink and cooling tubes. Moreover, the typical number of brazed components and the envisaged effects of local overheating due to failure in a single brazed junction stress the importance of having a set of NDE techniques developed that can ensure the flawless quality of the joint. The qualification and application of two NDE techniques (ultrasonic and thermographic analysis) for inspection of CFC-to-metal joints is described with particular regard to themore » annular geometry typical of macroblock/monoblock solutions for divertor high-heat-flux components. The results of the eddy current inspection are not reported. The development has been focused specifically on the joint between carbon-fiber composite and TZM molybdenum alloy; techniques for the production of reference defect samples have been devised and a set of reference defect samples produced. The comparative results of the NDE inspections are reported and discussed, also on the basis of the destructive examination of the samples. The nature and size of relevant and detectable defects are discussed together with hints for a possible NDE strategy for divertor high-heat-flux components.« less

Hydrogen transport behavior of metal coatings for plasma-facing components

NASA Astrophysics Data System (ADS)

Anderl, R. A.; Holland, D. F.; Longhurst, G. R.

1990-12-01

Plasma-facing components for experimental and commercial fusion reactor studies may include cladding or coatings of refractory metals like tungsten on metallic structural substrates such as copper, vanadium alloys and austenitic stainless steel. Issues of safety and fuel economy include the potential for inventory buildup and permeation of tritium implanted into the plasma-facing surface. This paper reports on laboratory-scale studies with 3 keV D +3 ion beams to investigate the hydrogen transport behavior in tungsten coatings on substrates of copper. These experiments entailed measurements of the deuterium re-emission and permeation rates for tungsten, copper, and tungsten-coated copper specimens at temperatures ranging from 638 to 825 K and implanting particle fluxes of approximately 5 × 10 19 D/m 2 s. Diffusion constants and surface recombination coefficients with enhancement factors due to sputtering were obtained from these measurements. These data may be used in calculations to estimate permeation rates and inventory buildups for proposed diverter designs.

First operation with the JET International Thermonuclear Experimental Reactor-like walla)

NASA Astrophysics Data System (ADS)

Neu, R.; Arnoux, G.; Beurskens, M.; Bobkov, V.; Brezinsek, S.; Bucalossi, J.; Calabro, G.; Challis, C.; Coenen, J. W.; de la Luna, E.; de Vries, P. C.; Dux, R.; Frassinetti, L.; Giroud, C.; Groth, M.; Hobirk, J.; Joffrin, E.; Lang, P.; Lehnen, M.; Lerche, E.; Loarer, T.; Lomas, P.; Maddison, G.; Maggi, C.; Matthews, G.; Marsen, S.; Mayoral, M.-L.; Meigs, A.; Mertens, Ph.; Nunes, I.; Philipps, V.; Pütterich, T.; Rimini, F.; Sertoli, M.; Sieglin, B.; Sips, A. C. C.; van Eester, D.; van Rooij, G.; JET-EFDA Contributors

2013-05-01

To consolidate International Thermonuclear Experimental Reactor (ITER) design choices and prepare for its operation, Joint European Torus (JET) has implemented ITER's plasma facing materials, namely, Be for 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 (≈ factor 10) has 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 D2/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 30% power threshold reduction, a distinct minimum density, and a pronounced shape dependence. The L-mode density limit was found to be 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 re-established only when using gas puff levels of a few 1021 es-1. On average, the confinement is lower with the new PFCs, but nevertheless, H factors up to 1 (H-Mode) and 1.3 (at βN≈3, hybrids) have been achieved with W concentrations well below the maximum acceptable level.

First Operation with the JET ITER-Like Wall

NASA Astrophysics Data System (ADS)

Neu, Rudolf

2012-10-01

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 (˜ 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 D2/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 10^21e/s. On average the confinement is lower with the new PFCs, but nevertheless, H factors around 1 (H-Mode) and 1.2 (at βN˜3, Hybrids) have been achieved with W concentrations well below the maximum acceptable level (<10-5).

Crystal-face-selective adsorption of Au nanoparticles onto polycrystalline diamond surfaces.

PubMed

Kondo, Takeshi; Aoshima, Shinsuke; Hirata, Kousuke; Honda, Kensuke; Einaga, Yasuaki; Fujishima, Akira; Kawai, Takeshi

2008-07-15

Crystal-face-selective adsorption of Au nanoparticles (AuNPs) was achieved on polycrystalline boron-doped diamond (BDD) surface via the self-assembly method combined with a UV/ozone treatment. To the best of our knowledge, this is the first report of crystal-face-selective adsorption on an inorganic solid surface. Hydrogen-plasma-treated BDD samples and those followed by UV/ozone treatment for 2 min or longer showed almost no adsorption of AuNP after immersion in the AuNP solution prepared by the citrate reduction method. However, the samples treated by UV/ozone for 10 s showed AuNP adsorption on their (111) facets selectively after the immersion. Moreover, the sample treated with UV/ozone for 40-60 s showed AuNP adsorption on the whole surface. These results indicate that the AuNP adsorption behavior can be controlled by UV/ozone treatment time. This phenomenon was highly reproducible and was applied to a two-step adsorption method, where AuNPs from different batches were adsorbed on the (111) and (100) surface in this order. Our findings may be of great value for the fabrication of advanced nanoparticle-based functional materials via bottom-up approaches with simple macroscale procedures.

Comparative analysis of the possibility of applying low-melting metals with the capillary-porous system in tokamak conditions

NASA Astrophysics Data System (ADS)

Lyublinski, I. E.; Vertkov, A. V.; Semenov, V. V.

2016-12-01

The use of capillary-porous systems (CPSs) with liquid Li, Ga, and Sn is considered as an alternative for solving the problem of creating plasma-facing elements (PFEs) of the fusion neutron source (FNS) and the DEMO-type reactor. The main advantages of CPSs with liquid metal compared with hard materials are their stability with respect to the degradation of properties in tokamak conditions and capability of surface self-restoration. The evaluation of applicability of liquid metals is performed on the basis of the analysis of their physical and chemical properties, the interaction with the tokamak plasma, and constructive and process features of in-vessel elements with CPSs implementing the application of these metals in a tokamak. It is shown that the upper limit of the PFE working temperature for all low-melting metals under consideration lies in the range of 550-600°C. The decisive factor for PFEs with Li is the limitation on the admissible atomic flux into plasma, while for those with Ga and Sn it is the corrosion resistance of construction materials. The upper limit of thermal loads in the steady-state operating mode for the considered promising PFE design with the use of Li, Ga, and Sn is close to 18-20 MW/m2. It is seen from the analysis that the use of metals with a low equilibrium vapor pressure of (Ga, Sn) gives no gain in extension of the region of admissible working temperatures of PFEs. However, with respect to the totality of properties, the possibility of implementing the self-restoration and stabilization effect of the liquid surface, the influence on the plasma discharge parameters, and the ability to protect the PFE surface in conditions of plasma perturbations and disruption, lithium is the most attractive liquid metal to create CPS-based PFEs for the tokamak.

Additive Manufacture of Plasma Diagnostic Components Final Report Phase II

DOE Office of Scientific and Technical Information (OSTI.GOV)

Woodruff, Simon; Romero-Talamas, Carlos; You, Setthivoine

There is now a well-established set of plasma diagnostics (see e.g. [3]), but these remain some of the mostexpensive assemblies in fusion systems since for every system they have to be custom built, and time fordiagnostic development can pace the project. Additive manufacturing (AM) has the potential to decreaseproduction cost and significantly lower design time of fusion diagnostic subsystems, which would realizesignificant cost reduction for standard diagnostics. In some cases, these basic components can be additivelymanufactured for less than 1/100th costs of conventional manufacturing.In our DOE Phase II SBIR, we examined the impact that AM can have on plasma diagnosticmore » cost bytaking 15 separate diagnostics through an engineering design using Conventional Manufacturing (CM) tech-niques, then optimizing the design to exploit the benefits of AM. The impact of AM techniques on cost isfound to be in several areas. First, the cost of materials falls because AM parts can be manufactured withlittle to no waste, and engineered to use less material than CM. Next, the cost of fabrication falls for AMparts relative to CM since the fabrication time can be computed exactly, and often no post-processing isrequired for the part to be functional. We find that AM techniques are well suited for plasma diagnosticssince typical diagnostic complexity comes at no additional cost. Cooling channels, for example, can be builtin to plasma-facing components at no extra cost. Fabrication costs associated with assembly are lower forAM parts because many components can be combined and printed as monoliths, thereby mitigating the needfor alignment or calibration. Finally, the cost of engineering is impacted by exploiting AM design tools thatallow standard components to be customized through web-interfaces. Furthermore, we find that conceptdesign costs can be impacted by scripting interfaces for online engineering design tools.« less

Comparative analysis of the possibility of applying low-melting metals with the capillary-porous system in tokamak conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lyublinski, I. E., E-mail: lyublinski@yandex.ru; Vertkov, A. V., E-mail: avertkov@yandex.ru; Semenov, V. V., E-mail: darkfenix2006@mail.ru

2016-12-15

The use of capillary-porous systems (CPSs) with liquid Li, Ga, and Sn is considered as an alternative for solving the problem of creating plasma-facing elements (PFEs) of the fusion neutron source (FNS) and the DEMO-type reactor. The main advantages of CPSs with liquid metal compared with hard materials are their stability with respect to the degradation of properties in tokamak conditions and capability of surface self-restoration. The evaluation of applicability of liquid metals is performed on the basis of the analysis of their physical and chemical properties, the interaction with the tokamak plasma, and constructive and process features of in-vesselmore » elements with CPSs implementing the application of these metals in a tokamak. It is shown that the upper limit of the PFE working temperature for all low-melting metals under consideration lies in the range of 550–600°Ð¡. The decisive factor for PFEs with Li is the limitation on the admissible atomic flux into plasma, while for those with Ga and Sn it is the corrosion resistance of construction materials. The upper limit of thermal loads in the steady-state operating mode for the considered promising PFE design with the use of Li, Ga, and Sn is close to 18–20 MW/m{sup 2}. It is seen from the analysis that the use of metals with a low equilibrium vapor pressure of (Ga, Sn) gives no gain in extension of the region of admissible working temperatures of PFEs. However, with respect to the totality of properties, the possibility of implementing the self-restoration and stabilization effect of the liquid surface, the influence on the plasma discharge parameters, and the ability to protect the PFE surface in conditions of plasma perturbations and disruption, lithium is the most attractive liquid metal to create CPS-based PFEs for the tokamak.« less

Limiter

DOEpatents

Cohen, Samuel A.; Hosea, Joel C.; Timberlake, John R.

1986-01-01

A limiter with a specially contoured front face accommodates the various power scrape-off distances .lambda..sub.p, which depend on the parallel velocity, V.sub..parallel., of the impacting particles. The front face of the limiter (the plasma-side face) is flat with a central indentation. In addition, the limiter shape is cylindrically symmetric so that the limiter can be rotated for greater heat distribution.

Solid expellant plasma generator

NASA Technical Reports Server (NTRS)

Stone, Nobie H. (Inventor); Poe, Garrett D. (Inventor); Rood, Robert (Inventor)

2010-01-01

An improved solid expellant plasma generator has been developed. The plasma generator includes a support housing, an electrode rod located in the central portion of the housing, and a mass of solid expellant material that surrounds the electrode rod within the support housing. The electrode rod and the solid expellant material are made of separate materials that are selected so that the electrode and the solid expellant material decompose at the same rate when the plasma generator is ignited. This maintains a point of discharge of the plasma at the interface between the electrode and the solid expellant material.

Enhancement of Edge Stability with Lithium Wall Coatings in NSTX

NASA Astrophysics Data System (ADS)

Maingi, R.; Bell, R. E.; Leblanc, B. P.; Kaita, R.; Kaye, S. M.; Kugel, H. W.; Mansfield, D. K.; Osborne, T. H.

2008-11-01

ELM reduction or elimination while maintaining high confinement is essential for ITER, which has been designed for H-mode operation. Large ELMs are thought to be triggered by exceeding either edge current density and/or pressure gradient limits (peeling, ballooning modes). Stability calculations show that spherical tori should have access to higher pressure gradients and pedestal heights than higher R/a tokamaks, owing to access to second stability regimes[...1]. An ELM-free regime was recently observed in the NSTX following the application of lithium onto the graphite plasma facing components[......2]. ELMs were eliminated in phases[.....3], with the resulting pressure gradients and pedestal widths increasing substantially. Calculations with TRANSP have shown that the edge bootstrap current increased substantially, consistent with second stability access. These ELM-free discharges have a substantial improvement in energy confinement, up to the global βN˜ 5.5 limit. * Supported by US DOE DE-FG02-04ER54520, DE-AC-76CH03073, and DE-FC02-04ER54698. [.1] P. B. Snyder, et. al., Plasma Phys. Contr. Fusion 46 (2004) A131. [2] H. W. Kugel, et. al., Phys. Plasma 15 (2008) #056118. [3] D. M. Mansfield, et. al., J. Nucl. Materials (2009) submitted.

Image quality method as a possible way of in situ monitoring of in-vessel mirrors in a fusion reactor

NASA Astrophysics Data System (ADS)

Konovalov, V. G.; Voitsenya, V. S.; Makhov, M. N.; Ryzhkov, I. V.; Shapoval, A. N.; Solodovchenko, S. I.; Stan, A. F.; Bondarenko, V. N.; Donné, A. J. H.; Litnovsky, A.

2016-09-01

The plasma-facing (first) mirrors in ITER will be subject to sputtering and/or contamination with rates that will depend on the precise mirror locations. The resulting influence of both these factors can reduce the mirror reflectance (R) and worsen the transmitted image quality (IQ). This implies that monitoring the mirror quality in situ is an actual desire, and the present work is an attempt towards a solution. The method we propose is able to elucidate the reason for degradation of the mirror reflectance: sputtering by charge exchange atoms or deposition of contaminated layers. In case of deposition of contaminants, the mirror can be cleaned in situ, but a rough mirror (due to sputtering) cannot be used anymore and has to be replaced. To demonstrate the feasibility of the IQ method, it was applied to mirror specimens coated with carbon film in laboratory conditions and to mirrors coated with contaminants during exposure in fusion devices (TRIAM-1M and Tore Supra), as well as to mirrors of different materials exposed to sputtering by plasma ions in the DSM-2 plasma stand (in IPP NSC KIPT).

Fuel Retention Improvement at High Temperatures in Tungsten-Uranium Dioxide Dispersion Fuel Elements by Plasma-Spray Cladding

NASA Technical Reports Server (NTRS)

Grisaffe, Salvatore J.; Caves, Robert M.

1964-01-01

An investigation was undertaken to determine the feasibility of depositing integrally bonded plasma-sprayed tungsten coatings onto 80-volume-percent tungsten - 20-volume-percent uranium dioxide composites. These composites were face clad with thin tungsten foil to inhibit uranium dioxide loss at elevated temperatures, but loss at the unclad edges was still significant. By preheating the composite substrates to approximately 3700 degrees F in a nitrogen environment, metallurgically bonded tungsten coatings could be obtained directly by plasma spraying. Furthermore, even though these coatings were thin and somewhat porous, they greatly inhibited the loss of uranium dioxide. For example, a specimen that was face clad but had no edge cladding lost 5.8 percent uranium dioxide after 2 hours at 4750 dgrees F in flowing hydrogen. A similar specimen with plasma-spray-coated edges, however, lost only 0.75 percent uranium dioxide under the same testing conditions.

High heat-flux self-rotating plasma-facing component: Concept and loading test in TEXTOR

NASA Astrophysics Data System (ADS)

Terra, A.; Sergienko, G.; Hubeny, M.; Huber, A.; Mertens, Ph.; Philipps, V.; The Textor Team

2015-08-01

This contribution reports on the concept of a circular self-rotating and temperature self-stabilising plasma-facing component (PFC), and test of a related prototype in TEXTOR tokamak. This PFC uses the Lorentz force induced by plasma current and magnet field (J × B) to create a torque applied on metallic discs which produce a rotational movement. Additional thermionic current, present at high operation temperatures, brings additional temperature stabilisation ability. This self-rotating disk limiter was exposed to plasma in the TEXTOR tokamak under different radial positions to vary the heat flux. This disk structure shows the interesting ability to stabilise its maximum temperature through the fact that the self-induced rotation is modulated by the thermal emission current. It was observed that the rotation speed increased following both the current collected by the limiter, and the temperature of the tungsten disks.

Ion radiation albedo effect: influence of surface roughness on ion implantation and sputtering of materials

NASA Astrophysics Data System (ADS)

Li, Yonggang; Yang, Yang; Short, Michael P.; Ding, Zejun; Zeng, Zhi; Li, Ju

2017-01-01

In fusion devices, ion retention and sputtering of materials are major concerns in the selection of compatible plasma-facing materials (PFMs), especially in the context of their microstructural conditions and surface morphologies. We demonstrate how surface roughness changes ion implantation and sputtering of materials under energetic ion irradiation. Using a new, sophisticated 3D Monte Carlo (MC) code, IM3D, and a random rough surface model, ion implantation and the sputtering yields of tungsten (W) with a surface roughness varying between 0-2 µm have been studied for irradiation by 0.1-1 keV D+, He+ and Ar+ ions. It is found that both ion backscattering and sputtering yields decrease with increasing roughness; this is hereafter called the ion radiation albedo effect. This effect is mainly dominated by the direct, line-of-sight deposition of a fraction of emitted atoms onto neighboring asperities. Backscattering and sputtering increase with more oblique irradiation angles. We propose a simple analytical formula to relate rough-surface and smooth-surface results.

Experiment attributes to establish tube with twisted tape insert performance cooling plasma facing components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clark, Emily; Ramirez, Emilio; Ruggles, Art E.

The modeling capability for tubes with twisted tape inserts is reviewed with reference to the application of cooling plasma facing components in magnetic confinement fusion devices. The history of experiments examining the cooling performance of tubes with twisted tape inserts is reviewed with emphasis on the manner of heating, flow stability limits and the details of the test section and fluid delivery system. Models for heat transfer, burnout, and onset of net vapor generation in straight tube flows and tube with twisted tape are compared. As a result, the gaps in knowledge required to establish performance limits of the plasmamore » facing components are identified and attributes of an experiment to close those gaps are presented.« less

Experiment attributes to establish tube with twisted tape insert performance cooling plasma facing components

DOE PAGES

Clark, Emily; Ramirez, Emilio; Ruggles, Art E.; ...

2015-08-18

The modeling capability for tubes with twisted tape inserts is reviewed with reference to the application of cooling plasma facing components in magnetic confinement fusion devices. The history of experiments examining the cooling performance of tubes with twisted tape inserts is reviewed with emphasis on the manner of heating, flow stability limits and the details of the test section and fluid delivery system. Models for heat transfer, burnout, and onset of net vapor generation in straight tube flows and tube with twisted tape are compared. As a result, the gaps in knowledge required to establish performance limits of the plasmamore » facing components are identified and attributes of an experiment to close those gaps are presented.« less

Method of laminating structural members

NASA Technical Reports Server (NTRS)

Heier, W. C. (Inventor)

1974-01-01

A laminate is obtained by providing a lightweight core material, such as a honeycombed plastic or metal, within the cavity defined by an annular mold cavity frame. Face sheets, which are to be bonded to the core material, are provided on opposite sides of the frame and extend over the frame, thus sealing the core material in the cavity. An adhesive is provided between the core material and the face sheets and the combined thickness of the core material and adhesive is a close fit within the opposed face sheets. A gas tight seal, such as an O-ring gasket, is provided between the frame and the face sheet members to form a gas tight cavity between the face sheet members and the frame. External heat and pressure are used to bond the face sheets to the core material. Gas pressure is introduced into the sealed cavity to minimize out-gasing of the adhesive.

Influence of the first wall material on the particle fuelling in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Lunt, T.; Reimold, F.; Wolfrum, E.; Carralero, D.; Feng, Y.; Schmid, K.; the ASDEX Upgrade Team

2017-05-01

In the period from 2002 to 2007 the material of the plasma facing components (PFCs) of ASDEX Upgrade (AUG) was changed from carbon (C) to tungsten (W). Comparing the measured density profiles of low-density L-mode discharges with little or no gas puff before and after this modification, a significantly higher pedestal-top density was found for W PFCs together with a steeper gradient and a lower pedestal temperature. This change can be explained by larger particle- and energy reflection coefficients for D on W compared to D on C, as shown by EMC3-EIRENE simulations of AUG discharges in similar conditions on a computational grid extending to the main chamber first wall. In the simulations, a change of the wall material at fixed separatrix density indeed shows that for W PFCs more neutrals cross the separatrix, resulting in a steeper density gradient. Analysis of the source resolved and poloidally resolved neutral flux densities across the separatrix show a dominant contribution of the divertor targets to the fuelling profile in the simulation of the low density case. Increasing the density decreases the electron temperature at the target and therefore the potential drop in the electrostatic sheath as well as the energy of the ions impinging on the surface. Neutrals with ∼eV energies, able to reach the separatrix, are then only produced via molecular dissociation processes in the plasma volume independently of the PFC material. Also the contribution of the main chamber PFCs to the fuelling is observed to increase at higher densities.

Electromagnetic Torque in Tokamaks with Toroidal Asymmetries

DOE Office of Scientific and Technical Information (OSTI.GOV)

Logan, Nikolas Christopher

2015-01-01

Lithium and boron coatings are applied to the walls of many tokamaks to enhance performance and protect the underlying substrates. Li and B-coated high-Z substrates are planned for use in NSTX-U and are a candidate plasma-facing component (PFC) for DEMO. However, previous measurements of Li evaporation and thermal sputtering on low-flux devices indicate that the Li temperature permitted on such devices may be unacceptably low. Thus it is crucial to characterize gross and net Li erosion rates under high-flux plasma bombardment. Additionally, no quantitative measurements have been performed of the erosion rate of a boron-coated PFC during plasma bombardment. Amore » realistic model for the compositional evolution of a Li layer under D bombardment was developed that incorporates adsorption, implantation, and diffusion. A model was developed for temperature-dependent mixed-material Li-D erosion that includes evaporation, physical sputtering, chemical sputtering, preferential sputtering, and thermal sputtering. The re-deposition fraction of a Li coating intersecting a linear plasma column was predicted using atomic physics information and by solving the Li continuity equation. These models were tested in the Magnum-PSI linear plasma device at ion fluxes of 10^23-10^24 m^-2 s^-1 and Li surface temperatures less than 800 degrees C. Li erosion was measured during bombardment with a neon plasma that will not chemically react with Li and the results agreed well with the erosion model. Next the ratio of the total D fluence to the areal density of the Li coating was varied to quantify differences in Li erosion under D plasma bombardment as a function of the D concentration. The ratio of D/Li atoms was calculated using the results of MD simulations and good agreement is observed between measurements and the predictions of the mixed-material erosion model. Li coatings are observed to disappear from graphite much faster than from TZM Mo, indicating that fast Li diffusion into the bulk graphite substrate occurred, as predicted. Li re-deposition fractions very close to unity are observed in Magnum-PSI, as predicted by modeling. Finally, predictions of Li coating lifetimes in the NSTX-U divertor are calculated. The gross erosion rate of boron coatings was also measured for the first time in a high-flux plasma device.« less

Advanced smart tungsten alloys for a future fusion power plant

NASA Astrophysics Data System (ADS)

Litnovsky, A.; Wegener, T.; Klein, F.; Linsmeier, Ch; Rasinski, M.; Kreter, A.; Tan, X.; Schmitz, J.; Mao, Y.; Coenen, J. W.; Bram, M.; Gonzalez-Julian, J.

2017-06-01

The severe particle, radiation and neutron environment in a future fusion power plant requires the development of advanced plasma-facing materials. At the same time, the highest level of safety needs to be ensured. The so-called loss-of-coolant accident combined with air ingress in the vacuum vessel represents a severe safety challenge. In the absence of a coolant the temperature of the tungsten first wall may reach 1200 °C. At such a temperature, the neutron-activated radioactive tungsten forms volatile oxide which can be mobilized into atmosphere. Smart tungsten alloys are being developed to address this safety issue. Smart alloys should combine an acceptable plasma performance with the suppressed oxidation during an accident. New thin film tungsten-chromium-yttrium smart alloys feature an impressive 105 fold suppression of oxidation compared to that of pure tungsten at temperatures of up to 1000 °C. Oxidation behavior at temperatures up to 1200 °C, and reactivity of alloys in humid atmosphere along with a manufacturing of reactor-relevant bulk samples, impose an additional challenge in smart alloy development. First exposures of smart alloys in steady-state deuterium plasma were made. Smart tungsten-chroimium-titanium alloys demonstrated a sputtering resistance which is similar to that of pure tungsten. Expected preferential sputtering of alloying elements by plasma ions was confirmed experimentally. The subsequent isothermal oxidation of exposed samples did not reveal any influence of plasma exposure on the passivation of alloys.

Antimicrobial cotton textiles with robust superhydrophobicity via plasma for oily water separation

NASA Astrophysics Data System (ADS)

Zhang, Ming; Pang, Jiuyin; Bao, Wenhui; Zhang, Wenbo; Gao, He; Wang, Chengyu; Shi, Junyou; Li, Jian

2017-10-01

During these decades, functional materials are facing the severe challenge of their weak surface structure. To solve this problem, plasma technology and spraying technology were utilized to improve the bonding effect between cotton substrates and coating structures. Herein, silica/silver nanoparticles (SiO2/Ag NPs) were prepared and introduced to the nano-/micro- structures on sample surface by spraying technology in the existence of polyurethane adhesive. Then the circles of spraying procedure containing adhesive and SiO2/Ag NPs had been discussed. After further fluorination, the samples still displayed an excellent waterproof property even after abrasion test with sand paper and various washing test by its solvent-acetone or harsh liquids with strong acidity/alkalinity, indicating their robust surfaces structures. More importantly, this product displayed the outstanding performance no matter in laboratory oil/water filtration or the extensive oil leakage and spill. At last, our modification also endowed the cotton sample with great antimicrobial property.

Safety and diagnostic systems on the Liquid Lithium Test Stand (LLTS)

NASA Astrophysics Data System (ADS)

Schwartz, J. A.; Jaworski, M. A.; Ellis, R.; Kaita, R.; Mozulay, R.

2013-10-01

The Liquid Lithium Test Stand (LLTS) is a test bed for development of flowing liquid lithium systems for plasma-facing components at PPPL. LLTS is designed to test operation of liquid lithium under vacuum, including flowing, solidifying (such as would be the case at the end of plasma operations), and re-melting. Constructed of stainless steel, LLTS is a closed loop of pipe with two reservoirs and a pump, as well as diagnostics for temperature, flow rate, and pressure. Since liquid lithium is a highly reactive material, special care must be taken when designing such a system. These include a permanent-magnet MHD pump and MHD flow meter that have no mechanical components in direct contact with the liquid lithium. The LLTS also includes an expandable 24-channel leak-detector interlock system which cuts power to heaters and the pump if any lithium leaks from a pipe joint. Design for the interlock systems and flow meter are presented. This work is supported by US DOE Contract DE-AC02-09CH11466.

Face Detection Technique as Interactive Audio/Video Controller for a Mother-Tongue-Based Instructional Material

NASA Astrophysics Data System (ADS)

Guidang, Excel Philip B.; Llanda, Christopher John R.; Palaoag, Thelma D.

2018-03-01

Face Detection Technique as a strategy in controlling a multimedia instructional material was implemented in this study. Specifically, it achieved the following objectives: 1) developed a face detection application that controls an embedded mother-tongue-based instructional material for face-recognition configuration using Python; 2) determined the perceptions of the students using the Mutt Susan’s student app review rubric. The study concludes that face detection technique is effective in controlling an electronic instructional material. It can be used to change the method of interaction of the student with an instructional material. 90% of the students perceived the application to be a great app and 10% rated the application to be good.

Softening due to Grain Boundary Cavity Formation and its Competition with Hardening in Helium Implanted Nanocrystalline Tungsten

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cunningham, W. Streit; Gentile, Jonathan M.; El-Atwani, Osman

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 viamore » 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.« less

Softening due to Grain Boundary Cavity Formation and its Competition with Hardening in Helium Implanted Nanocrystalline Tungsten

DOE PAGES

Cunningham, W. Streit; Gentile, Jonathan M.; El-Atwani, Osman; ...

2018-02-13

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 viamore » 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.« less

Sequential and simultaneous thermal and particle exposure of tungsten

NASA Astrophysics Data System (ADS)

Steudel, I.; Huber, A.; Kreter, A.; Linke, J.; Sergienko, G.; Unterberg, B.; Wirtz, M.

2016-02-01

The broad array of expected loading conditions in a fusion reactor such as ITER necessitates high requirements on the plasma facing materials (PFMs). Tungsten, the PFM for the divertor region, the most affected part of the in-vessel components, must thus sustain severe, distinct exposure conditions. Accordingly, comprehensive experiments investigating sequential and simultaneous thermal and particle loads were performed on double forged pure tungsten, not only to investigate whether the thermal and particle loads cause damage but also if the sequence of exposure maintains an influence. The exposed specimens showed various kinds of damage such as roughening, blistering, and cracking at a base temperature where tungsten could be ductile enough to compensate the induced stresses exclusively by plastic deformation (Pintsuk et al 2011 J. Nucl. Mater. 417 481-6). It was found out that hydrogen has an adverse effect on the material performance and the loading sequence on the surface modification.

Compact steady-state and high-flux Falcon ion source for tests of plasma-facing materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Girka, O.; Bizyukov, I.; Sereda, K.

2012-08-15

This paper describes the design and operation of the Falcon ion source. It is based on conventional design of anode layer thrusters. This ion source is a versatile, compact, affordable, and highly functional in the research field of the fusion materials. The reversed magnetic field configuration of the source allows precise focusing of the ion beam into small spot of Almost-Equal-To 3 mm and also provides the limited capabilities for impurity mass-separation. As the result, the source generates steady-state ion beam, which irradiates surface with high heat (0.3 - 21 MW m{sup -2}) and particle fluxes (4 Multiplication-Sign 10{sup 21}-more » 3 Multiplication-Sign 10{sup 23} m{sup -2}s{sup -1}), which approaches the upper limit for the flux range expected in ITER.« less

An experiment on the dynamics of ion implantation and sputtering of surfaces

NASA Astrophysics Data System (ADS)

Wright, G. M.; Barnard, H. A.; Kesler, L. A.; Peterson, E. E.; Stahle, P. W.; Sullivan, R. M.; Whyte, D. G.; Woller, K. B.

2014-02-01

A major impediment towards a better understanding of the complex plasma-surface interaction is the limited diagnostic access to the material surface while it is undergoing plasma exposure. The Dynamics of ION Implantation and Sputtering Of Surfaces (DIONISOS) experiment overcomes this limitation by uniquely combining powerful, non-perturbing ion beam analysis techniques with a steady-state helicon plasma exposure chamber, allowing for real-time, depth-resolved in situ measurements of material compositions during plasma exposure. Design solutions are described that provide compatibility between the ion beam analysis requirements in the presence of a high-intensity helicon plasma. The three primary ion beam analysis techniques, Rutherford backscattering spectroscopy, elastic recoil detection, and nuclear reaction analysis, are successfully implemented on targets during plasma exposure in DIONISOS. These techniques measure parameters of interest for plasma-material interactions such as erosion/deposition rates of materials and the concentration of plasma fuel species in the material surface.

An experiment on the dynamics of ion implantation and sputtering of surfaces.

PubMed

Wright, G M; Barnard, H A; Kesler, L A; Peterson, E E; Stahle, P W; Sullivan, R M; Whyte, D G; Woller, K B

2014-02-01

A major impediment towards a better understanding of the complex plasma-surface interaction is the limited diagnostic access to the material surface while it is undergoing plasma exposure. The Dynamics of ION Implantation and Sputtering Of Surfaces (DIONISOS) experiment overcomes this limitation by uniquely combining powerful, non-perturbing ion beam analysis techniques with a steady-state helicon plasma exposure chamber, allowing for real-time, depth-resolved in situ measurements of material compositions during plasma exposure. Design solutions are described that provide compatibility between the ion beam analysis requirements in the presence of a high-intensity helicon plasma. The three primary ion beam analysis techniques, Rutherford backscattering spectroscopy, elastic recoil detection, and nuclear reaction analysis, are successfully implemented on targets during plasma exposure in DIONISOS. These techniques measure parameters of interest for plasma-material interactions such as erosion/deposition rates of materials and the concentration of plasma fuel species in the material surface.

Plasma technologies application for building materials surface modification

NASA Astrophysics Data System (ADS)

Volokitin, G. G.; Skripnikova, N. K.; Volokitin, O. G.; Shehovtzov, V. V.; Luchkin, A. G.; Kashapov, N. F.

2016-01-01

Low temperature arc plasma was used to process building surface materials, such as silicate brick, sand lime brick, concrete and wood. It was shown that building surface materials modification with low temperature plasma positively affects frost resistance, water permeability and chemical resistance with high adhesion strength. Short time plasma processing is rather economical than traditional processing thermic methods. Plasma processing makes wood surface uniquely waterproof and gives high operational properties, dimensional and geometrical stability. It also increases compression resistance and decreases inner tensions level in material.

The hybrid reactor project based on the straight field line mirror concept

NASA Astrophysics Data System (ADS)

Ågren, O.; Noack, K.; Moiseenko, V. E.; Hagnestâl, A.; Källne, J.; Anglart, H.

2012-06-01

The straight field line mirror (SFLM) concept is aiming towards a steady-state compact fusion neutron source. Besides the possibility for steady state operation for a year or more, the geometry is chosen to avoid high loads on materials and plasma facing components. A comparatively small fusion hybrid device with "semi-poor" plasma confinement (with a low fusion Q factor) may be developed for industrial transmutation and energy production from spent nuclear fuel. This opportunity arises from a large fission to fusion energy multiplication ratio, Qr = Pfis/Pfus>>1. The upper bound on Qr is primarily determined by geometry and reactor safety. For the SFLM, the upper bound is Qr≈150, corresponding to a neutron multiplicity of keff=0.97. Power production in a mirror hybrid is predicted for a substantially lower electron temperature than the requirement Te≈10 keV for a fusion reactor. Power production in the SFLM seems possible with Q≈0.15, which is 10 times lower than typically anticipated for hybrids (and 100 times smaller than required for a fusion reactor). This relaxes plasma confinement demands, and broadens the range for use of plasmas with supra-thermal ions in hybrid reactors. The SFLM concept is based on a mirror machine stabilized by qudrupolar magnetic fields and large expander tanks beyond the confinement region. The purpose of the expander tanks is to distribute axial plasma loss flow over a sufficiently large area so that the receiving plates can withstand the heat. Plasma stability is not relying on a plasma flow into the expander regions. With a suppressed plasma flow into the expander tanks, a possibility arise for higher electron temperature. A brief presentation will be given on basic theory for the SFLM with plasma stability and electron temperature issues, RF heating computations with sloshing ion formation, neutron transport computations with reactor safety margins and material load estimates, magnetic coil designs as well as a discussion on the implications of the geometry for possible diagnostics. Reactor safety issues are addressed and a vertical orientation of the device could assist passive coolant circulation. Specific attention is put to a device with a 25 m long confinement region and 40 cm plasma radius in the mid-plane. In an optimal case (keff = 0.97) with a fusion power of only 10 MW, such a device may be capable of producing a power of 1.5 GWth.

In-Vessel Tritium Retention and Removal in ITER-FEAT

NASA Astrophysics Data System (ADS)

Federici, G.; Brooks, J. N.; Iseli, M.; Wu, C. H.

Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruptions, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ˜350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R&D work is required to narrow the remaining uncertainties are also briefly discussed.

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. Ono, M. Jaworski, R. Kaita, C. N. Skinner, J.P. Allain, R. Maingi, F. Scotti, V.A. Soukhanovskii, and the NSTX-U Team

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 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.more » Combined with the radiative divertor concept, the snow-flake configuration has reduced the divertor heat flux by 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, Hmode 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.« less

Recent progress in plasma-assisted synthesis and modification of 2D materials

NASA Astrophysics Data System (ADS)

Han, Zhao Jun; Murdock, Adrian T.; Seo, Dong Han; Bendavid, Avi

2018-07-01

Plasma represents an important technique for both the synthesis and modification of two-dimensional (2D) materials, owing to the unique plasma-material interactions which can enable effective energy transfer at the nanoscale. Non-equilibrium and non-thermal plasma techniques have been widely applied on various 2D materials, including graphene, silicene, germanene, phosphorene, hexagonal boron nitride (h-BN), and transition metal dichalcogenides such as MoS2 and WS2. Here, we review the recent progress in plasma-assisted synthesis and modification (e.g. functionalisation, doping and etching) of 2D materials and discuss the potential applications of this unique branch of 2D materials. Challenges and future research opportunities in the relevant research field are also discussed. The primary aim of this Review is to provide a better understanding of the plasma-assisted processes and to promote the utilization of 2D materials for advanced electronic, optoelectronic, sensing and energy storage applications.

Limiter

DOEpatents

Cohen, S.A.; Hosea, J.C.; Timberlake, J.R.

1984-10-19

A limiter with a specially contoured front face is provided. The front face of the limiter (the plasma-side face) is flat with a central indentation. In addition, the limiter shape is cylindrically symmetric so that the limiter can be rotated for greater heat distribution. This limiter shape accommodates the various power scrape-off distances lambda p, which depend on the parallel velocity, V/sub parallel/, of the impacting particles.

Towards a programme of testing and qualification for structural and plasma-facing materials in ‘fusion neutron’ environments

NASA Astrophysics Data System (ADS)

Stork, D.; Heidinger, R.; Muroga, T.; Zinkle, S. J.; Moeslang, A.; Porton, M.; Boutard, J.-L.; Gonzalez, S.; Ibarra, A.

2017-09-01

Materials damage by 14.1MeV neutrons from deuterium-tritium (D-T) fusion reactions can only be characterised definitively by subjecting a relevant configuration of test materials to high-intensity ‘fusion-neutron spectrum sources’, i.e. those simulating closely D-T fusion-neutron spectra. This provides major challenges to programmes to design and construct a demonstration fusion reactor prior to having a large-scale, high-intensity source of such neutrons. In this paper, we discuss the different aspects related to these ‘relevant configuration’ tests, including: • generic issues in materials qualification/validation, comparing safety requirements against those of investment protection; • lessons learned from the fission programme, enabling a reduced fusion materials testing programme; • the use and limitations of presently available possible irradiation sources to optimise a fusion neutron testing program including fission-neutron irradiation of isotopically and chemically tailored steels, ion damage by high-energy helium ions and self-ion beams, or irradiation studies with neutron sources of non-fusion spectra; and • the different potential sources of simulated fusion neutron spectra and the choice using stripping reactions from deuterium-beam ions incident on light-element targets.

Plasma reactor waste management systems

NASA Technical Reports Server (NTRS)

Ness, Robert O., Jr.; Rindt, John R.; Ness, Sumitra R.

1992-01-01

The University of North Dakota is developing a plasma reactor system for use in closed-loop processing that includes biological, materials, manufacturing, and waste processing. Direct-current, high-frequency, or microwave discharges will be used to produce plasmas for the treatment of materials. The plasma reactors offer several advantages over other systems, including low operating temperatures, low operating pressures, mechanical simplicity, and relatively safe operation. Human fecal material, sunflowers, oats, soybeans, and plastic were oxidized in a batch plasma reactor. Over 98 percent of the organic material was converted to gaseous products. The solids were then analyzed and a large amount of water and acid-soluble materials were detected. These materials could possibly be used as nutrients for biological systems.

Effects of lithium-implantation on the hydrogen retention in both a-C:H and a-SiC:H materials submitted to deuterium bombardment

NASA Astrophysics Data System (ADS)

Barbier, G.; Ross, G. G.; El Khakani, M. A.; Chevarier, N.; Chevarier, A.

1997-02-01

The hydrogen release in plasma facing materials is a challenging problem for the hydrogen recycling. The hydrogen desorption from the a-C:H and a-SiC:H materials induced by deuterium bombardment has been investigated. Prior to the deuterium bombardment, both materials were implanted with different fluences of lithium ions. Before and after each irradiation, depth profiles of H, Li and deuterium were determined by nuclear microanalysis. After deuterium bombardment, it is shown that the retention of the initial hydrogen in both materials was enhanced by increasing the total dose of the implanted Li. For the a-C:H samples, the hydrogen desorption under deuterium bombardment was strongly reduced by lithium implantation. This effect was also evidenced in a-SiC:H samples, even though it is less spectacular than in a-C:H. Also, nuclear analyses showed that the retained dose of deuterium decreases when the lithium concentration increases. This could be a result of the formation of LiH bonds which occurs to the detriment of deuterium retention in both a-C:H and a-SiC:H materials. Preliminary results of both materials exposed to TdeV tokamak discharges confirms the role of Li in hydrogen retention, already observed in deuterium bombardment exposure.

Impact and effects of simultaneous MeV-ion irradiation and helium plasma exposure to the formation of tungsten nano-tendrils

NASA Astrophysics Data System (ADS)

Wright, Graham; Kesler, Leigh Ann; Whyte, Dennis

2013-10-01

The extrusion of nano-tendrils from high temperature (>1000 K) tungsten (W) targets exposed to helium (He) plasma ions remains a concern for future fusion reactors. Previous work on the Alcator C-Mod tokamak has demonstrated it is possible to form these structures in a tokamak environment. However, one area where Alcator C-Mod and a fusion reactor differ is total neutron flux at the wall and the displacement damage these neutrons produce in the plasma-facing materials. This dsiplacement damage may affect the size and number He bubbles precipitating in the W target, which is a key factor in the formation and growth of the nano-tendrils. The DIONISOS experiment directly measures the impact of the displacement damage by simultaneously bombarding high temperature W targets with MeV-range ions (to simulate the displacement damage caused by neutron flux) and high flux of He plasma ions. Different combinations of irradiating ion species and W target temperatures are used to vary the different processes and rates that are involved such as He trapping rate, vacancy production and annealing rates, and nano-tendril growth rate. The nano-tendril growth is characterized by SEM imaging and focused ion beam (FIB) cross-sectioning and compared to nano-tendril formation without the presence of the irradiating ion beam. This work is supported by US DOE award DE-SC00-02060.

Interactions of Deuterium Plasma with Lithiated and Boronized Surfaces in NSTX-U

NASA Astrophysics Data System (ADS)

Krstic, Predrag

2015-09-01

The main research goal of the presented research has been to understand the changes in surface composition and chemistry at the nanoscopic temporal and spatial scales for long pulse Plasma Facing Components (PFCs) and link these to the overall machine performance of the National Spherical Torus Experiment Upgrade (NSTX-U). A study is presented of the lithium surface science, with atomic spatial and temporal resolutions. The dynamic surface responds and evolves in a mixed material environments (D, Li, C, B, O, Mo, W) with impingement of plasma particles in the energy range below 100 eV. The results, obtained by quantum-classical molecular dynamics, include microstructure changes, erosion, surface chemistry, deuterium implantation and permeation. Main objectives of the research are i) a comparison of Li and B deposition on carbon, ii) the role of oxygen and other impurities e.g. boron, carbon in the lithium performance, and iii) how this performance will change when lithium is applied to a high-Z refractory metal substrate (Mo, W). In addition to predicting and understanding the phenomenology of the processes, we will show plasma induced erosion of PFCs, including chemical and physical sputtering yields at various temperatures (300-700 K) as well as deuterium uptake/recycling. This work is supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Science, Award Number DE-SC0013752.

Process for forming exoergic structures with the use of a plasma

DOEpatents

Kelly, M.D.

1987-05-29

A method of forming exoergic structures, as well as exoergic structures produced by the method, is provided. The method comprises the steps of passing a plasma-forming gas through a plasma spray gun, forming a plasma spray, introducing exoergic material into the plasma spray and directing the plasma spray toward a substrate, and allowing the exoergic material to become molten in the plasma spray and to thereafter impinge on the substrate to form a solid mass of exoergic material, the shape of which corresponds to the shape of the substrate.

Method for minimizing decarburization and other high temperature oxygen reactions in a plasma sprayed material

DOEpatents

Lenling, William J.; Henfling, Joseph A.; Smith, Mark F.

1993-06-08

A method is disclosed for spray coating material which employs a plasma gun that has a cathode, an anode, an arc gas inlet, a first powder injection port, and a second powder injection port. A suitable arc gas is introduced through the arc gas inlet, and ionization of the arc gas between the cathode and the anode forms a plasma. The plasma is directed to emenate from an open-ended chamber defined by the boundary of the anode. A coating is deposited upon a base metal part by suspending a binder powder within a carrier gas that is fed into the plasma through the first powder injection port; a material subject to degradation by high temperature oxygen reactions is suspended within a carrier gas that is fed into the plasma through the second injection port. The material fed through the second injection port experiences a cooler portion of the plasma and has a shorter dwell time within the plasma to minimize high temperature oxygen reactions. The material of the first port and the material of the second port intermingle within the plasma to form a uniform coating having constituent percentages related to the powder-feed rates of the materials through the respective ports.

Design, Analysis and R&D of the EAST In-Vessel Components

NASA Astrophysics Data System (ADS)

Yao, Damao; Bao, Liman; Li, Jiangang; Song, Yuntao; Chen, Wenge; Du, Shijun; Hu, Qingsheng; Wei, Jing; Xie, Han; Liu, Xufeng; Cao, Lei; Zhou, Zibo; Chen, Junling; Mao, Xinqiao; Wang, Shengming; Zhu, Ning; Weng, Peide; Wan, Yuanxi

2008-06-01

In-vessel components are important parts of the EAST superconducting tokamak. They include the plasma facing components, passive plates, cryo-pumps, in-vessel coils, etc. The structural design, analysis and related R&D have been completed. The divertor is designed in an up-down symmetric configuration to accommodate both double null and single null plasma operation. Passive plates are used for plasma movement control. In-vessel coils are used for the active control of plasma vertical movements. Each cryo-pump can provide an approximately 45 m3/s pumping rate at a pressure of 10-1 Pa for particle exhaust. Analysis shows that, when a plasma current of 1 MA disrupts in 3 ms, the EM loads caused by the eddy current and the halo current in a vertical displacement event (VDE) will not generate an unacceptable stress on the divertor structure. The bolted divertor thermal structure with an active cooling system can sustain a load of 2 MW/m2 up to a 60 s operation if the plasma facing surface temperature is limited to 1500 °C. Thermal testing and structural optimization testing were conducted to demonstrate the analysis results.

High-Performance Computational Modeling of ICRF Physics and Plasma-Surface Interactions in Alcator C-Mod

NASA Astrophysics Data System (ADS)

Jenkins, Thomas; Smithe, David

2016-10-01

Inefficiencies and detrimental physical effects may arise in conjunction with ICRF heating of tokamak plasmas. Large wall potential drops, associated with sheath formation near plasma-facing antenna hardware, give rise to high-Z impurity sputtering from plasma-facing components and subsequent radiative cooling. Linear and nonlinear wave excitations in the plasma edge/SOL also dissipate injected RF power and reduce overall antenna efficiency. Recent advances in finite-difference time-domain (FDTD) modeling techniques allow the physics of localized sheath potentials, and associated sputtering events, to be modeled concurrently with the physics of antenna near- and far-field behavior and RF power flow. The new methods enable time-domain modeling of plasma-surface interactions and ICRF physics in realistic experimental configurations at unprecedented spatial resolution. We present results/animations from high-performance (10k-100k core) FDTD/PIC simulations spanning half of Alcator C-Mod at mm-scale resolution, exploring impurity production due to localized sputtering (in response to self-consistent sheath potentials at antenna surfaces) and the physics of parasitic slow wave excitation near the antenna hardware and SOL. Supported by US DoE (Award DE-SC0009501) and the ALCC program.

Retention and release of hydrogen isotopes in tungsten plasma-facing components: the role of grain boundaries and the native oxide layer from a joint experiment-simulation integrated approach

NASA Astrophysics Data System (ADS)

Hodille, E. A.; Ghiorghiu, F.; Addab, Y.; Založnik, A.; Minissale, M.; Piazza, Z.; Martin, C.; Angot, T.; Gallais, L.; Barthe, M.-F.; Becquart, C. S.; Markelj, S.; Mougenot, J.; Grisolia, C.; Bisson, R.

2017-07-01

Fusion fuel retention (trapping) and release (desorption) from plasma-facing components are critical issues for ITER and for any future industrial demonstration reactors such as DEMO. Therefore, understanding the fundamental mechanisms behind the retention of hydrogen isotopes in first wall and divertor materials is necessary. We developed an approach that couples dedicated experimental studies with modelling at all relevant scales, from microscopic elementary steps to macroscopic observables, in order to build a reliable and predictive fusion reactor wall model. This integrated approach is applied to the ITER divertor material (tungsten), and advances in the development of the wall model are presented. An experimental dataset, including focused ion beam scanning electron microscopy, isothermal desorption, temperature programmed desorption, nuclear reaction analysis and Auger electron spectroscopy, is exploited to initialize a macroscopic rate equation wall model. This model includes all elementary steps of modelled experiments: implantation of fusion fuel, fuel diffusion in the bulk or towards the surface, fuel trapping on defects and release of trapped fuel during a thermal excursion of materials. We were able to show that a single-trap-type single-detrapping-energy model is not able to reproduce an extended parameter space study of a polycrystalline sample exhibiting a single desorption peak. It is therefore justified to use density functional theory to guide the initialization of a more complex model. This new model still contains a single type of trap, but includes the density functional theory findings that the detrapping energy varies as a function of the number of hydrogen isotopes bound to the trap. A better agreement of the model with experimental results is obtained when grain boundary defects are included, as is consistent with the polycrystalline nature of the studied sample. Refinement of this grain boundary model is discussed as well as the inclusion in the model of a thin defective oxide layer following the experimental observation of the presence of an oxygen layer on the surface even after annealing to 1300 K.

Buckling coefficients for simply supported and camped flat, rectangular sandwich panels under edgewise compression

Treesearch

Edward W. Kuenzi; Charles B. Norris; Paul M. Jenkinson

1964-01-01

“This report presents curves of coefficients and formulas for use in calculating the buckling of flat panels of sandwich construction under edgewise compressive loads. The curves were derived for sandwich panels having one facing of either of two orthotropic materials, the other facing of an isotropic material; both facings of orthotropic material; both facings of...

Plasma discharge elemental detector for a mass spectrometer

NASA Astrophysics Data System (ADS)

Heppner, R. A.

1983-06-01

A material to be analyzed is injected into a mirowave-induced plasma discharge unit, in which the material is carried with a flow of buffer gas through an intense microwave energy field which produces a plasma discharge in the buffer gas. As the material exits from the plasma discharge, the material is sampled and conveyed along a capillary transfer tube to a mass spectrometer where it is analyzed. The plasma discharge causes dissociation of complex organic molecules into simpler molecules which return to the neutral ground state before they are analyzed in the mass spectrometer. The buffer gas is supplied to one end portion of the discharge tube and is withdrawn from the other end portion by a vacuum pump which maintains a subatmospheric pressure in the discharge tube. The sample material is injected by a capillary injection tube into the buffer gas flow as it enters the plasma discharge zone. The dissociated materials are sampled by an axial sampling tube having an entrance where the buffer gas exits from the plasma discharge zone. The sample material may be supplied by a gas chromatography having a capillary effluent line connected to the capillary injection tube, so that the effluent material is injected into the microwave induced plasma discharge. The microwave field is produced by a cavity resonator through which the discharge tube passes.

Present status of liquid metal research for a fusion reactor

NASA Astrophysics Data System (ADS)

Tabarés, Francisco L.

2016-01-01

Although the use of solid materials as targets of divertor plasmas in magnetic fusion research is accepted as the standard solution for the very challenging issue of power and particle handling in a fusion reactor, a generalized feeling that the present options chosen for ITER will not represent the best choice for a reactor is growing up. The problems found for tungsten, the present selection for the divertor target of ITER, in laboratory tests and in hot plasma fusion devices suggest so. Even in the absence of the strong neutron irradiation expected in a reactor, issues like surface melting, droplet ejection, surface cracking, dust generation, etc., call for alternative solutions in a long pulse, high efficient fusion energy-producing continuous machine. Fortunately enough, decades of research on plasma facing materials based on liquid metals (LMs) have produced a wealth of appealing ideas that could find practical application in the route to the realization of a commercial fusion power plant. The options presently available, although in a different degree of maturity, range from full coverage of the inner wall of the device with liquid metals, so that power and particle exhaust together with neutron shielding could be provided, to more conservative combinations of liquid metal films and conventional solid targets basically representing a sort of high performance, evaporative coating for the alleviation of the surface degradation issues found so far. In this work, an updated review of worldwide activities on LM research is presented, together with some open issues still remaining and some proposals based on simple physical considerations leading to the optimization of the most conservative alternatives.

The efficacy of post porosity plasma protection against vacuum-ultraviolet damage in porous low-k materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lionti, K.; Volksen, W.; Darnon, M.

2015-03-21

As of today, plasma damage remains as one of the main challenges to the reliable integration of porous low-k materials into microelectronic devices at the most aggressive node. One promising strategy to limit damage of porous low-k materials during plasma processing is an approach we refer to as post porosity plasma protection (P4). In this approach, the pores of the low-k material are filled with a sacrificial agent prior to any plasma treatment, greatly minimizing the total damage by limiting the physical interactions between plasma species and the low-k material. Interestingly, the contribution of the individual plasma species to themore » total plasma damage is not fully understood. In this study, we investigated the specific damaging effect of vacuum-ultraviolet (v-UV) photons on a highly porous, k = 2.0 low-k material and we assessed the P4 protective effect against them. It was found that the impact of the v-UV radiation varied depending upon the v-UV emission lines of the plasma. More importantly, we successfully demonstrated that the P4 process provides excellent protection against v-UV damage.« less

Onlay bone augmentation on mouse calvarial bone using a hydroxyapatite/collagen composite material with total blood or platelet-rich plasma.

PubMed

Ohba, Seigo; Sumita, Yoshinori; Umebayashi, Mayumi; Yoshimura, Hitoshi; Yoshida, Hisato; Matsuda, Shinpei; Kimura, Hideki; Asahina, Izumi; Sano, Kazuo

2016-01-01

The aim of this study was to assess newly formed onlay bone on mouse calvarial bone using a new artificial bone material, a hydroxyapatite/collagen composite, with total blood or platelet-rich plasma. The hydroxyapatite/collagen composite material with normal saline, total blood or platelet-rich plasma was transplanted on mouse calvarial bone. The mice were sacrificed and the specimens were harvested four weeks after surgery. The newly formed bone area was measured on hematoxylin and eosin stained specimens using Image J software. The hydroxyapatite/collagen composite materials with total blood or platelet-rich plasma induced a significantly greater amount of newly formed bone than that with normal saline. Moreover, bone marrow was observed four weeks after surgery in the transplanted materials with total blood or platelet-rich plasma but not with normal saline. However, there were no significant differences in the amount of newly formed bone between materials used with total blood versus platelet-rich plasma. The hydroxyapatite/collagen composite material was valid for onlay bone augmentation and this material should be soaked in total blood or platelet-rich plasma prior to transplantation. Copyright © 2015 Elsevier Ltd. All rights reserved.

Use of Atmospheric-Pressure Plasma Jet for Polymer Surface Modification: An Overview

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuettner, Lindsey A.

Atmospheric-pressure plasma jets (APPJs) are playing an increasingly important role in materials processing procedures. Plasma treatment is a useful tool to modify surface properties of materials, especially polymers. Plasma reacts with polymer surfaces in numerous ways thus the type of process gas and plasma conditions must be explored for chosen substrates and materials to maximize desired properties. This report discusses plasma treatments and looks further into atmospheric-pressure plasma jets and the effects of gases and plasma conditions. Following the short literature review, a general overview of the future work and research at Los Alamos National Laboratory (LANL) is discussed.

A Multi-ring Ionospheric Plasma Probe

NASA Technical Reports Server (NTRS)

Sheldon, J. W.

1972-01-01

An ionospheric plasma probe was constructed which consists of a long cylinder with the end facing the flow closed by an end plate made up of multiple annular rings and a center disk. A theoretical argument is given which yields the plasma potential and electron temperature in terms of known plasma parameters and the currents to the various rings of the end plate. This probe was successfully operated in an ionospheric flow simulation facility and the resulting plasma potential is in excellent agreement with the traditional Langmuir analysis (1.22 volts).

Process for forming exoergic structures with the use of a plasma

DOEpatents

Kelly, Michael D.

1989-02-21

A method of forming exoergic structures, as well as exoergic structures produced by the method, is provided. The method comprises the steps of passing a plasma-forming gas through a plasma spray gun, forming a plasma spray, introducing exoergic material into the plasma spray and directing the plasma spray toward a substrate, and allowing the exoergic material to become molten, without chemically reacting in the plasma spray and to thereafter impinge on the substrate to form a solid mass of exoergic material, the shape of which corresponds to the shape of the substrate.

Acute tryptophan depletion attenuates conscious appraisal of social emotional signals in healthy female volunteers

PubMed Central

Gray, Marcus A.; Minati, Ludovico; Whale, Richard; Harrison, Neil A.; Critchley, Hugo D.

2010-01-01

Rationale Acute tryptophan depletion (ATD) decreases levels of central serotonin. ATD thus enables the cognitive effects of serotonin to be studied, with implications for the understanding of psychiatric conditions, including depression. Objective To determine the role of serotonin in conscious (explicit) and unconscious/incidental processing of emotional information. Materials and methods A randomized, double-blind, cross-over design was used with 15 healthy female participants. Subjective mood was recorded at baseline and after 4 h, when participants performed an explicit emotional face processing task, and a task eliciting unconscious processing of emotionally aversive and neutral images presented subliminally using backward masking. Results ATD was associated with a robust reduction in plasma tryptophan at 4 h but had no effect on mood or autonomic physiology. ATD was associated with significantly lower attractiveness ratings for happy faces and attenuation of intensity/arousal ratings of angry faces. ATD also reduced overall reaction times on the unconscious perception task, but there was no interaction with emotional content of masked stimuli. ATD did not affect breakthrough perception (accuracy in identification) of masked images. Conclusions ATD attenuates the attractiveness of positive faces and the negative intensity of threatening faces, suggesting that serotonin contributes specifically to the appraisal of the social salience of both positive and negative salient social emotional cues. We found no evidence that serotonin affects unconscious processing of negative emotional stimuli. These novel findings implicate serotonin in conscious aspects of active social and behavioural engagement and extend knowledge regarding the effects of ATD on emotional perception. PMID:20596858

Plasma Sterilization Technology for Spacecraft Applications

NASA Technical Reports Server (NTRS)

Fraser, S. J.; Olson, R. L.; Leavens, W. M.

1975-01-01

The application of plasma gas technology to sterilization and decontamination of spacecraft components is considered. Areas investigated include: effective sterilizing ranges of four separate gases; lethal constituents of a plasma environment; effectiveness of plasma against a diverse group of microorganisms; penetrating efficiency of plasmas for sterilization; and compatibility of spacecraft materials with plasma environments. Results demonstrated that plasma gas, specifically helium plasma, is a highly effective sterilant and is compatible with spacecraft materials.

Innovative potential of plasma technology

NASA Astrophysics Data System (ADS)

Budaev, V. P.

2017-11-01

The review summarizes recent experimental observations of materials exposed to extreme hot plasma loads in fusion devices and plasma facilities with high-temperature plasma. Plasma load on the material in such devices lead to the stochastic clustering and fractal growth of the surface on scales from tens of nanometers to hundreds of micrometers forming statistical self-similarity of the surface roughness with extremely high specific area. Statistical characteristics of hierarchical granularity and scale invariance of such materials surface qualitatively differ from the properties of the roughness of the ordinary Brownian surface which provides a potential of innovative plasma technologies for synthesis of new nanostructured materials with programmed roughness properties, for hypersonic technologies, for biotechnology and biomedical applications.

Estimates of RF-induced erosion at antenna-connected beryllium plasma-facing components in JET

DOE PAGES

Klepper, C. C.; Borodin, D.; Groth, M.; ...

2016-01-18

Radio-frequency (RF)-enhanced surface erosion of beryllium (Be) plasma-facing components is explored, for the first time, using the ERO code. We applied the code in order to measure the RF-enhanced edge Be line emission at JET Be outboard limiters, in the presence of high-power, ion cyclotronresonance heating (ICRH) in L-mode discharges. In this first modelling study, the RF sheath effect from an ICRH antenna on a magnetically connected, limiter region is simulated by adding a constant potential to the local sheath, in an attempt to match measured increases in local Be I and Be II emission of factors of 2 3.more » It was found that such increases are readily simulated with added potentials in the range of 100 200 V, which is compatible with expected values for potentials arising from rectification of sheath voltage oscillations from ICRH antennas in the scrape-off layer plasma. We also estimated absolute erosion values within the uncertainties in local plasma conditions.« less

Bio-inert interfaces via biomimetic anchoring of a zwitterionic copolymer on versatile substrates.

PubMed

Dizon, Gian Vincent; Chou, Ying-Nien; Yeh, Lu-Chen; Venault, Antoine; Huang, James; Chang, Yung

2018-05-22

Bio-inert biomaterial design is vital for fields like biosensors, medical implants, and drug delivery systems. Bio-inert materials are generally hydrophilic and electrical neutral. One limitation faced in the design of bio-inert materials is that most of the modifiers used are specific to their substrate. In this work, we synthesized a novel zwitterionic copolymer containing a catechol group, a non-substrate dependent biomimetic anchoring segment, that can form a stable coating on various materials. No previous study was conducted using a grafting-to approach and determined the critical amount of catechol groups needed to effectively modify a material. The synthesized copolymers of sulfobetaine acrylamide (SBAA) and dopamine methacrylamide (DMA) in this work contains varying numbers of catechol groups, in which the critical number of catechol groups that had effectively modified substrates to have the bio-inert property was determined. The bio-inert property and capability to do coating on versatile substrates were evaluated in contact with human blood by coating different material groups such as ceramic, metallic, and polymeric groups. The novel structure and the simple grafting-to approach provides bio-inert property on various materials, giving them non-specific adsorption and attachment of biomolecules such as plasma proteins, erythrocytes, thrombocytes, bacteria, and tissue cells (85-95% reduction). Copyright © 2018 Elsevier Inc. All rights reserved.

Plasma of argon enhances the adhesion of murine osteoblasts on different graft materials.

PubMed

Canullo, Luigi; Genova, Tullio; Naenni, Nadja; Nakajima, Yasushi; Masuda, Katsuhiko; Mussano, Federico

2018-04-25

plasma of argon treatment was demonstrated to increase material surface energy leading to stronger and faster interaction with cells. The aim of the present in vitro study was to test the effect of plasma treatment on different graft materials. synthetic hydroxyapatite (Mg-HA), biphasic calcium phosphate (BCP), cancellous and cortical xenogeneic bone matrices (CaBM, CoBM) were used representing commonly used classes of bone substitute materials. Fifty serially numbered disks with a 10mm-diameter from each graft material were randomly divided into two groups: Test group (argon plasma treatment) and Control group (absence of treatment). Cell morphology (using pre-osteoblastic murine cells) and protein adsorption were analyzed at all samples from both the test and control group. Differences between groups were analyzed using the Mann-Whitney test setting the level of significance at p<0.05. plasma treatment significantly increased the protein adsorption at all samples. Similarly, plasma treatment significantly increased cell adhesion in all groups. data confirmed that non-atmospheric plasma of argon treatment led to an increase of protein adsorption and cell adhesion in all groups of graft material to a similar extent. plasma of argon is able to improve the surface conditions of graft materials. Copyright © 2018 Elsevier GmbH. All rights reserved.

Efficacy of autologous platelet-rich plasma combined with fractional ablative carbon dioxide resurfacing laser in treatment of facial atrophic acne scars: A split-face randomized clinical trial.

PubMed

Faghihi, Gita; Keyvan, Shima; Asilian, Ali; Nouraei, Saeid; Behfar, Shadi; Nilforoushzadeh, Mohamad Ali

2016-01-01

Autologous platelet-rich plasma has recently attracted significant attention throughout the medical field for its wound-healing ability. This study was conducted to investigate the potential of platelet-rich plasma combined with fractional laser therapy in the treatment of acne scarring. Sixteen patients (12 women and 4 men) who underwent split-face therapy were analyzed in this study. They received ablative fractional carbon dioxide laser combined with intradermal platelet-rich plasma treatment on one half of their face and ablative fractional carbon dioxide laser with intradermal normal saline on the other half. The injections were administered immediately after laser therapy. The treatment sessions were repeated after an interval of one month. The clinical response was assessed based on patient satisfaction and the objective evaluation of serial photographs by two blinded dermatologists at baseline, 1 month after the first treatment session and 4 months after the second. The adverse effects including erythema and edema were scored by participants on days 0, 2, 4, 6, 8, 15 and 30 after each session. Overall clinical improvement of acne scars was higher on the platelet-rich plasma-fractional carbon dioxide laser treated side but the difference was not statistically significant either 1 month after the first treatment session (P = 0.15) or 4 months after the second (P = 0.23). In addition, adverse effects (erythema and edema) on the platelet-rich plasma-fractional carbon dioxide laser-treated side were more severe and of longer duration. Small sample size, absence of all skin phototypes within the study group and lack of objective methods for the evaluation of response to treatment and adverse effects were the limitations. This study demonstrated that adding platelet-rich plasma to fractional carbon dioxide laser treatment did not produce any statistically significant synergistic effects and also resulted in more severe side effects and longer downtime.

Development of 2D imaging of SXR plasma radiation by means of GEM detectors

NASA Astrophysics Data System (ADS)

Chernyshova, M.; Czarski, T.; Jabłoński, S.; Kowalska-Strzeciwilk, E.; Poźniak, K.; Kasprowicz, G.; Zabołotny, W.; Wojeński, A.; Byszuk, A.; Burza, M.; Juszczyk, B.; Zienkiewicz, P.

2014-11-01

Presented 2D gaseous detector system has been developed and designed to provide energy resolved fast dynamic plasma radiation imaging in the soft X-Ray region with 0.1 kHz exposure frequency for online, made in real time, data acquisition (DAQ) mode. The detection structure is based on triple Gas Electron Multiplier (GEM) amplification structure followed by the pixel readout electrode. The efficiency of detecting unit was adjusted for the radiation energy region of tungsten in high-temperature plasma, the main candidate for the plasma facing material for future thermonuclear reactors. Here we present preliminary laboratory results and detector parameters obtained for the developed system. The operational characteristics and conditions of the detector were designed to work in the X-Ray range of 2-17 keV. The detector linearity was checked using the fluorescence lines of different elements and was found to be sufficient for good photon energy reconstruction. Images of two sources through various screens were performed with an X-Ray laboratory source and 55Fe source showing a good imaging capability. Finally offline stream-handling data acquisition mode has been developed for the detecting system with timing down to the ADC sampling frequency rate (~13 ns), up to 2.5 MHz of exposure frequency, which could pave the way to invaluable physics information about plasma dynamics due to very good time resolving ability. Here we present results of studied spatial resolution and imaging properties of the detector for conditions of laboratory moderate counting rates and high gain.

Reactor plasma facing component designs based on liquid metal concepts supported in porous systems

NASA Astrophysics Data System (ADS)

Tabarés, F. L.; Oyarzabal, E.; Martin-Rojo, A. B.; Tafalla, D.; de Castro, A.; Soleto, A.

2017-01-01

The use of liquid metals (LMs) as plasma facing components in fusion devices was proposed as early as 1970 for a field reversed concept and inertial fusion reactors. The idea was extensively developed during the APEX Project, at the turn of the century, and it is the subject at present of the biennial International Symposium on Lithium Applications (ISLA), whose fourth meeting took place in Granada, Spain at the end of September 2015. While liquid metal flowing concepts were specially addressed in USA research projects, the idea of embedding the metal in a capillary porous system (CPS) was put forwards by Russian teams in the 1990s, thus opening the possibility of static concepts. Since then, many ideas and accompanying experimental tests in fusion devices and laboratories have been produced, involving a large fraction of countries within the international fusion community. Within the EUROFusion Roadmap, these activities are encompassed into the working programs of the plasma facing components (PFC) and divertor tokamak test (DTT) packages. In this paper, a review of the state of the art in concepts based on the CPS set-up for a fusion reactor divertor target, aimed at preventing the ejection of the liquid metal by electro-magnetic (EM) forces generated under plasma operation, is described and required R+D activities on the topic, including ongoing work at CIEMAT specifically oriented to filling the remaining gaps, are stressed.

Plasma Processing of Metallic and Semiconductor Thin Films in the Fisk Plasma Source

NASA Technical Reports Server (NTRS)

Lampkin, Gregory; Thomas, Edward, Jr.; Watson, Michael; Wallace, Kent; Chen, Henry; Burger, Arnold

1998-01-01

The use of plasmas to process materials has become widespread throughout the semiconductor industry. Plasmas are used to modify the morphology and chemistry of surfaces. We report on initial plasma processing experiments using the Fisk Plasma Source. Metallic and semiconductor thin films deposited on a silicon substrate have been exposed to argon plasmas. Results of microscopy and chemical analyses of processed materials are presented.

Assembly & Metrology of First Wall Components of SST-1

NASA Astrophysics Data System (ADS)

Parekh, Tejas; Santra, Prosenjit; Biswas, Prabal; Patel, Hiteshkumar; Paravastu, Yuvakiran; Jaiswal, Snehal; Chauhan, Pradeep; Babu, Gattu Ramesh; A, Arun Prakash; Bhavsar, Dhaval; Raval, Dilip C.; Khan, Ziauddin; Pradhan, Subrata

2017-04-01

First Wall Components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma comprises of limiters, divertors, baffles, passive stabilizers are designed to operate long duration (1000 s) discharges of elongated plasma. All FWC consists of a copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at ring & port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 undergoing a meticulous planning of assembly sequence, quality checks at every stage of the assembly process. This paper will present the metrology aspects & procedure of each FWC, both outside the vacuum vessel, and inside the vessel, assembly tolerances, tools, equipment and jig/fixtures, used at each stage of assembly, starting from location of support bases on vessel rings, fixing of copper modules on support structures, around 3800 graphite tile mounting on 136 copper modules with proper tightening torques, till final toroidal and poloidal geometry of the in-vessel components are obtained within acceptable limits, also ensuring electrical continuity of passive stabilizers to form a closed saddle loop, electrical isolation of passive stabilizers from vacuum vessel.

Method of processing materials using an inductively coupled plasma

DOEpatents

Hull, Donald E.; Bieniewski, Thomas M.

1989-01-01

A method for coating surfaces or implanting ions in an object using an inductively coupled plasma. The method provides a gas-free environment, since the plasma is formed without using a gas. The coating material or implantation material is intitially in solid form.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morgan, O.B. Jr.; Berry, L.A.; Sheffield, J.

This annual report on fusion energy discusses the progress on work in the following main topics: toroidal confinement experiments; atomic physics and plasma diagnostics development; plasma theory and computing; plasma-materials interactions; plasma technology; superconducting magnet development; fusion engineering design center; materials research and development; and neutron transport. (LSP)

Carbon materials modified by plasma treatment as electrodes for supercapacitors

NASA Astrophysics Data System (ADS)

Lota, Grzegorz; Tyczkowski, Jacek; Kapica, Ryszard; Lota, Katarzyna; Frackowiak, Elzbieta

The carbon material was modified by RF plasma with various reactive gases: O 2, Ar and CO 2. Physicochemical properties of the final carbon products were characterized using different techniques such as gas adsorption method and XPS. Plasma modified materials enriched in oxygen functionalities were investigated as electrodes for supercapacitors in acidic medium. The electrochemical measurements have been carried out using cyclic voltammetry, galvanostatic charge/discharge and impedance spectroscopy. The electrochemical measurements have confirmed that capacity characteristics are closely connected with a type of plasma exposition. Modification processes have an influence on the kind and amount of surface functional groups in the carbon matrix. The moderate increase of capacity of carbon materials modified by plasma has been observed using symmetric two-electrode systems. Whereas investigations made in three-electrode system proved that the suitable selection of plasma modification parameters allows to obtain promising negative and positive electrode materials for supercapacitor application.

Development of barrier coatings for cellulosic-based materials by cold plasma methods

NASA Astrophysics Data System (ADS)

Denes, Agnes Reka

Cellulose-based materials are ideal candidates for future industries that need to be based on environmentally safe technologies and renewable resources. Wood represents an important raw material and its application as construction material is well established. Cellophane is one of the most important cellulosic material and it is widely used as packaging material in the food industry. Outdoor exposure of wood causes a combination of physical and chemical degradation processes due to the combined effects of sunlight, moisture, fungi, and bacteria. Cold-plasma-induced surface modifications are an attractive way for tailoring the characteristics of lignocellulosic substrates to prevent weathering degradation. Plasma-polymerized hexamethyldisiloxane (PPHMDSO) was deposited onto wood surfaces to create water repellent characteristics. The presence of a crosslinked macromolecular structure was detected. The plasma coated samples exhibited very high water contact angle values indicating the existence of hydrophobic surfaces. Reflective and electromagnetic radiation-absorbent substances were incorporated with a high-molecular-weight polydimethylsiloxane polymer in liquid phase and deposited as thin layers on wood surfaces. The macromolecular films, containing the dispersed materials, were then converted into a three dimensional solid state network by exposure to a oxygen-plasma. It was demonstrated that both UV-absorbent and reflectant components incorporated into the plasma-generated PDMSO matrix protected the wood from weathering degradation. Reduced oxidation and less degradation was observed after simulated weathering. High water contact angle values indicated a strong hydrophobic character of the oxygen plasma-treated PDMSO-coated samples. Plasma-enhanced surface modifications and coatings were employed to create water-vapor barrier layers on cellophane substrate surfaces. HMDSO was selected as a plasma gas and oxygen was used to ablate amorphous regions. Oxygen plasma treated cellophane and oxygen plasma treated and PPHMDSO coated cellophane surfaces were comparatively analyzed and the corresponding surface wettability characteristics were evaluated. The plasma generated surface topographies controlled the morphology of the PPHMDSO layers. Higher temperature HMDSO plasma-state environments lead to insoluble, crosslinked layers. Continuous and pulsed Csb2Fsb6 plasmas were also used for surface modification and excellent surface fluorination was achieved under the pulsed plasma conditions.

X-ray And EUV Spectroscopy Of Highly Charged Tungsten Ions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Biedermann, Christoph; Radtke, Rainer

2009-09-10

The Berlin EBIT has been established by the Max-Planck-Institut fuer Plasmaphysik to generate atomic physics data in support of research in the field of controlled nuclear fusion, by measuring the radiation from highly charged ions in the x-ray, extreme ultraviolet and visible spectral ranges and providing valuable diagnostics for high temperature plasmas. In future fusion devices, for example ITER, currently being constructed at Cadarache, France, the plasma facing components will be armored with high-Z materials, most likely tungsten, due to the favorable properties of this element. At the same time the tremendous radiation cooling of these high-Z materials represents amore » threat to fusion and obliges one to monitor carefully the radiation. With EBIT a selected ensemble of ions in specific charge states can be produced, stored and excited for spectroscopic investigations. Employing this technique, we have for example resolved the wide structure observed around 5 nm at the ASDEX Upgrade tokamak as originating from E1-transitions into the open 4d shell of tungsten ions in charge states 25+ to 37+ producing a band-like emission pattern. Further, these ions emit well-separated M1 lines in the EUV range around 65 nm suitable for plasma diagnostics. Kr-like to Cr-like tungsten ions (38+ to 50+) show strong soft-x-ray lines in the range 0.5 to 2 and 5 to 15 nm. Lines of even higher charged tungsten ions, up to Ne-like W{sup 64+}, abundant in the core plasma of present and future fusion test devices, have been investigated with high resolution Bragg-crystal spectroscopy at 0.13 nm. Recently, x-ray spectroscopic measurements of the dielectronic recombination LMn resonances of W{sup 60+} to W{sup 67+} ions have been preformed and compare well with atomic structure calculations.« less

Microstructure and hardness evolution of nanochannel W films irradiated by helium at high temperature

NASA Astrophysics Data System (ADS)

Qin, Wenjing; Wang, Yongqiang; Tang, Ming; Ren, Feng; Fu, Qiang; Cai, Guangxu; Dong, Lan; Hu, Lulu; Wei, Guo; Jiang, Changzhong

2018-04-01

Plasma facing materials (PFMs) face one of the most serious challenges in fusion reactors, including unprecedented harsh environment such as 14.1 MeV neutron and transmutation gas irradiation at high temperature. Tungsten (W) is considered to be one of the most promising PFM, however, virtually insolubility of helium (He) in W causes new material issues such as He bubbles and W "fuzz" microstructure. In our previous studies, we presented a new strategy using nanochannel structure designed in the W film to increase the releasing of He atoms and thus to minimize the He nucleation and "fuzz" formation behavior. In this work, we report the further study on the diffusion of He atoms in the nanochannel W films irradiated at a high temperature of 600 °C. More specifically, the temperature influences on the formation and growth of He bubbles, the lattice swelling, and the mechanical properties of the nanochannel W films were investigated. Compared with the bulk W, the nanochannel W films possessed smaller bubble size and lower bubble areal density, indicating that noticeable amounts of He atoms have been released out along the nanochannels during the high temperature irradiations. Thus, with lower He concentration in the nanochannel W films, the formation of the bubble superlattice is delayed, which suppresses the lattice swelling and reduces hardening. These aspects indicate the nanochannel W films have better radiation resistance even at high temperature irradiations.

Faces are special but not too special: Spared face recognition in amnesia is based on familiarity

PubMed Central

Aly, Mariam; Knight, Robert T.; Yonelinas, Andrew P.

2014-01-01

Most current theories of human memory are material-general in the sense that they assume that the medial temporal lobe (MTL) is important for retrieving the details of prior events, regardless of the specific type of materials. Recent studies of amnesia have challenged the material-general assumption by suggesting that the MTL may be necessary for remembering words, but is not involved in remembering faces. We examined recognition memory for faces and words in a group of amnesic patients, which included hypoxic patients and patients with extensive left or right MTL lesions. Recognition confidence judgments were used to plot receiver operating characteristics (ROCs) in order to more fully quantify recognition performance and to estimate the contributions of recollection and familiarity. Consistent with the extant literature, an analysis of overall recognition accuracy showed that the patients were impaired at word memory but had spared face memory. However, the ROC analysis indicated that the patients were generally impaired at high confidence recognition responses for faces and words, and they exhibited significant recollection impairments for both types of materials. Familiarity for faces was preserved in all patients, but extensive left MTL damage impaired familiarity for words. These results suggest that face recognition may appear to be spared because performance tends to rely heavily on familiarity, a process that is relatively well preserved in amnesia. The findings challenge material-general theories of memory, and suggest that both material and process are important determinants of memory performance in amnesia, and different types of materials may depend more or less on recollection and familiarity. PMID:20833190

Plasma Spraying of Ceramics with Particular Difficulties in Processing

NASA Astrophysics Data System (ADS)

Mauer, G.; Schlegel, N.; Guignard, A.; Jarligo, M. O.; Rezanka, S.; Hospach, A.; Vaßen, R.

2015-01-01

Emerging new applications and growing demands of plasma-sprayed coatings initiate the development of new materials. Regarding ceramics, often complex compositions are employed to achieve advanced material properties, e.g., high thermal stability, low thermal conductivity, high electronic and ionic conductivity as well as specific thermo-mechanical properties and microstructures. Such materials however, often involve particular difficulties in processing by plasma spraying. The inhomogeneous dissociation and evaporation behavior of individual constituents can lead to changes of the chemical composition and the formation of secondary phases in the deposited coatings. Hence, undesired effects on the coating characteristics are encountered. In this work, examples of such challenging materials are investigated, namely pyrochlores applied for thermal barrier coatings as well as perovskites for gas separation membranes. In particular, new plasma spray processes like suspension plasma spraying and plasma spray-physical vapor deposition are considered. In some cases, plasma diagnostics are applied to analyze the processing conditions.

Secondary electron emission from lithium and lithium compounds

DOE PAGES

Capece, A. M.; Patino, M. I.; Raitses, Y.; ...

2016-07-06

In this work, measurements of electron-induced secondary electron emission ( SEE) yields of lithium as a function of composition are presented. The results are particularly relevant for magnetic fusion devices such as tokamaks, field-reversed configurations, and stellarators that consider Li as a plasma-facing material for improved plasma confinement. SEE can reduce the sheath potential at the wall and cool electrons at the plasma edge, resulting in large power losses. These effects become significant as the SEE coefficient, γ e, approaches one, making it imperative to maintain a low yield surface. This work demonstrates that the yield from Li strongly dependsmore » on chemical composition and substantially increases after exposure to oxygen and water vapor. The total yield was measured using a retarding field analyzer in ultrahigh vacuum for primary electron energies of 20-600 eV. The effect of Li composition was determined by introducing controlled amounts of O 2 and H 2O vapor while monitoring film composition with Auger electron spectroscopy and temperature programmed desorption. The results show that the energy at which γ e = 1 decreases with oxygen content and is 145 eV for a Li film that is 17% oxidized and drops to less than 25 eV for a fully oxidized film. This work has important implications for laboratory plasmas operating under realistic vacuum conditions in which oxidation significantly alters the electron emission properties of Li walls. Published by AIP Publishing.« less

Secondary electron emission from lithium and lithium compounds

DOE Office of Scientific and Technical Information (OSTI.GOV)

Capece, A. M., E-mail: capecea@tcnj.edu; Department of Physics, The College of New Jersey, Ewing, New Jersey 08628; Patino, M. I.

2016-07-04

In this work, measurements of electron-induced secondary electron emission (SEE) yields of lithium as a function of composition are presented. The results are particularly relevant for magnetic fusion devices such as tokamaks, field-reversed configurations, and stellarators that consider Li as a plasma-facing material for improved plasma confinement. SEE can reduce the sheath potential at the wall and cool electrons at the plasma edge, resulting in large power losses. These effects become significant as the SEE coefficient, γ{sub e}, approaches one, making it imperative to maintain a low yield surface. This work demonstrates that the yield from Li strongly depends onmore » chemical composition and substantially increases after exposure to oxygen and water vapor. The total yield was measured using a retarding field analyzer in ultrahigh vacuum for primary electron energies of 20–600 eV. The effect of Li composition was determined by introducing controlled amounts of O{sub 2} and H{sub 2}O vapor while monitoring film composition with Auger electron spectroscopy and temperature programmed desorption. The results show that the energy at which γ{sub e} = 1 decreases with oxygen content and is 145 eV for a Li film that is 17% oxidized and drops to less than 25 eV for a fully oxidized film. This work has important implications for laboratory plasmas operating under realistic vacuum conditions in which oxidation significantly alters the electron emission properties of Li walls.« less

Ohmic ignition with high engineering beta based on the RFP

NASA Astrophysics Data System (ADS)

Sarff, J. S.; Anderson, J. K.; Chapman, B. E.; McCollam, K. J.

2017-10-01

The RFP configuration allows the possibility of ohmic ignition for fusion energy, eliminating the need for auxiliary heating by rf or neutral beam injection. Complex plasma-facing antennas and NBI sources are therefore not required, simplifying the difficult fusion materials challenge. While all toroidal configurations require a volume-average 〈 B 〉 >= 5 T, the field strength at the magnet in the RFP is only Bcoil 3T since plasma current generates almost all of the field. Engineering beta is therefore maximized. We summarize access to ohmic ignition by examining a Lawson-like power balance for an RFP fusion plasma comparable to the ARIES-AT advanced tokamak, which generates neutron wall loading Pn / A 5 MW/m2. The required energy confinement for ohmic ignition in an RFP is similar to that for a tokamak. Confinement in MST is comparable to a same-size, same-field tokamak plasma, but 〈 B 〉 in MST is only 1/20th that required for fusion. While transport could ultimately be dominated by micro turbulence, extrapolation of stochastic transport using Lundquist number scaling for MHD tearing indicates standard RFP confinement (not enhanced by current profile control) could be sufficient to access ohmic ignition. This bolsters the possibility for steady-state inductive sustainment using oscillating field current drive. The high beta and classical energetic ion confinement measured in MST also bolster the RFP's fusion potential. Work supported by U.S. DoE.

Thermomechanical and chemical properties of porous W/liquid Li hybrid systems as plasma-facing self-healing surfaces

NASA Astrophysics Data System (ADS)

Kapat, Aveek; Lang, Eric; Neff, Anton; Allain, Jean Paul

2017-10-01

The environmental conditions at the plasma-material interface of a future nuclear fusion reactor interacting will be extreme. The incident plasma will carry heat fluxes of the order of 100's of MWm-2 and particle fluxes that can average 1024 m-2s-1. The fusion reactor wall would need to operate at high temperatures near 800 C and the incident energy of particles will vary from a few eV ions to MeV neutrons. A hybrid system, inspired by self-healing solid-state concepts, combines the ductile phase of liquid Li within a solid phase porous W. The liquid Li serves to control hydrogen retention and provide vapor shielding, within the framework of a tunable porosity to optimize edge plasma conditions [2]. Additionally, the porous interface can also provide for effective defect sinks for high duty cycle neutron damage. The surface chemistry of liquid Li on a porous surface varied with D irradiation is studied and its effect on retention. Prior results with refractory alloys have demonstrated effective wetting properties [3]. These hybrid systems, as well as traditional W samples, are bombarded with 500eV D2+and Ar+ at 230oC and 300oC. The Li, O, and C XPS peaks were examined and compared to controls. Additionally, the porous W is characterized for thermo-mechanical properties. Work supported by USDOE Contract DE- DE-SC0014267.

Etude fondamentale des mecanismes de gravure par plasma de materiaux de pointe: Application a la fabrication de dispositifs photoniques

NASA Astrophysics Data System (ADS)

Stafford, Luc

Advances in electronics and photonics critically depend upon plasma-based materials processing either for transferring small lithographic patterns into underlying materials (plasma etching) or for the growth of high-quality films. This thesis deals with the etching mechanisms of materials using high-density plasmas. The general objective of this work is to provide an original framework for the plasma-material interaction involved in the etching of advanced materials by putting the emphasis on complex oxides such as SrTiO3, (Ba,Sr)TiO 3 and SrBi2Ta2O9 films. Based on a synthesis of the descriptions proposed by different authors to explain the etching characteristics of simple materials in noble and halogenated plasma mixtures, we propose comprehensive rate models for physical and chemical plasma etching processes. These models have been successfully validated using experimental data published in literature for Si, Pt, W, SiO2 and ZnO. As an example, we have been able to adequately describe the simultaneous dependence of the etch rate on ion and reactive neutral fluxes and on the ion energy. From an exhaustive experimental investigation of the plasma and etching properties, we have also demonstrated that the validity of the proposed models can be extended to complex oxides such as SrTiO3, (Ba,Sr)TiO 3 and SrBi2Ta2O9 films. We also reported for the first time physical aspects involved in plasma etching such as the influence of the film microstructural properties on the sputter-etch rate and the influence of the positive ion composition on the ion-assisted desorption dynamics. Finally, we have used our deep investigation of the etching mechanisms of STO films and the resulting excellent control of the etch rate to fabricate a ridge waveguide for photonic device applications. Keywords: plasma etching, sputtering, adsorption and desorption dynamics, high-density plasmas, plasma diagnostics, advanced materials, photonic applications.

Numerical band structure calculations of plasma metamaterials

NASA Astrophysics Data System (ADS)

Pederson, Dylan; Kourtzanidis, Konstantinos; Raja, Laxminarayan

2015-09-01

Metamaterials (MM) are materials engineered to display negative macroscopic permittivity and permeability. These materials allow for designed control over electromagnetic energy flow, especially at frequencies where natural materials do not interact. Plasmas have recently found application in MM as a negative permittivity component. The permittivity of a plasma depends on its electron density, which can be controlled by an applied field. This means that plasmas can be used in MM to actively control the transmission or reflection of incident waves. This work focuses on a plasma MM geometry in which microplasmas are generated in perforations in a metal plate. We characterizethis material by its band structure, which describes its interaction with incident waves. The plasma-EM interactions are obtained by coupling Maxwell's equations to a simplified plasma momentum equation. A plasma density profile is prescribed, and its effect on the band structure is investigated. The band structure calculations are typically done for static structures, whereas our current density responds to the incident waves. The resulting band structures are compared with experimental results.

Heat flux estimates of power balance on Proto-MPEX with IR imaging

DOE Office of Scientific and Technical Information (OSTI.GOV)

Showers, M., E-mail: mshower1@vols.utk.edu; Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831; Biewer, T. M.

The Prototype Material Plasma Exposure eXperiment (Proto-MPEX) at Oak Ridge National Laboratory (ORNL) is a precursor linear plasma device to the Material Plasma Exposure eXperiment (MPEX), which will study plasma material interactions (PMIs) for future fusion reactors. This paper will discuss the initial steps performed towards completing a power balance on Proto-MPEX to quantify where energy is lost from the plasma, including the relevant diagnostic package implemented. Machine operating parameters that will improve Proto-MPEX’s performance may be identified, increasing its PMI research capabilities.

Organic crystalline films for optical applications and related methods of fabrication

NASA Technical Reports Server (NTRS)

Leyderman, Alexander (Inventor)

2001-01-01

A method for forming an optical device includes the steps of providing a first plate having a first face defining a recess, filling the recess with a material which can be crystallized, and covering the first face and the recess with a second plate having a second face, so that the second face is in contact with the first face and the material in the recess is completely enclosed by the first and second plates. The material in the recess is thereby protected from chemical and mechanical damage, as well as evaporation. In addition, the plates can be transparent, allowing the material in the recess to be visually monitored. A grown crystalline film packed in the cell can be used as a non-liner and/or electro-optical device.

Elimination of columnar microstructure in N-face InAlN, lattice-matched to GaN, grown by plasma-assisted molecular beam epitaxy in the N-rich regime

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahmadi, Elaheh; Wienecke, Steven; Keller, Stacia

2014-02-17

The microstructure of N-face InAlN layers, lattice-matched to GaN, was investigated by scanning transmission electron microscopy and atom probe tomography. These layers were grown by plasma-assisted molecular beam epitaxy (PAMBE) in the N-rich regime. Microstructural analysis shows an absence of the lateral composition modulation that was previously observed in InAlN films grown by PAMBE. A room temperature two-dimensional electron gas (2DEG) mobility of 1100 cm{sup 2}/V s and 2DEG sheet charge density of 1.9 × 10{sup 13} cm{sup −2} was measured for N-face GaN/AlN/GaN/InAlN high-electron-mobility transistors with lattice-matched InAlN back barriers.

Multi-scale modeling to relate Be surface temperatures, concentrations and molecular sputtering yields

NASA Astrophysics Data System (ADS)

Lasa, Ane; Safi, Elnaz; Nordlund, Kai

2015-11-01

Recent experiments and Molecular Dynamics (MD) simulations show erosion rates of Be exposed to deuterium (D) plasma varying with surface temperature and the correlated D concentration. Little is understood how these three parameters relate for Be surfaces, despite being essential for reliable prediction of impurity transport and plasma facing material lifetime in current (JET) and future (ITER) devices. A multi-scale exercise is presented here to relate Be surface temperatures, concentrations and sputtering yields. Kinetic Monte Carlo (MC) code MMonCa is used to estimate equilibrium D concentrations in Be at different temperatures. Then, mixed Be-D surfaces - that correspond to the KMC profiles - are generated in MD, to calculate Be-D molecular erosion yields due to D irradiation. With this new database implemented in the 3D MC impurity transport code ERO, modeling scenarios studying wall erosion, such as RF-induced enhanced limiter erosion or main wall surface temperature scans run at JET, can be revisited with higher confidence. Work supported by U.S. DOE under Contract DE-AC05-00OR22725.

Stochastic clustering of material surface under high-heat plasma load

NASA Astrophysics Data System (ADS)

Budaev, Viacheslav P.

2017-11-01

The results of a study of a surface formed by high-temperature plasma loads on various materials such as tungsten, carbon and stainless steel are presented. High-temperature plasma irradiation leads to an inhomogeneous stochastic clustering of the surface with self-similar granularity - fractality on the scale from nanoscale to macroscales. Cauliflower-like structure of tungsten and carbon materials are formed under high heat plasma load in fusion devices. The statistical characteristics of hierarchical granularity and scale invariance are estimated. They differ qualitatively from the roughness of the ordinary Brownian surface, which is possibly due to the universal mechanisms of stochastic clustering of material surface under the influence of high-temperature plasma.

PREFACE: 26th Symposium on Plasma Science for Materials (SPSM-26)

NASA Astrophysics Data System (ADS)

2014-06-01

26th Symposium on Plasma Science for Materials (SPSM-26) Takayuki Watanabe The 26th Symposium on Plasma Science for Materials (SPSM-26) was held in Fukuoka, Japan on September 23-24, 2013. SPSM has been held annually since 1988 under the sponsorship of The 153rd Committee on Plasma Materials Science, Japan Society for the Promotion of Science (JSPS). This symposium is one of the major activities of the Committee, which is organized by researchers in academia and industry for the purpose of advancing intersectional scientific information exchange and discussion of science and technology of plasma materials processing. Plasma processing have attracted extensive attention due to their unique advantages, and it is expected to be utilized for a number of innovative industrial applications such as synthesis of high-quality and high-performance nanomaterials. The advantages of plasmas including high chemical reactivity in accordance with required chemical reactions are beneficial for innovative processing. In recent years, plasma materials processing with reactive plasmas has been extensively employed in the fields of environmental issues and biotechnology. This conference seeks to bring different scientific communities together to create a forum for discussing the latest developments and issues. The conference provides a platform for the exploration of both fundamental topics and new applications of plasmas by the contacts between science, technology, and industry. The conference was organized in plenary lectures, invited, contributed oral presentations, and poster sessions. At this meeting, we had 142 participants from 10 countries and 104 presentations, including 11 invited presentations. This year, we arranged special topical sessions that cover Plasma Medicine and Biotechnologies, Business and Academia Cooperation, Plasma with Liquids, Plasma Processes for Nanomaterials, together with Basic, Electronics, and Thermal Plasma sessions. This special issue presents 28 papers that are selected via strict peer-review process from full papers submitted for the proceedings of the conference. The topics range from basic physics and chemistry of plasma processing to a broad variety of materials processing and environmental applications. This proceeding offers an overview on the recent advances in thermal and non-equilibrium plasmas as well as the challenges ahead in the field of plasma research and applications among engineers and scientists. It is an honor to present this volume of Journal of Physics: Conference Series and we deeply thank the authors for their enthusiastic and high-grade contribution. The editors hope that this proceeding will be useful and helpful for deepening our understanding of science and technology of plasma materials processing and also for stimulating further development of the plasma technology. Finally, I would like to thank the organizing committee and organizing secretariat of SPSM-26, and the participants of the conference for contribution to a successful and exciting meeting. The conference was chaired by Prof. Masaharu Shiratani, Kyushu University. I would also like to thank the financial support from The 153rd Committee on Plasma Materials Science. Editors of SPMS-26 Prof Takayuki Watanabe, Kyushu University, Japan Prof Makoto Sekine, Nagoya University, Japan Prof Takanori Ichiki, The University of Tokyo, Japan Prof Masaharu Shiratani, Kyushu University, Japan Prof Akimitsu Hatta, Kochi University of Technology, Japan Sponsors and Supporting Organization: The 153rd Committee on Plasma Materials Science, Japan Society for the Promotion of Science

An assessment of surface emissivity variation effects on plasma uniformity analysis using IR cameras

NASA Astrophysics Data System (ADS)

Greenhalgh, Abigail; Showers, Melissa; Biewer, Theodore

2017-10-01

The Prototype-Material Plasma Exposure eXperiment (Proto-MPEX) is a linear plasma device operating at Oak Ridge National Laboratory (ORNL). Its purpose is to test plasma source and heating concepts for the planned Material Plasma Exposure eXperiment (MPEX), which has the mission to test the plasma-material interactions under fusion reactor conditions. In this device material targets will be exposed to high heat fluxes (>10 MW/m2). To characterize the heat fluxes to the target a IR thermography system is used taking up to 432 frames per second videos. The data is analyzed to determine the surface temperature on the target in specific regions of interest. The IR analysis has indicated a low level of plasma uniformity; the plasma often deposits more heat to the edge of the plate than the center. An essential parameter for IR temperature calculation is the surface emissivity of the plate (stainless steel). A study has been performed to characterize the variation in the surface emissivity of the plate as its temperature changes and its surface finish is modified by plasma exposure.

Spacecraft Charging Modeling -- Nascap-2k 2014 Annual Report

DTIC Science & Technology

2014-09-19

i ) ’ "’"’ 2rrm" T (2) For a surface directly facing the .ram at a typical low- Earth - orbit speed of 7.500 m/ s in a 0.1 eV plasma . the surface is...of modeling the charging of spacecraft with a low- Earth -orbit plasma within Nascap-2k. This work resulted in a paper presented at the Spacecraft...approaches used to model spacecraft charging in cold. dense plasma . such as found in low- Earth -orbit The range of plasma properties under

Experimental results of near real-time protection system for plasma facing components in Wendelstein 7-X at GLADIS

NASA Astrophysics Data System (ADS)

Ali, A.; Jakubowski, M.; Greuner, H.; Böswirth, B.; Moncada, V.; Sitjes, A. Puig; Neu, R.; Pedersen, T. S.; the W7-X Team

2017-12-01

One of the aims of stellarator Wendelstein 7-X (W7-X), is to investigate steady state operation, for which power exhaust is an important issue. The predominant fraction of the energy lost from the confined plasma region will be absorbed by an island divertors, which is designed for 10 {{MWm}}-2 steady state operation. In order to protect the divertor targets from overheating, 10 state-of-the-art infrared endoscopes will be installed at W7-X. In this work, we present the experimental results obtained at the high heat flux test facility GLADIS (Garching LArge DIvertor Sample test facility in IPP Garching) [1] during tests of a new plasma facing components (PFCs) protection algorithm designed for W7-X. The GLADIS device is equipped with two ion beams that can generate a heat load in the range from 3 MWm-2 to 55 MWm-2. The algorithms developed at W7-X to detect defects and hot spots are based on the analysis of surface temperature evolution and are adapted to work in near real-time. The aim of this work was to test the near real-time algorithms in conditions close to those expected in W7-X. The experiments were performed on W7-X pre-series tiles to detect CFC/Cu delaminations. For detection of surface layers, carbon fiber composite (CFC) blocks from the divertor of the Wendelstein 7-AS stellarator were used to observe temporal behavior of fully developed surface layers. These layers of re-deposited materials, like carbon, boron, oxygen and iron, were formed during the W7-AS operation. A detailed analysis of the composition and their thermal response to high heat fluxes (HHF) are described in [2]. The experiments indicate that the automatic detection of critical events works according to W7-X PFC protection requirements.

Power Balance Analysis of the Prototype-Material Plasma Exposure eXperiment

NASA Astrophysics Data System (ADS)

Showers, M. A.; Biewer, T. M.; Caneses, J. F.; Caughman, J. B. O.; Lumsdaine, A.; Owen, L.; Rapp, J.; Youchison, D.; Beers, C. J.; Donovan, D. C.; Kafle, N.; Ray, H. B.

2017-10-01

The Prototype-Material Plasma Exposure eXperiment (Proto-MPEX) is a test bed for the plasma source concept for the planned Material Plasma Exposure eXperiment (MPEX), a steady-state linear device studying plasma material interactions for fusion reactors. A power balance of Proto-MPEX attempts to identify machine operating parameters that will improve Proto-MPEX's performance, potentially impacting the MPEX design concept. A power balance has been performed utilizing an extensive diagnostic suite to identify mechanisms and locations of power loss from the main plasma. The diagnostic package includes infrared cameras, double Langmuir probes, fluoroptic probes, Mach probes, a Thomson scattering diagnostic, a McPherson spectrometer and in-vessel thermocouples. Radiation losses are estimated with absolute calibrated spectroscopic signals. This work was supported by the U.S. D.O.E. contract DE-AC05-00OR22725.

Properties of Vacancy Complexes with Hydrogen and Helium Atoms in Tungsten from First Principles

DOE PAGES

Samolyuk, German D.; Osetsky, Yury N.; Stoller, Roger E.

2016-12-03

Tungsten and its alloys are the primary candidate materials for plasma-facing components in fusion reactors. The material is exposed to high-energy neutrons and the high flux of helium and hydrogen atoms. In this paper, we have studied the properties of vacancy clusters and their interaction with H and He in W using density functional theory. Convergence of calculations with respect to modeling cell size was investigated. It is demonstrated that vacancy cluster formation energy converges with small cells with a size of 6 × 6 × 6 (432 lattice sites) enough to consider a microvoid of up to six vacanciesmore » with high accuracy. Most of the vacancy clusters containing fewer than six vacancies are unstable. Introducing He or H atoms increases their binding energy potentially making gas-filled bubbles stable. Finally, according to the results of the calculations, the H 2 molecule is unstable in clusters containing six or fewer vacancies.« less

Radioactivity evaluation for the KSTAR tokamak.

PubMed

Kim, Hyunduk; Lee, Hee-Seock; Hong, Sukmo; Kim, Minho; Chung, Chinwha; Kim, Changsuk

2005-01-01

The deuterium-deuterium (D-D) reaction in the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak generates neutrons with a peak yield of 2.5 x 10(16) s(-1) through a pulse operation of 300 s. Since the structure material of the tokamak is irradiated with neutrons, this environment will restrict work around and inside the tokamak from a radiation protection physics point of view after shutdown. Identification of neutron-produced radionuclides and evaluation of absorbed dose in the structure material are needed to develop a guiding principle for radiation protection. The activation level was evaluated by MCNP4C2 and an inventory code, FISPACT. The absorbed dose in the working area decreased by 4.26 x 10(-4) mrem h(-1) in the inner vessel 1.5 d after shutdown. Furthermore, tritium strongly contributes to the contamination in the graphite tile. The amount of tritium produced by neutrons was 3.03 x 10(6) Bq kg(-1) in the carbon graphite of a plasma-facing wall.

3D toroidal physics: testing the boundaries of symmetry breaking

NASA Astrophysics Data System (ADS)

Spong, Don

2014-10-01

Toroidal symmetry is an important concept for plasma confinement; it allows the existence of nested flux surface MHD equilibria and conserved invariants for particle motion. However, perfect symmetry is unachievable in realistic toroidal plasma devices. For example, tokamaks have toroidal ripple due to discrete field coils, optimized stellarators do not achieve exact quasi-symmetry, the plasma itself continually seeks lower energy states through helical 3D deformations, and reactors will likely have non-uniform distributions of ferritic steel near the plasma. Also, some level of designed-in 3D magnetic field structure is now anticipated for most concepts in order to lead to a stable, steady-state fusion reactor. Such planned 3D field structures can take many forms, ranging from tokamaks with weak 3D ELM-suppression fields to stellarators with more dominant 3D field structures. There is considerable interest in the development of unified physics models for the full range of 3D effects. Ultimately, the questions of how much symmetry breaking can be tolerated and how to optimize its design must be addressed for all fusion concepts. Fortunately, significant progress is underway in theory, computation and plasma diagnostics on many issues such as magnetic surface quality, plasma screening vs. amplification of 3D perturbations, 3D transport, influence on edge pedestal structures, MHD stability effects, modification of fast ion-driven instabilities, prediction of energetic particle heat loads on plasma-facing materials, effects of 3D fields on turbulence, and magnetic coil design. A closely coupled program of simulation, experimental validation, and design optimization is required to determine what forms and amplitudes of 3D shaping and symmetry breaking will be compatible with future fusion reactors. The development of models to address 3D physics and progress in these areas will be described. This work is supported both by the US Department of Energy under Contract DE-AC05-00OR22725 with UT-Battelle, LLC and under the US DOE SciDAC GSEP Center.

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

NASA Astrophysics Data System (ADS)

Majeski, Dick

2016-10-01

High edge electron temperatures (200 eV or greater) have been measured at the wall-limited plasma boundary in the Lithium Tokamak eXperiment (LTX). High edge temperatures, with flat electron temperature profiles, are a long-predicted consequence of low recycling boundary conditions. The temperature profile in LTX, measured by Thomson scattering, varies by as little as 10% from the plasma axis to the boundary, determined by the lithium-coated high field-side wall. The hydrogen plasma density in the outer scrape-off layer is very low, 2-3 x 1017 m-3 , consistent with a low recycling metallic lithium boundary. The plasma surface interaction in LTX is characterized by a low flux of high energy protons to the lithium PFC, with an estimated Debye sheath potential approaching 1 kV. Plasma-material interactions in LTX are consequently in a novel regime, where the impacting proton energy exceeds the peak in the sputtering yield for the lithium wall. In this regime, further increases in the edge temperature will decrease, rather than increase, the sputtering yield. Despite the high edge temperature, the core impurity content is low. Zeff is 1.2 - 1.5, with a very modest contribution (<0.1) from lithium. So far 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 which includes a 35A, 20 kV neutral beam injector to provide core fueling to maintain constant density, as well as auxiliary heating, is underway. Two beam systems have been loaned to LTX by Tri Alpha Energy. Additional results from LTX, as well as progress on the upgrade - LTX- β - will be discussed. Work supported by US DOE contracts DE-AC02-09CH11466 and DE-AC05-00OR22725.

Residual gas analysis for long-pulse, advanced tokamak operation.

PubMed

Klepper, C C; Hillis, D L; Bucalossi, J; Douai, D; Oddon, P; Vartanian, S; Colas, L; Manenc, L; Pégourié, B

2010-10-01

A shielded residual gas analyzer (RGA) system on Tore Supra can function during plasma operation and is set up to monitor the composition of the neutral gas in one of the pumping ducts of the toroidal pumped limited. This "diagnostic RGA" has been used in long-pulse (up to 6 min) discharges for continuous monitoring of up to 15 masses simultaneously. Comparison of the RGA-measured evolution of the H(2)/D(2) isotopic ratio in the exhaust gas to that measured by an energetic neutral particle analyzer in the plasma core provides a way to monitor the evolution of particle balance. RGA monitoring of corrective H(2) injection to maintain proper minority heating is providing a database for improved ion cyclotron resonance heating, potentially with RGA-base feedback control. In very long pulses (>4 min) absence of significant changes in the RGA-monitored, hydrocarbon particle pressures is an indication of proper operation of the actively cooled, carbon-based plasma facing components. Also H(2) could increase due to thermodesorption of overheated plasma facing components.

POD analysis of flow over a backward-facing step forced by right-angle-shaped plasma actuator.

PubMed

Wang, Bin; Li, Huaxing

2016-01-01

This study aims to present flow control over the backward-facing step with specially designed right-angle-shaped plasma actuator and analyzed the influence of various scales of flow structures on the Reynolds stress through snapshot proper orthogonal decomposition (POD). 2D particle image velocimetry measurements were conducted on region (x/h = 0-2.25) and reattachment zone in the x-y plane over the backward-facing step at a Reynolds number of Re h  = 27,766 (based on step height [Formula: see text] and free stream velocity [Formula: see text]. The separated shear layer was excited by specially designed right-angle-shaped plasma actuator under the normalized excitation frequency St h  ≈ 0.345 along the 45° direction. The spatial distribution of each Reynolds stress component was reconstructed using an increasing number of POD modes. The POD analysis indicated that the flow dynamic downstream of the step was dominated by large-scale flow structures, which contributed to streamwise Reynolds stress and Reynolds shear stress. The intense Reynolds stress localized to a narrow strip within the shear layer was mainly affected by small-scale flow structures, which were responsible for the recovery of the Reynolds stress peak. With plasma excitation, a significant increase was obtained in the vertical Reynolds stress peak. Under the dimensionless frequencies St h  ≈ 0.345 and [Formula: see text] which are based on the step height and momentum thickness, the effectiveness of the flow control forced by the plasma actuator along the 45° direction was ordinary. Only the vertical Reynolds stress was significantly affected.

Plasma under control: Advanced solutions and perspectives for plasma flux management in material treatment and nanosynthesis

NASA Astrophysics Data System (ADS)

Baranov, O.; Bazaka, K.; Kersten, H.; Keidar, M.; Cvelbar, U.; Xu, S.; Levchenko, I.

2017-12-01

Given the vast number of strategies used to control the behavior of laboratory and industrially relevant plasmas for material processing and other state-of-the-art applications, a potential user may find themselves overwhelmed with the diversity of physical configurations used to generate and control plasmas. Apparently, a need for clearly defined, physics-based classification of the presently available spectrum of plasma technologies is pressing, and the critically summary of the individual advantages, unique benefits, and challenges against key application criteria is a vital prerequisite for the further progress. To facilitate selection of the technological solutions that provide the best match to the needs of the end user, this work systematically explores plasma setups, focusing on the most significant family of the processes—control of plasma fluxes—which determine the distribution and delivery of mass and energy to the surfaces of materials being processed and synthesized. A novel classification based on the incorporation of substrates into plasma-generating circuitry is also proposed and illustrated by its application to a wide variety of plasma reactors, where the effect of substrate incorporation on the plasma fluxes is emphasized. With the key process and material parameters, such as growth and modification rates, phase transitions, crystallinity, density of lattice defects, and others being linked to plasma and energy fluxes, this review offers direction to physicists, engineers, and materials scientists engaged in the design and development of instrumentation for plasma processing and diagnostics, where the selection of the correct tools is critical for the advancement of emerging and high-performance applications.

Plasma dye coating as straightforward and widely applicable procedure for dye immobilization on polymeric materials.

PubMed

De Smet, Lieselot; Vancoillie, Gertjan; Minshall, Peter; Lava, Kathleen; Steyaert, Iline; Schoolaert, Ella; Van De Walle, Elke; Dubruel, Peter; De Clerck, Karen; Hoogenboom, Richard

2018-03-16

Here, we introduce a novel concept for the fabrication of colored materials with significantly reduced dye leaching through covalent immobilization of the desired dye using plasma-generated surface radicals. This plasma dye coating (PDC) procedure immobilizes a pre-adsorbed layer of a dye functionalized with a radical sensitive group on the surface through radical addition caused by a short plasma treatment. The non-specific nature of the plasma-generated surface radicals allows for a wide variety of dyes including azobenzenes and sulfonphthaleins, functionalized with radical sensitive groups to avoid significant dye degradation, to be combined with various materials including PP, PE, PA6, cellulose, and PTFE. The wide applicability, low consumption of dye, relatively short procedure time, and the possibility of continuous PDC using an atmospheric plasma reactor make this procedure economically interesting for various applications ranging from simple coloring of a material to the fabrication of chromic sensor fabrics as demonstrated by preparing a range of halochromic materials.

Novel biomaterials: plasma-enabled nanostructures and functions

NASA Astrophysics Data System (ADS)

Levchenko, Igor; Keidar, Michael; Cvelbar, Uroš; Mariotti, Davide; Mai-Prochnow, Anne; Fang, Jinghua; (Ken Ostrikov, Kostya

2016-07-01

Material processing techniques utilizing low-temperature plasmas as the main process tool feature many unique capabilities for the fabrication of various nanostructured materials. As compared with the neutral-gas based techniques and methods, the plasma-based approaches offer higher levels of energy and flux controllability, often leading to higher quality of the fabricated nanomaterials and sometimes to the synthesis of the hierarchical materials with interesting properties. Among others, nanoscale biomaterials attract significant attention due to their special properties towards the biological materials (proteins, enzymes), living cells and tissues. This review briefly examines various approaches based on the use of low-temperature plasma environments to fabricate nanoscale biomaterials exhibiting high biological activity, biological inertness for drug delivery system, and other features of the biomaterials make them highly attractive. In particular, we briefly discuss the plasma-assisted fabrication of gold and silicon nanoparticles for bio-applications; carbon nanoparticles for bioimaging and cancer therapy; carbon nanotube-based platforms for enzyme production and bacteria growth control, and other applications of low-temperature plasmas in the production of biologically-active materials.

Metallurgy and properties of plasma spray formed materials

NASA Technical Reports Server (NTRS)

Mckechnie, T. N.; Liaw, Y. K.; Zimmerman, F. R.; Poorman, R. M.

1992-01-01

Understanding the fundamental metallurgy of vacuum plasma spray formed materials is the key to enhancing and developing full material properties. Investigations have shown that the microstructure of plasma sprayed materials must evolve from a powder splat morphology to a recrystallized grain structure to assure high strength and ductility. A fully, or near fully, dense material that exhibits a powder splat morphology will perform as a brittle material compared to a recrystallized grain structure for the same amount of porosity. Metallurgy and material properties of nickel, iron, and copper base alloys will be presented and correlated to microstructure.

Recovery from Bell Palsy after Transplantation of Peripheral Blood Mononuclear Cells and Platelet-Rich Plasma.

PubMed

Seffer, Istvan; Nemeth, Zoltan

2017-06-01

Peripheral blood mononuclear cells (PBMCs) are multipotent, and plasma contains growth factors involving tissue regeneration. We hypothesized that transplantation of PBMC-plasma will promote the recovery of paralyzed facial muscles in Bell palsy. This case report describes the effects of PBMC-plasma transplantations in a 27-year-old female patient with right side Bell palsy. On the affected side of the face, the treatment resulted in both morphological and functional recovery including voluntary facial movements. These findings suggest that PBMC-plasma has the capacity of facial muscle regeneration and provides a promising treatment strategy for patients suffering from Bell palsy or other neuromuscular disorders.

Singled-walled carbon nanotubes produced by induction thermal plasma: Cytotoxicity evaluation of the feedstock materials and the final product for a potential bone application

NASA Astrophysics Data System (ADS)

Alinejad, Yasaman

One of the most challenging issues that the technologies related to nanomaterials face is the impact they have on human health and environment. It is therefore of great importance to investigate the toxicological impacts of these technologies prior to their widespread utilization in different fields of application. Therefore, in this study, the cytotoxicity of the materials present throughout the process of single-walled carbon nanotubes (SWCNTs) synthesis by induction thermal plasma (from the feedstock materials to the final product) was evaluated. First of all, the influence of the induction thermal plasma process on the physico-chemical and cytotoxic properties of feedstock materials (i.e. commercial Co, Ni, Y2O3, Mo catalysts and carbon black) was investigated. The strongest cytotoxicity was observed for commercial Co compared to other catalysts. Although the thermal plasma process affected the properties of all catalysts, only the cytotoxicity of Ni was increased. Comparing the properties and cytotoxicity of the plasma treated Ni particles with commercial Ni nanoparticles revealed that the particles with similar surface area had different cytotoxicities. Plus, the observed cytotoxicity of the catalysts was not mainly due to the release of ions. In order to evaluate the capacity of the RF induction thermal plasma process to produce high quality SWCNTs using non-toxic catalysts, the effects of the type and quantity of three catalyst mixtures (Ni-Y2O 3, Ni-Co-Y2O3, and Ni-Mo-Y2O3 ) on SWCNTs synthesis were examined. Thermodynamic calculations, in gas and particularly in liquid solution phases, were also performed. The results showed that catalyst type affected the quality of the SWCNT final product and similar quality SWCNTs was produced when the same amount of Co was replaced by Ni. Then, to investigate the cytotoxicity of the SWCNTs produced with the three catalyst mixtures, their effect was evaluated on the behavior of murine MC3T3-E1 preosteoblasts. Either SWCNTs were added on the attached cells or cells were seeded on the SWCNT-covered culture plates. SWCNTs which were added on the attached cells reduced cell viability drastically in a dose-dependent manner. However, the viability of the cells seeded on SWCNTs was only slightly decreased at 24 h, even on those produced with Ni-Co-Y2O3 . Moreover, cells could proliferate within 48 h. Thus, except mechanical membrane disturbance, thermal plasma grown SWCNTs seemed to induce no severe cytotoxicity on MC3T3-E1 preosteoblasts. Consequently, SWCNTs were purified and their influence on the viability and proliferation of MC3T3-E1 preosteoblasts was determined. The impact of SWCNTs on Smad activation and cell differentiation induced by BMP-2 and BMP-9 was also studied. SWCNTs pre-treatment accelerated the Smad1/5/8 activation induced by both BMP-2 and BMP-9. It did not reduce the viability of preosteoblasts but slightly affected their proliferation at 48 h. Furthermore, after 72 h incubation with BMP-2 or BMP-9, preosteoblasts pre-treated with SWCNTs for 24 h could express genes encoding osteogenic markers such as osterix and osteocalcin and showed high alkaline phosphatase activity. Interestingly, BMP-9 favored the differentiation of preosteoblasts pre-treated with SWCNTs more remarkably than BMP-2. Therefore, combination of BMP-9 with SWCNTs seems to be a promising avenue for bone regeneration. Keywords: Carbon nanotubes, metallic nanoparticles, induction thermal plasma, cytotoxicity, cell proliferation, mitochondrial enzymatic activity, lactate dehydrogenase, osteogenesis.

Mobile inductively coupled plasma system

DOEpatents

D'Silva, Arthur P.; Jaselskis, Edward J.

1999-03-30

A system for sampling and analyzing a material located at a hazardous site. A laser located remote from the hazardous site is connected to an optical fiber, which directs laser radiation proximate the material at the hazardous site. The laser radiation abates a sample of the material. An inductively coupled plasma is located remotely from the material. An aerosol transport system carries the ablated particles to a plasma, where they are dissociated, atomized and excited to provide characteristic optical reduction of the elemental constituents of the sample. An optical spectrometer is located remotely from the site. A second optical fiber is connected to the optical spectrometer at one end and the plasma source at the other end to carry the optical radiation from the plasma source to the spectrometer.

Sequential infiltration synthesis for advanced lithography

DOEpatents

Darling, Seth B.; Elam, Jeffrey W.; Tseng, Yu-Chih; Peng, Qing

2015-03-17

A plasma etch resist material modified by an inorganic protective component via sequential infiltration synthesis (SIS) and methods of preparing the modified resist material. The modified resist material is characterized by an improved resistance to a plasma etching or related process relative to the unmodified resist material, thereby allowing formation of patterned features into a substrate material, which may be high-aspect ratio features. The SIS process forms the protective component within the bulk resist material through a plurality of alternating exposures to gas phase precursors which infiltrate the resist material. The plasma etch resist material may be initially patterned using photolithography, electron-beam lithography or a block copolymer self-assembly process.

Electro-optic device with gap-coupled electrode

DOEpatents

Deri, Robert J.; Rhodes, Mark A.; Bayramian, Andrew J.; Caird, John A.; Henesian, Mark A.; Ebbers, Christopher A.

2013-08-20

An electro-optic device includes an electro-optic crystal having a predetermined thickness, a first face and a second face. The electro-optic device also includes a first electrode substrate disposed opposing the first face. The first electrode substrate includes a first substrate material having a first thickness and a first electrode coating coupled to the first substrate material. The electro-optic device further includes a second electrode substrate disposed opposing the second face. The second electrode substrate includes a second substrate material having a second thickness and a second electrode coating coupled to the second substrate material. The electro-optic device additionally includes a voltage source electrically coupled to the first electrode coating and the second electrode coating.

The effect of MTHFR(C677T) genotype on plasma homocysteine concentrations in healthy children is influenced by gender.

PubMed

Papoutsakis, C; Yiannakouris, N; Manios, Y; Papaconstantinou, E; Magkos, F; Schulpis, K H; Zampelas, A; Matalas, A L

2006-02-01

To explore the influence of gender, together with folate status, on the relation between the common methylenetetrahydrofolate reductase (MTHFR) C677T polymorphism and plasma total homocysteine (tHcy) concentrations in healthy children. Cross-sectional study by face-to-face interview. A total of 186 sixth-grade students participated from twelve randomly selected primary schools in Volos, Greece. Fasting tHcy, folate, and vitamin B(12) were measured in plasma. The MTHFR genotypes were determined. Anthropometric and dietary intake data by 24-h recall were collected. Geometric means for plasma tHcy, plasma folate and energy-adjusted dietary folate did not differ between females and males. The homozygous mutant TT genotype was associated with higher tHcy only in children with lower plasma folate concentrations (<19.9 nmol/l, P = 0.012). As a significant gender interaction was observed (P = 0.050), we stratified the lower plasma folate group by gender and found that the association between the genotype and tHcy was restricted to males (P = 0.026). Similar results were obtained when folate status was based on estimated dietary folate. Specifically, only TT males that reported lower dietary folate consumption (<37 microg/MJ/day) had tHcy that was significantly higher than tHcy levels of C-allele carriers (P = 0.001). Under conditions of lower folate status (as estimated by either plasma concentration or reported dietary consumption), gender modifies the association of the MTHFR(C677T) polymorphism with tHcy concentrations in healthy children. Kellog Europe.

Effects of fusion relevant transient energetic radiation, plasma and thermal load on PLANSEE double forged tungsten samples in a low-energy plasma focus device

NASA Astrophysics Data System (ADS)

Javadi, S.; Ouyang, B.; Zhang, Z.; Ghoranneviss, M.; Salar Elahi, A.; Rawat, R. S.

2018-06-01

Tungsten is the leading candidate for plasma facing component (PFC) material for thermonuclear fusion reactors and various efforts are ongoing to evaluate its performance or response to intense fusion relevant radiation, plasma and thermal loads. This paper investigates the effects of hot dense decaying pinch plasma, highly energetic deuterium ions and fusion neutrons generated in a low-energy (3.0 kJ) plasma focus device on the structure, morphology and hardness of the PLANSEE double forged tungsten (W) samples surfaces. The tungsten samples were provided by Forschungszentrum Juelich (FZJ), Germany via International Atomic Energy Agency, Vienna, Austria. Tungsten samples were irradiated using different number of plasma focus (PF) shots (1, 5 and 10) at a fixed axial distance of 5 cm from the anode top and also at various distances from the top of the anode (5, 7, 9 and 11 cm) using fixed number (5) of plasma focus shots. The virgin tungsten sample had bcc structure (α-W phase). After PF irradiation, the XRD analysis showed (i) the presence of low intensity new diffraction peak corresponding to β-W phase at (211) crystalline plane indicating the partial structural phase transition in some of the samples, (ii) partial amorphization, and (iii) vacancy defects formation and compressive stress in irradiated tungsten samples. Field emission scanning electron microscopy showed the distinctive changes to non-uniform surface with nanometer sized particles and particle agglomerates along with large surface cracks at higher number of irradiation shots. X-ray photoelectron spectroscopy analysis demonstrated the reduction in relative tungsten oxide content and the increase in metallic tungsten after irradiation. Hardness of irradiated samples initially increased for one shot exposure due to reduction in tungsten oxide phase, but then decreased with increasing number of shots due to increasing concentration of defects. It is demonstrated that the plasma focus device provides appropriate intense fusion relevant pulses for testing the structural, morphological and mechanical changes on irradiated tungsten samples.

Effect of the relative shift between the electron density and temperature pedestal position on the pedestal stability in JET-ILW and comparison with JET-C

NASA Astrophysics Data System (ADS)

Stefanikova, E.; Frassinetti, L.; Saarelma, S.; Loarte, A.; Nunes, I.; Garzotti, L.; Lomas, P.; Rimini, F.; Drewelow, P.; Kruezi, U.; Lomanowski, B.; de la Luna, E.; Meneses, L.; Peterka, M.; Viola, B.; Giroud, C.; Maggi, C.; contributors, JET

2018-05-01

The electron temperature and density pedestals tend to vary in their relative radial positions, as observed in DIII-D (Beurskens et al 2011 Phys. Plasmas 18 056120) and ASDEX Upgrade (Dunne et al 2017 Plasma Phys. Control. Fusion 59 14017). This so-called relative shift has an impact on the pedestal magnetohydrodynamic (MHD) stability and hence on the pedestal height (Osborne et al 2015 Nucl. Fusion 55 063018). The present work studies the effect of the relative shift on pedestal stability of JET ITER-like wall (JET-ILW) baseline low triangularity (δ) unseeded plasmas, and similar JET-C discharges. As shown in this paper, the increase of the pedestal relative shift is correlated with the reduction of the normalized pressure gradient, therefore playing a strong role in pedestal stability. Furthermore, JET-ILW tends to have a larger relative shift compared to JET carbon wall (JET-C), suggesting a possible role of the plasma facing materials in affecting the density profile location. Experimental results are then compared with stability analysis performed in terms of the peeling-ballooning model and with pedestal predictive model EUROPED (Saarelma et al 2017 Plasma Phys. Control. Fusion). Stability analysis is consistent with the experimental findings, showing an improvement of the pedestal stability, when the relative shift is reduced. This has been ascribed mainly to the increase of the edge bootstrap current, and to minor effects related to the increase of the pedestal pressure gradient and narrowing of the pedestal pressure width. Pedestal predictive model EUROPED shows a qualitative agreement with experiment, especially for low values of the relative shift.

Method of processing materials using an inductively coupled plasma

DOEpatents

Hull, D.E.; Bieniewski, T.M.

1987-04-13

A method of processing materials. The invention enables ultrafine, ultrapure powders to be formed from solid ingots in a gas free environment. A plasma is formed directly from an ingot which insures purity. The vaporized material is expanded through a nozzle and the resultant powder settles on a cold surface. An inductively coupled plasma may also be used to process waste chemicals. Noxious chemicals are directed through a series of plasma tubes, breaking molecular bonds and resulting in relatively harmless atomic constituents. 3 figs.

Silicon Carbide as a tritium permeation barrier in tungsten plasma-facing components

NASA Astrophysics Data System (ADS)

Wright, G. M.; Durrett, M. G.; Hoover, K. W.; Kesler, L. A.; Whyte, D. G.

2015-03-01

The control of tritium inventory is of great importance in future fusion reactors, not only from a safety standpoint but also to maximize a reactor's efficiency. Due to the high mobility of hydrogenic species in tungsten (W) one concern is the loss of tritium from the system via permeation through the tungsten plasma-facing components (PFC). This can lead to loss of tritium through the cooling channels of the wall thereby mandating tritium monitoring and recovery methods for the cooling system of the first wall. The permeated tritium is then out of the fuel cycle and cannot contribute to energy production until it is recovered and recycled into the system.

Plasma power recycling at the divertor surface

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tang, Xian -Zhu; Guo, Zehua

With a divertor made of solid materials like carbon and tungsten, plasma ions are expected to be recycled at the divertor surface with a time-averaged particle recycling coefficient very close to unity in steady-state operation. This means that almost every plasma ion (hydrogen and helium) will be returned to the plasma, mostly as neutrals. The power flux deposited by the plasma on the divertor surface, on the other hand, can have varying recycling characteristics depending on the material choice of the divertor; the run-time atomic composition of the surface, which can be modified by material mix due to impurity migrationmore » in the chamber; and the surface morphology change over time. In general, a high-Z–material (such as tungsten) surface tends to reflect light ions and produce stronger power recycling, while a low-Z–material (such as carbon) surface tends to have a larger sticking coefficient for light ions and hence lower power recycling. Here, an explicit constraint on target plasma density and temperature is derived from the truncated bi-Maxwellian sheath model, in relation to the absorbed power load and power recycling coefficient at the divertor surface. Lastly, it is shown that because of the surface recombination energy flux, the attached plasma has a sharper response to power recycling in comparison to a detached plasma.« less

Plasma power recycling at the divertor surface

DOE PAGES

Tang, Xian -Zhu; Guo, Zehua

2016-12-03

With a divertor made of solid materials like carbon and tungsten, plasma ions are expected to be recycled at the divertor surface with a time-averaged particle recycling coefficient very close to unity in steady-state operation. This means that almost every plasma ion (hydrogen and helium) will be returned to the plasma, mostly as neutrals. The power flux deposited by the plasma on the divertor surface, on the other hand, can have varying recycling characteristics depending on the material choice of the divertor; the run-time atomic composition of the surface, which can be modified by material mix due to impurity migrationmore » in the chamber; and the surface morphology change over time. In general, a high-Z–material (such as tungsten) surface tends to reflect light ions and produce stronger power recycling, while a low-Z–material (such as carbon) surface tends to have a larger sticking coefficient for light ions and hence lower power recycling. Here, an explicit constraint on target plasma density and temperature is derived from the truncated bi-Maxwellian sheath model, in relation to the absorbed power load and power recycling coefficient at the divertor surface. Lastly, it is shown that because of the surface recombination energy flux, the attached plasma has a sharper response to power recycling in comparison to a detached plasma.« less

Very thick mixture oxide ion beam sputtering films for investigation of nonlinear material properties

NASA Astrophysics Data System (ADS)

Steinecke, Morten; Kiedrowski, Kevin; Jupé, Marco; Ristau, Detlev

2017-11-01

Currently, optical coating technology is facing a multitude of new challenges. Some of the new requirements are addressed to the spectral behavior of complex coatings, but in addition, the power handling capabilities gain in importance. Often, both demands are combined in the same component, for example in chirped mirrors for ultra-short pulse applications. The consequent demands on the accuracy of the layer thicknesses and the stability of the refractive indices require a deposition by sputtering processes. For high end components, Ion Beam Sputtering (IBS) is often the method of choice. Utilizing the Co-sputtering technique, IBS additionally allows a higher flexibility in the possible coating materials by mixing two pure oxides into one ternary composite material. These composite materials are also advantageous for researching third order nonlinear effects, which can limit the functionality of optics at high powers. The layer thicknesses required for this fundamental research often exceed 100 µm, which therefore makes low stress and absorption in the layer materials mandatory. A reduction of these decisive properties can be achieved by a thermal treatment of the sample. Usually, this is performed by a post-deposition annealing. Alternatively, the coating temperature can be increased. This is rarely done for IBS processes, but it can be assumed, that the effect is comparable to that of ex-situ annealing. In this work, different ternary mixtures of Al2O3/SiO2, HfO2/Al2O3 as well as Nb2O5/Al2O3 were investigated for their layer stress and absorption, applying both, in-situ temperature treatment as well as post manufacturing annealing. It is observed that suitable thermal treatment as well as material composition can significantly reduce layer stress and absorption in the deposited layer. This enabled the manufacturing of layers with thicknesses of over 180 µm as well as the measurement of nonlinear properties of the deposited materials. Contribution to the topical issue "Plasma Sources and Plasma Processes (PSPP)", edited by Luis Lemos Alves, Thierry Belmonte and Tiberiu Minea

Method of processing materials using an inductively coupled plasma

DOEpatents

Hull, Donald E.; Bieniewski, Thomas M.

1990-01-01

A method for making fine power using an inductively coupled plasma. The method provides a gas-free environment, since the plasma is formed without using a gas. The starting material used in the method is in solid form.

Process to make core-shell structured nanoparticles

DOEpatents

Luhrs, Claudia; Phillips, Jonathan; Richard, Monique N

2014-01-07

Disclosed is a process for making a composite material that contains core-shell structured nanoparticles. The process includes providing a precursor in the form of a powder a liquid and/or a vapor of a liquid that contains a core material and a shell material, and suspending the precursor in an aerosol gas to produce an aerosol containing the precursor. In addition, the process includes providing a plasma that has a hot zone and passing the aerosol through the hot zone of the plasma. As the aerosol passes through the hot zone of the plasma, at least part of the core material and at least part of the shell material in the aerosol is vaporized. Vapor that contains the core material and the shell material that has been vaporized is removed from the hot zone of the plasma and allowed to condense into core-shell structured nanoparticles.

29 CFR 1910.133 - Eye and face protection.

Code of Federal Regulations, 2011 CFR

2011-07-01

... side shields) meeting the pertinent requirements of this section are acceptable. (3) The employer shall...) less than 500 10 Arc cutting (Heavy) 500-1000 11 Plasma arc welding less than 20 6 20-100 8 100-400 10 400-800 11 Plasma arc cutting (light)** less than 300 8 (medium)** 300-400 9 (heavy)** 400-800 10...

29 CFR 1910.133 - Eye and face protection.

Code of Federal Regulations, 2012 CFR

2012-07-01

... side shields) meeting the pertinent requirements of this section are acceptable. (3) The employer shall...) less than 500 10 Arc cutting (Heavy) 500-1000 11 Plasma arc welding less than 20 6 20-100 8 100-400 10 400-800 11 Plasma arc cutting (light)** less than 300 8 (medium)** 300-400 9 (heavy)** 400-800 10...

29 CFR 1910.133 - Eye and face protection.

Code of Federal Regulations, 2010 CFR

2010-07-01

... side shields) meeting the pertinent requirements of this section are acceptable. (3) The employer shall...) less than 500 10 Arc cutting (Heavy) 500-1000 11 Plasma arc welding less than 20 6 20-100 8 100-400 10 400-800 11 Plasma arc cutting (light)** less than 300 8 (medium)** 300-400 9 (heavy)** 400-800 10...

29 CFR 1910.133 - Eye and face protection.

Code of Federal Regulations, 2014 CFR

2014-07-01

... side shields) meeting the pertinent requirements of this section are acceptable. (3) The employer shall...) less than 500 10 Arc cutting (Heavy) 500-1000 11 Plasma arc welding less than 20 6 20-100 8 100-400 10 400-800 11 Plasma arc cutting (light)** less than 300 8 (medium)** 300-400 9 (heavy)** 400-800 10...

29 CFR 1910.133 - Eye and face protection.

Code of Federal Regulations, 2013 CFR

2013-07-01

... side shields) meeting the pertinent requirements of this section are acceptable. (3) The employer shall...) less than 500 10 Arc cutting (Heavy) 500-1000 11 Plasma arc welding less than 20 6 20-100 8 100-400 10 400-800 11 Plasma arc cutting (light)** less than 300 8 (medium)** 300-400 9 (heavy)** 400-800 10...

Mobile inductively coupled plasma system

DOEpatents

D`Silva, A.P.; Jaselskis, E.J.

1999-03-30

A system is described for sampling and analyzing a material located at a hazardous site. A laser located remotely from the hazardous site is connected to an optical fiber, which directs laser radiation proximate the material at the hazardous site. The laser radiation abates a sample of the material. An inductively coupled plasma is located remotely from the material. An aerosol transport system carries the ablated particles to a plasma, where they are dissociated, atomized and excited to provide characteristic optical reduction of the elemental constituents of the sample. An optical spectrometer is located remotely from the site. A second optical fiber is connected to the optical spectrometer at one end and the plasma source at the other end to carry the optical radiation from the plasma source to the spectrometer. 10 figs.

Helicon plasma ion temperature measurements and observed ion cyclotron heating in proto-MPEX

NASA Astrophysics Data System (ADS)

Beers, C. J.; Goulding, R. H.; Isler, R. C.; Martin, E. H.; Biewer, T. M.; Caneses, J. F.; Caughman, J. B. O.; Kafle, N.; Rapp, J.

2018-01-01

The Prototype-Material Plasma Exposure eXperiment (Proto-MPEX) linear plasma device is a test bed for exploring and developing plasma source concepts to be employed in the future steady-state linear device Material Plasma Exposure eXperiment (MPEX) that will study plasma-material interactions for the nuclear fusion program. The concept foresees using a helicon plasma source supplemented with electron and ion heating systems to reach necessary plasma conditions. In this paper, we discuss ion temperature measurements obtained from Doppler broadening of spectral lines from argon ion test particles. Plasmas produced with helicon heating alone have average ion temperatures downstream of the Helicon antenna in the range of 3 ± 1 eV; ion temperature increases to 10 ± 3 eV are observed with the addition of ion cyclotron heating (ICH). The temperatures are higher at the edge than the center of the plasma either with or without ICH. This type of profile is observed with electrons as well. A one-dimensional RF antenna model is used to show where heating of the plasma is expected.

Sequential infiltration synthesis for advanced lithography

DOE Office of Scientific and Technical Information (OSTI.GOV)

Darling, Seth B.; Elam, Jeffrey W.; Tseng, Yu-Chih

A plasma etch resist material modified by an inorganic protective component via sequential infiltration synthesis (SIS) and methods of preparing the modified resist material. The modified resist material is characterized by an improved resistance to a plasma etching or related process relative to the unmodified resist material, thereby allowing formation of patterned features into a substrate material, which may be high-aspect ratio features. The SIS process forms the protective component within the bulk resist material through a plurality of alternating exposures to gas phase precursors which infiltrate the resist material. The plasma etch resist material may be initially patterned usingmore » photolithography, electron-beam lithography or a block copolymer self-assembly process.« less

A computer model of solar panel-plasma interactions

NASA Technical Reports Server (NTRS)

Cooke, D. L.; Freeman, J. W.

1980-01-01

High power solar arrays for satellite power systems are presently being planned with dimensions of kilometers, and with tens of kilovolts distributed over their surface. Such systems face many plasma interaction problems, such as power leakage to the plasma, particle focusing, and anomalous arcing. These effects cannot be adequately modeled without detailed knowledge of the plasma sheath structure and space charge effects. Laboratory studies of 1 by 10 meter solar array in a simulated low Earth orbit plasma are discussed. The plasma screening process is discussed, program theory is outlined, and a series of calibration models is presented. These models are designed to demonstrate that PANEL is capable of accurate self consistant space charge calculations. Such models include PANEL predictions for the Child-Langmuir diode problem.

Elastic-plastic adhesive impacts of tungsten dust with metal surfaces in plasma environments

NASA Astrophysics Data System (ADS)

Ratynskaia, S.; Tolias, P.; Shalpegin, A.; Vignitchouk, L.; De Angeli, M.; Bykov, I.; Bystrov, K.; Bardin, S.; Brochard, F.; Ripamonti, D.; den Harder, N.; De Temmerman, G.

2015-08-01

Dust-surface collisions impose size selectivity on the ability of dust grains to migrate in scrape-off layer and divertor plasmas and to adhere to plasma-facing components. Here, we report first experimental evidence of dust impact phenomena in plasma environments concerning low-speed collisions of tungsten dust with tungsten surfaces: re-bouncing, adhesion, sliding and rolling. The results comply with the predictions of the model of elastic-perfectly plastic adhesive spheres employed in the dust dynamics code MIGRAINe for sub- to several meters per second impacts of micrometer-range metal dust.

Planar controlled zone microwave plasma system

DOEpatents

Ripley, Edward B [Knoxville, TN; Seals, Roland D [Oak Ridge, TN; Morrell, Jonathan S [Knoxvlle, TN

2011-10-04

An apparatus and method for initiating a process gas plasma. A conductive plate having a plurality of conductive fingers is positioned in a microwave applicator. An arc forms between the conductive fingers to initiate the formation of a plasma. A transport mechanism may convey process materials through the plasma. A spray port may be provided to expel processed materials.

Controlled zone microwave plasma system

DOEpatents

Ripley, Edward B [Knoxville, TN; Seals, Roland D [Oak Ridge, TN; Morrell, Jonathan S [Knoxville, TN

2009-10-20

An apparatus and method for initiating a process gas plasma. A conductive plate having a plurality of conductive fingers is positioned in a microwave applicator. An arc forms between the conductive fingers to initiate the formation of a plasma. A transport mechanism may convey process materials through the plasma. A spray port may be provided to expel processed materials.

Vacuum arc plasma thrusters with inductive energy storage driver

NASA Technical Reports Server (NTRS)

Schein, Jochen (Inventor); Gerhan, Andrew N. (Inventor); Woo, Robyn L. (Inventor); Au, Michael Y. (Inventor); Krishnan, Mahadevan (Inventor)

2004-01-01

An apparatus for producing a vacuum arc plasma source device using a low mass, compact inductive energy storage circuit powered by a low voltage DC supply acts as a vacuum arc plasma thruster. An inductor is charged through a switch, subsequently the switch is opened and a voltage spike of Ldi/dt is produced initiating plasma across a resistive path separating anode and cathode. The plasma is subsequently maintained by energy stored in the inductor. Plasma is produced from cathode material, which allows for any electrically conductive material to be used. A planar structure, a tubular structure, and a coaxial structure allow for consumption of cathode material feed and thereby long lifetime of the thruster for long durations of time.

Composite electrode for use in electrochemical cells

DOEpatents

Vanderborgh, N.E.; Huff, J.R.; Leddy, J.

1987-10-16

A porous composite electrode for use in electrochemical cells. The electrode has a first face and a second face defining a relatively thin section therebetween. The electrode is comprised of an ion conducting material, an electron conducting material, and an electrocatalyst. The volume concentration of the ion conducting material is greatest at the first face and is decreased across the section, while the volume concentration of the electron conducting material is greatest at the second face and decreases across the section of the electrode. Substantially all of the electrocatalyst is positioned within the electrode section in a relatively narrow zone where the rate of electron transport of the electrode is approximately equal to the rate of ion transport of the electrode. 4 figs., 1 tab.

Composite electrode for use in electrochemical cells

DOEpatents

Vanderborgh, Nicholas E.; Huff, James R.; Leddy, Johna

1989-01-01

A porous composite electrode for use in electrochemical cells. The electrode has a first face and a second face defining a relatively thin section therebetween. The electrode is comprised of an ion conducting material, an electron conducting material, and an electrocatalyst. The volume concentration of the ion conducting material is greatest at the first face and is decreased across the section, while the volume concentration of the electron conducting material is greatest at the second face and decreases across the section of the electrode. Substantially all of the electrocatalyst is positioned within the electrode section in a relatively narrow zone where the rate of electron transport of the electrode is approximately equal to the rate of ion transport of the electrode.

Pulse thermal processing of functional materials using directed plasma arc

DOEpatents

Ott, Ronald D [Knoxville, TN; Blue, Craig A [Knoxville, TN; Dudney, Nancy J [Knoxville, TN; Harper, David C [Kingston, TN

2007-05-22

A method of thermally processing a material includes exposing the material to at least one pulse of infrared light emitted from a directed plasma arc to thermally process the material, the pulse having a duration of no more than 10 s.

Progress in the Development of a High Power Helicon Plasma Source for the Materials Plasma Exposure Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goulding, Richard Howell; Caughman, John B.; Rapp, Juergen

Proto-MPEX is a linear plasma device being used to study a novel RF source concept for the planned Material Plasma Exposure eXperiment (MPEX), which will address plasma-materials interaction (PMI) for nuclear fusion reactors. Plasmas are produced using a large diameter helicon source operating at a frequency of 13.56 MHz at power levels up to 120 kW. In recent experiments the helicon source has produced deuterium plasmas with densities up to ~6 × 1019 m–3 measured at a location 2 m downstream from the antenna and 0.4 m from the target. Previous plasma production experiments on Proto-MPEX have generated lower densitymore » plasmas with hollow electron temperature profiles and target power deposition peaked far off axis. The latest experiments have produced flat Te profiles with a large portion of the power deposited on the target near the axis. This and other evidence points to the excitation of a helicon mode in this case.« less

Diffusivity and solubility of hydrogen in the carbon fibre composite SEP N11

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alberici, S.; Perujo, A.; Camposilvan, J.

1995-10-01

In this paper we present the hydrogen diffusivity and solubility in the carbon fibre composite (CFC) SEP N11 with tri-directional fibres structure that is a possible candidate as armour material for plasma facing components (PFC). The technique used for these measurements is a gas evolution method and the measurements were carried out in the temperature range 900 - 1200 K with a loading hydrogen pressure of 100 kPa. The results obtained showed that the Sieverts` constant K{sub s} is of the same order of magnitude as those previously obtained for several graphites, while the diffusivity is about five to sixmore » orders of magnitude higher as compared to graphites. Furthermore, CFC presents an endothermic behaviour in contrast to graphites. 10 refs., 3 figs.« less

Suppression of tritium retention in remote areas of ITER by nonperturbative reactive gas injection.

PubMed

Tabarés, F L; Ferreira, J A; Ramos, A; van Rooij, G; Westerhout, J; Al, R; Rapp, J; Drenik, A; Mozetic, M

2010-10-22

A technique based on reactive gas injection in the afterglow region of the divertor plasma is proposed for the suppression of tritium-carbon codeposits in remote areas of ITER when operated with carbon-based divertor targets. Experiments in a divertor simulator plasma device indicate that a 4  nm/min deposition can be suppressed by addition of 1  Pa·m³ s⁻¹ ammonia flow at 10 cm from the plasma. These results bolster the concept of nonperturbative scavenger injection for tritium inventory control in carbon-based fusion plasma devices, thus paving the way for ITER operation in the active phase under a carbon-dominated, plasma facing component background.

Evaluation of the growth of carbonaceous deposit in steady state Tore Supra using infrared thermography

NASA Astrophysics Data System (ADS)

Mitteau, R.; Guilhem, D.; Reichle, R.; Vallet, J. C.; Roche, H.; Buravand, Y.; Chantant, M.; Tsitrone, E.; Brosset, C.; Grosman, A.; Chappuis, P.

2006-03-01

Fusion devices with carbon as the main armour material are experiencing a growth in carbonaceous deposits at the surface of the plasma facing components. Tore Supra presents such deposits, and has specific features which influence their growth: long pulse operation and cooled walls. Deposits have a low thermal transfer to the cooled structure so that they appear as hot areas with the infrared imaging system looking at the elements surface temperature during plasma discharges. A 'degree of (carbon) deposit' on the toroidal pumped limiter is estimated by establishing the ratio between the apparent power on the limiter derived from the infrared measure and the actual one, deduced from a power balance analysis between the injected and the radiated power. This criterion is used to monitor the evolution of the deposit average thermal resistance. Successive shots have a similar 'degree of deposit', showing that the evaluation makes sense. Two years of data have been compiled (2003 and 2004), representing 3000 discharges (13 h of plasma, including 30 discharges longer than one minute). A three-fold increase in the 'degree of deposit' over six months is evidenced, following a limiter clean-up early in 2003. A comparison with calorimetric data produces a similar result, albeit less pronounced. Large steps in the degree of deposit are sometimes observed, usually correlated with identified events such as disruption, vessel opening, conditioning or plasma parameters change. It indicates that the deposit thermal resistance can change rapidly, although a systematic correlation with the above mentioned events could not be established.

Methods for Using Durable Adhesively Bonded Joints for Sandwich Structures

NASA Technical Reports Server (NTRS)

Smeltzer, Stanley S., III (Inventor); Lundgren, Eric C. (Inventor)

2016-01-01

Systems, methods, and apparatus for increasing durability of adhesively bonded joints in a sandwich structure. Such systems, methods, and apparatus includes an first face sheet and an second face sheet as well as an insert structure, the insert structure having a first insert face sheet, a second insert face sheet, and an insert core material. In addition, sandwich core material is arranged between the first face sheet and the second face sheet. A primary bondline may be coupled to the face sheet(s) and the splice. Further, systems, methods, and apparatus of the present disclosure advantageously reduce the load, provide a redundant path, reduce structural fatigue, and/or increase fatigue life.

Systems, Apparatuses, and Methods for Using Durable Adhesively Bonded Joints for Sandwich Structures

NASA Technical Reports Server (NTRS)

Smeltzer, III, Stanley S. (Inventor); Lundgren, Eric C. (Inventor)

2014-01-01

Systems, methods, and apparatus for increasing durability of adhesively bonded joints in a sandwich structure. Such systems, methods, and apparatus includes an first face sheet and an second face sheet as well as an insert structure, the insert structure having a first insert face sheet, a second insert face sheet, and an insert core material. In addition, sandwich core material is arranged between the first face sheet and the second face sheet. A primary bondline may be coupled to the face sheet(s) and the splice. Further, systems, methods, and apparatus of the present disclosure advantageously reduce the load, provide a redundant path, reduce structural fatigue, and/or increase fatigue life.

Development of active porous medium filters based on plasma textiles

NASA Astrophysics Data System (ADS)

Kuznetsov, Ivan A.; Saveliev, Alexei V.; Rasipuram, Srinivasan; Kuznetsov, Andrey V.; Brown, Alan; Jasper, Warren

2012-05-01

Inexpensive, flexible, washable, and durable materials that serve as antimicrobial filters and self-decontaminating fabrics are needed to provide active protection to people in areas regularly exposed to various biohazards, such as hospitals and bio research labs working with pathogens. Airlines and cruise lines need such material to combat the spread of infections. In households these materials can be used in HVAC filters to fight indoor pollution, which is especially dangerous to people suffering from asthma. Efficient filtering materials are also required in areas contaminated by other types of hazardous dust particulates, such as nuclear dust. The primary idea that guided the undertaken study is that a microplasma-generating structure can be embedded in a textile fabric to generate a plasma sheath ("plasma shield") that kills bacterial agents coming in contact with the fabric. The research resulted in the development of a plasma textile that can be used for producing new types of self-decontaminating garments, fabrics, and filter materials, capable of activating a plasma sheath that would filter, capture, and destroy any bacteriological agent deposited on its surface. This new material relies on the unique antimicrobial and catalytic properties of cold (room temperature) plasma that is benign to people and does not cause thermal damage to many polymer textiles, such as Nomex and polypropylene. The uniqueness of cold plasma as a disinfecting agent lies in the inability of bacteria to develop resistance to plasma exposure, as they can for antibiotics. Plasma textiles could thus be utilized for microbial destruction in active antimicrobial filters (for continuous decontamination and disinfection of large amounts of air) as well as in self-decontaminating surfaces and antibacterial barriers (for example, for creating local antiseptic or sterile environments around wounds and burns).

Development of active porous medium filters based on plasma textiles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuznetsov, Ivan A.; Saveliev, Alexei V.; Rasipuram, Srinivasan

2012-05-15

Inexpensive, flexible, washable, and durable materials that serve as antimicrobial filters and self-decontaminating fabrics are needed to provide active protection to people in areas regularly exposed to various biohazards, such as hospitals and bio research labs working with pathogens. Airlines and cruise lines need such material to combat the spread of infections. In households these materials can be used in HVAC filters to fight indoor pollution, which is especially dangerous to people suffering from asthma. Efficient filtering materials are also required in areas contaminated by other types of hazardous dust particulates, such as nuclear dust. The primary idea that guidedmore » the undertaken study is that a microplasma-generating structure can be embedded in a textile fabric to generate a plasma sheath (''plasma shield'') that kills bacterial agents coming in contact with the fabric. The research resulted in the development of a plasma textile that can be used for producing new types of self-decontaminating garments, fabrics, and filter materials, capable of activating a plasma sheath that would filter, capture, and destroy any bacteriological agent deposited on its surface. This new material relies on the unique antimicrobial and catalytic properties of cold (room temperature) plasma that is benign to people and does not cause thermal damage to many polymer textiles, such as Nomex and polypropylene. The uniqueness of cold plasma as a disinfecting agent lies in the inability of bacteria to develop resistance to plasma exposure, as they can for antibiotics. Plasma textiles could thus be utilized for microbial destruction in active antimicrobial filters (for continuous decontamination and disinfection of large amounts of air) as well as in self-decontaminating surfaces and antibacterial barriers (for example, for creating local antiseptic or sterile environments around wounds and burns).« less

Plasma-assisted conversion of solid hydrocarbon to diamond

DOEpatents

Valone, Steven M.; Pattillo, Stevan G.; Trkula, Mitchell; Coates, Don M.; Shah, S. Ismat

1996-01-01

A process of preparing diamond, e.g., diamond fiber, by subjecting a hydrocarbon material, e.g., a hydrocarbon fiber, to a plasma treatment in a gaseous feedstream for a sufficient period of time to form diamond, e.g., a diamond fiber is disclosed. The method generally further involves pretreating the hydrocarbon material prior to treatment with the plasma by heating within an oxygen-containing atmosphere at temperatures sufficient to increase crosslinking within said hydrocarbon material, but at temperatures insufficient to melt or decompose said hydrocarbon material, followed by heating at temperatures sufficient to promote outgassing of said crosslinked hydrocarbon material, but at temperatures insufficient to convert said hydrocarbon material to carbon.

Characterization of Liquid Lithium Wetting and Thermoelectric Properties for Nuclear Fusion Applications

NASA Astrophysics Data System (ADS)

Fiflis, Peter; Xu, Wenyu; Christenson, Michael; Andruczyk, Daniel; Curreli, Davide; Ruzic, David

2013-10-01

Critical to the implementation of flowing liquid lithium plasma facing components is understanding the interactions of liquid lithium with various surfaces. Presented here are experiments investigating the material compatibility, wetting characteristics, and relative thermopower of liquid lithium with a variety of potential substrate candidates for the LiMIT concept. Wetting experiments with lithium used the contact angle as a metric. Among those materials investigated are 316 SS, Mo, Ta, and W. The contact angle, as well as its dependence on temperature was measured. For example, at 200 C, tungsten registers a contact angle of 130°, whereas above its wetting temperature of 350 C, the contact angle is less than 80°. Several methods were found to decrease the critical wetting temperature of various materials and are presented here. The thermopower of W, Mo, Ta, Li, Ga, Wood's metal and Sn has been measured relative to stainless steel, and the Seebeck coefficient of has then been calculated. For molybdenum the Seebeck coefficient has a linear rise with temperature from SMo = 3.9 μVK-1 at 30 °C to 7.5 μVK-1 at 275 °C. On Assignment at PPPL

The cathode material for a plasma-arc heater

NASA Astrophysics Data System (ADS)

Yelyutin, A. V.; Berlin, I. K.; Averyanov, V. V.; Kadyshevskii, V. S.; Savchenko, A. A.; Putintseva, R. G.

1983-11-01

The cathode of a plasma arc heater experiences a large thermal load. The temperature of its working surface, which is in contact with the plasma, reaches high values, as a result of which the electrode material is subject to erosion. Refractory metals are usually employed for the cathode material, but because of the severe erosion do not usually have a long working life. The most important electrophysical characteristic of the electrode is the electron work function. The use of materials with a low electron work function allows a decrease in the heat flow to the cathode, and this leads to an increase in its erosion resistance and working life. The electroerosion of certain materials employed for the cathode in an electric arc plasma generator in the process of reduction smelting of refractory metals was studied.

Platelet-Rich Plasma with Basic Fibroblast Growth Factor for Treatment of Wrinkles and Depressed Areas of the Skin.

PubMed

Kamakura, Tatsuro; Kataoka, Jiro; Maeda, Kazuhiko; Teramachi, Hideaki; Mihara, Hisayuki; Miyata, Kazuhiro; Ooi, Kouichi; Sasaki, Naomi; Kobayashi, Miyuki; Ito, Kouhei

2015-11-01

There are several treatments for wrinkles and depressed areas of the face, hands, and body. Hyaluronic acid is effective, but only for 6 months to 1 year. Autologous fat grafting may cause damage during tissue harvest. In this study, patients were injected with platelet-rich plasma plus basic fibroblast growth factor (bFGF). Platelet-rich plasma was prepared by collecting blood and extracting platelets using double centrifugation. Basic fibroblast growth factor diluted with normal saline was added to platelet-rich plasma. There were 2005 patients who received platelet-rich plasma plus bFGF therapy. Of the 2005 patients treated, 1889 were female and 116 were male patients; patients had a mean age of 48.2 years. Treated areas inlcuded 1461 nasolabial folds, 437 marionette lines, 1413 nasojugal grooves, 148 supraorbital grooves, 253 midcheek grooves, 304 foreheads, 49 temples, and 282 glabellae. Results on the Global Aesthetic Improvement Scale indicated that the level of patient satisfaction was 97.3 percent and the level of investigator satisfaction was 98.4 percent. The period for the therapy's effectiveness to become apparent was an average of 65.4 days. Platelet-rich plasma plus bFGF therapy resulted in an improved grade on the Wrinkle Severity Rating Scale. Improvement was 0.55 for a Wrinkle Severity Rating Scale grade of 2, 1.13 for a Wrinkle Severity Rating Scale grade of 3, 1.82 for a Wrinkle Severity Rating Scale grade of 4, and 2.23 for a Wrinkle Severity Rating Scale grade of 5. Platelet-rich plasma plus bFGF is effective in treating wrinkles and depressed areas of the skin of the face and body. The study revealed that platelet-rich plasma plus bFGF is an innovative therapy that causes minimal complications. Therapeutic, IV.

Theory and experiments characterizing hypervelocity impact plasmas on biased spacecraft materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Nicolas; Close, Sigrid; Goel, Ashish

2013-03-15

Space weather including solar activity and background plasma sets up spacecraft conditions that can magnify the threat from hypervelocity impacts. Hypervelocity impactors include both meteoroids, traveling between 11 and 72 km/s, and orbital debris, with typical impact speeds of 10 km/s. When an impactor encounters a spacecraft, its kinetic energy is converted over a very short timescale into energy of vaporization and ionization, resulting in a small, dense plasma. This plasma can produce radio frequency (RF) emission, causing electrical anomalies within the spacecraft. In order to study this phenomenon, we conducted ground-based experiments to study hypervelocity impact plasmas using amore » Van de Graaff dust accelerator. Iron projectiles ranging from 10{sup -16} g to 10{sup -11} g were fired at speeds of up to 70 km/s into a variety of target materials under a range of surface charging conditions representative of space weather effects. Impact plasmas associated with bare metal targets as well as spacecraft materials were studied. Plasma expansion models were developed to determine the composition and temperature of the impact plasma, shedding light on the plasma dynamics that can lead to spacecraft electrical anomalies. The dependence of these plasma properties on target material, impact speed, and surface charge was analyzed. Our work includes three major results. First, the initial temperature of the impact plasma is at least an order of magnitude lower than previously reported, providing conditions more favorable for sustained RF emission. Second, the composition of impact plasmas from glass targets, unlike that of impact plasmas from tungsten, has low dependence on impact speed, indicating a charge production mechanism that is significant down to orbital debris speeds. Finally, negative ion formation has a strong dependence on target material. These new results can inform the design and operation of spacecraft in order to mitigate future impact-related space weather anomalies and failures.« less

Theory and experiments characterizing hypervelocity impact plasmas on biased spacecraft materials

NASA Astrophysics Data System (ADS)

Lee, Nicolas; Close, Sigrid; Goel, Ashish; Lauben, David; Linscott, Ivan; Johnson, Theresa; Strauss, David; Bugiel, Sebastian; Mocker, Anna; Srama, Ralf

2013-03-01

Space weather including solar activity and background plasma sets up spacecraft conditions that can magnify the threat from hypervelocity impacts. Hypervelocity impactors include both meteoroids, traveling between 11 and 72 km/s, and orbital debris, with typical impact speeds of 10 km/s. When an impactor encounters a spacecraft, its kinetic energy is converted over a very short timescale into energy of vaporization and ionization, resulting in a small, dense plasma. This plasma can produce radio frequency (RF) emission, causing electrical anomalies within the spacecraft. In order to study this phenomenon, we conducted ground-based experiments to study hypervelocity impact plasmas using a Van de Graaff dust accelerator. Iron projectiles ranging from 10-16 g to 10-11 g were fired at speeds of up to 70 km/s into a variety of target materials under a range of surface charging conditions representative of space weather effects. Impact plasmas associated with bare metal targets as well as spacecraft materials were studied. Plasma expansion models were developed to determine the composition and temperature of the impact plasma, shedding light on the plasma dynamics that can lead to spacecraft electrical anomalies. The dependence of these plasma properties on target material, impact speed, and surface charge was analyzed. Our work includes three major results. First, the initial temperature of the impact plasma is at least an order of magnitude lower than previously reported, providing conditions more favorable for sustained RF emission. Second, the composition of impact plasmas from glass targets, unlike that of impact plasmas from tungsten, has low dependence on impact speed, indicating a charge production mechanism that is significant down to orbital debris speeds. Finally, negative ion formation has a strong dependence on target material. These new results can inform the design and operation of spacecraft in order to mitigate future impact-related space weather anomalies and failures.

Shunting arc plasma source for pure carbon ion beam.

PubMed

Koguchi, H; Sakakita, H; Kiyama, S; Shimada, T; Sato, Y; Hirano, Y

2012-02-01

A plasma source is developed using a coaxial shunting arc plasma gun to extract a pure carbon ion beam. The pure carbon ion beam is a new type of deposition system for diamond and other carbon materials. Our plasma device generates pure carbon plasma from solid-state carbon material without using a hydrocarbon gas such as methane gas, and the plasma does not contain any hydrogen. The ion saturation current of the discharge measured by a double probe is about 0.2 mA∕mm(2) at the peak of the pulse.

Plasma chemistry for inorganic materials

NASA Technical Reports Server (NTRS)

Matsumoto, O.

1980-01-01

Practical application of plasma chemistry to the development of inorganic materials using both low temperature and warm plasmas are summarized. Topics cover: the surface nitrification and oxidation of metals; chemical vapor deposition; formation of minute oxide particles; the composition of oxides from chloride vapor; the composition of carbides and nitrides; freezing high temperature phases by plasma arc welding and plasma jet; use of plasma in the development of a substitute for petroleum; the production of silicon for use in solar cell batteries; and insulating the inner surface of nuclear fusion reactor walls.

Shunting arc plasma source for pure carbon ion beam

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koguchi, H.; Sakakita, H.; Kiyama, S.

2012-02-15

A plasma source is developed using a coaxial shunting arc plasma gun to extract a pure carbon ion beam. The pure carbon ion beam is a new type of deposition system for diamond and other carbon materials. Our plasma device generates pure carbon plasma from solid-state carbon material without using a hydrocarbon gas such as methane gas, and the plasma does not contain any hydrogen. The ion saturation current of the discharge measured by a double probe is about 0.2 mA/mm{sup 2} at the peak of the pulse.

Cutaneous plasmacytosis: A rare entity with unique presentation.

PubMed

Dhar, Subhra; Liani, Lalthleng; Patole, Kamlakar; Dhar, Sandipan

2017-01-01

Primary cutaneous plasmacytosis is a rare cutaneous disorder with extensive cutaneous plaques/papules mainly on the trunk and face. Cases have mostly been documented from Japan. We present here a rare case of cutaneous plasmacytosis from India of Mongolian descent. This 50-year-old female from Mizoram had extensive maculo-papular violaceous plaques distributed on the face, axillae, trunk and lower extremities. Initial and repeat skin biopsy revealed dense perivascular and periadnexal mature plasma cells. She also had lymphadenopathy. Serum protein electrophoresis did not reveal any M band and the Bence Jones protein was negative in urine. The patient had multiple superficial lymph nodes and a biopsy from the cervical lymph node showed effacement of normal nodal architecture by sheets of plasma cells. Immuno histochemistry was done from both skin and lymph node biopsies. The kappa and lambda tight chains were not restricted; there by proving the polyclonal nature of the plasma cells. The novelty of the case lies in its classical clinical presentation with histopathological documentation.

Scanning retarding field analyzer for plasma profile measurements in the boundary of the Alcator C-Mod tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brunner, D.; LaBombard, B.; Ochoukov, R.

2013-03-15

A new Retarding Field Analyzer (RFA) head has been created for the outer-midplane scanning probe system on the Alcator C-Mod tokamak. The new probe head contains back-to-back retarding field analyzers aligned with the local magnetic field. One faces 'upstream' into the field-aligned plasma flow and the other faces 'downstream' away from the flow. The RFA was created primarily to benchmark ion temperature measurements of an ion sensitive probe; it may also be used to interrogate electrons. However, its construction is robust enough to be used to measure ion and electron temperatures up to the last-closed flux surface in C-Mod. Amore » RFA probe of identical design has been attached to the side of a limiter to explore direct changes to the boundary plasma due to lower hybrid heating and current drive. Design of the high heat flux (>100 MW/m{sup 2}) handling probe and initial results are presented.« less

Hydrogen retention in Li and Li-C-O films

NASA Astrophysics Data System (ADS)

Buzi, Luxherta; Nelson, Andrew O.; Yang, Yuxin; Kaita, Robert; Koel, Bruce E.

2017-10-01

The efficiency of Li in binding H isotopes has led to reduced recycling in magnetic fusion devices and improved plasma performance. Since elemental Li surfaces are challenging to maintain in fusion devices due to the presence of impurities, parameterizing and understanding the mechanisms for H retention in various Li compounds (Li-C-O), in addition to pure Li, is crucial for Li plasma-facing material applications. To determine H retention in Li and Li-C-O films, measurements were done under ultrahigh vacuum conditions using temperature programmed desorption (TPD). Thin Li films (20 monolayers) were deposited on a nickel single crystal substrate and irradiated with 500 eV H2+ions at surface temperatures from 90K to 520K. Initial measurements on Li and Li-O films showed that the retention was comparable and dropped exponentially with surface temperature, from 95% at 90 K to 35% at 520 K. Auger electron spectroscopy and TPD showed that H was retained as lithium hydride (LiH) in pure Li and as lithium hydroxide (LiOH) in Li2O, which decomposed to H2O and Li2O at temperatures higher than 470K. H retention in Li-C and Li-C-O films will be determined over a similar temperature range, and the sputtering rate of these layers with H ions will also be reported. This material is based upon work supported by the U.S. Department of Energy, Office of Science/Fusion Energy Sciences under Award Number DE-SC0012890.

Development of laser-based techniques for in situ characterization of the first wall in ITER and future fusion devices

NASA Astrophysics Data System (ADS)

Philipps, V.; Malaquias, A.; Hakola, A.; Karhunen, J.; Maddaluno, G.; Almaviva, S.; Caneve, L.; Colao, F.; Fortuna, E.; Gasior, P.; Kubkowska, M.; Czarnecka, A.; Laan, M.; Lissovski, A.; Paris, P.; van der Meiden, H. J.; Petersson, P.; Rubel, M.; Huber, A.; Zlobinski, M.; Schweer, B.; Gierse, N.; Xiao, Q.; Sergienko, G.

2013-09-01

Analysis and understanding of wall erosion, material transport and fuel retention are among the most important tasks for ITER and future devices, since these questions determine largely the lifetime and availability of the fusion reactor. These data are also of extreme value to improve the understanding and validate the models of the in vessel build-up of the T inventory in ITER and future D-T devices. So far, research in these areas is largely supported by post-mortem analysis of wall tiles. However, access to samples will be very much restricted in the next-generation devices (such as ITER, JT-60SA, W7-X, etc) with actively cooled plasma-facing components (PFC) and increasing duty cycle. This has motivated the development of methods to measure the deposition of material and retention of plasma fuel on the walls of fusion devices in situ, without removal of PFC samples. For this purpose, laser-based methods are the most promising candidates. Their feasibility has been assessed in a cooperative undertaking in various European associations under EFDA coordination. Different laser techniques have been explored both under laboratory and tokamak conditions with the emphasis to develop a conceptual design for a laser-based wall diagnostic which is integrated into an ITER port plug, aiming to characterize in situ relevant parts of the inner wall, the upper region of the inner divertor, part of the dome and the upper X-point region.

Self-assembly of ordered nanostructures

NASA Astrophysics Data System (ADS)

Yin, Jinsong

2000-10-01

Several different kinds of nanostructure materials were studied in this thesis: self-assembled monodispersive nanocrystals, photonic crystals, ordered mesoporous silica and hierarchically ordered nanostructured materials. Tetrahedral nanocrystals of CoO, with edge-lengths of 4.4 +/- 0.2 nm, were synthesized at high purity and monodispersity. The size, shape and phase selections of the nanocrystals were performed using a novel magnetic field separation technique. These nanocrystals behave like molecules, forming a face-centered cubic self-assembly of nanocrystal superlattices. In-situ behavior of self-assembled CoO nanocrystal arrays was also analyzed using transmission electron microscopy and associated techniques. The surface passivation layer started to evaporate/decompose at temperatures as low as ˜200°C, but the exposed cores of nanocrystals preserved the geometrical configuration of the assembly due to the strong adhesion of the carbon substrate. As the temperature is further increased from 300 to 600°C, the intrinsic crystal structure of the CoO nanoparticles experiences a replacement reaction, resulting in the formation of cobalt carbides. Two-dimensional self-assembling of cobalt nanocrystals with an average particle size of 9.2 nm and polydispersity of 9% is processed. Phtonic crystals were processed by a template-assisted method. Ordered self-assembly of pores of titania nanocrystals formed a face-centered cubic packing structure. The walls of the pores were made of anatase nanocrystals of ˜8 nm in diameter. Cobalt can be doped into the walls of the pores by solution infiltration of cobalt carbonyl. Cobalt titanium oxide may be formed on the internal surface of the ordered pore structure. This type of structure is likely to be an excellent supporting material for catalysis. The experimental results suggest that transition metal elements can be incorporated into porous titania without blocking the interconnected pores. Hierarchically ordered nanostructured materials with high porosity at dual length-scale were prepared by a single annealing procedure. The plasma energy of this porous materials shifts about 1.2 eV to lower energy, compared to the fully densed silica spheres. This type of material is expected to have not only large surface area for catalysis, but also low dielectric constant for low-loss dielectric applications.

Structure and characteristics of functional powder composite materials obtained by spark plasma sintering

NASA Astrophysics Data System (ADS)

Oglezneva, S. A.; Kachenyuk, M. N.; Kulmeteva, V. B.; Ogleznev, N. B.

2017-07-01

The article describes the results of spark plasma sintering of ceramic materials based on titanium carbide, titanium carbosilicide, ceramic composite materials based on zirconium oxide, strengthened by carbon nanostructures and composite materials of electrotechnical purpose based on copper with addition of carbon structures and titanium carbosilicide. The research shows that the spark plasma sintering can achieve relative density of the material up to 98%. The effect of sintering temperature on the phase composition, density and porosity of the final product has been studied. It was found that with addition of carbon nanostructures the relative density and hardness decrease, but the fracture strength of ZrO2 increases up to times 2. The relative erosion resistance of the electrodes made of composite copper-based powder materials, obtained by spark plasma sintering during electroerosion treatment of tool steel exceeds that parameter of pure copper up to times 15.

Plasma characterization studies for materials processing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pfender, E.; Heberlein, J.

New applications for plasma processing of materials require a more detailed understanding of the fundamental processes occurring in the processing reactors. We have developed reactors offering specific advantages for materials processing, and we are using modeling and diagnostic techniques for the characterization of these reactors. The emphasis is in part set by the interest shown by industry pursuing specific plasma processing applications. In this paper we report on the modeling of radio frequency plasma reactors for use in materials synthesis, and on the characterization of the high rate diamond deposition process using liquid precursors. In the radio frequency plasma torchmore » model, the influence of specific design changes such as the location of the excitation coil on the enthalpy flow distribution is investigated for oxygen and air as plasma gases. The diamond deposition with liquid precursors has identified the efficient mass transport in form of liquid droplets into the boundary layer as responsible for high growth, and the chemical properties of the liquid for the film morphology.« less

Thermophysical properties of plasma sprayed coatings

NASA Technical Reports Server (NTRS)

Wilkes, K. E.; Lagedrost, J. F.

1973-01-01

Thermophysical properties of plasma sprayed materials were determined for the following plasma sprayed materials: CaO - stabilized ZrO2, Y2O3 - stabilized ZerO2, Al2O3, HfO2 Mo, nichrome, NiAl, Mo-ZrO2, and MoAl2O3 mixtures. In all cases the thermal conductivity of the as-sprayed materials was found to be considerably lower than that of the bulk material. The flash-laser thermal diffusivity technique was used both for diffusivity determination of single-layer materials and to determine the thermal contact resistance at the interface of two-layer specimens.

Method for depositing high-quality microcrystalline semiconductor materials

DOEpatents

Guha, Subhendu [Bloomfield Hills, MI; Yang, Chi C [Troy, MI; Yan, Baojie [Rochester Hills, MI

2011-03-08

A process for the plasma deposition of a layer of a microcrystalline semiconductor material is carried out by energizing a process gas which includes a precursor of the semiconductor material and a diluent with electromagnetic energy so as to create a plasma therefrom. The plasma deposits a layer of the microcrystalline semiconductor material onto the substrate. The concentration of the diluent in the process gas is varied as a function of the thickness of the layer of microcrystalline semiconductor material which has been deposited. Also disclosed is the use of the process for the preparation of an N-I-P type photovoltaic device.

How 'blended' is blended learning?: students' perceptions of issues around the integration of online and face-to-face learning in a Continuing Professional Development (CPD) health care context.

PubMed

Glogowska, Margaret; Young, Pat; Lockyer, Lesley; Moule, Pam

2011-11-01

This paper explores students' perceptions of blended learning modules delivered in a Continuing Professional Development (CPD) health care context in the UK. 'Blended learning' is the term used to describe a hybrid model of learning where traditional face-to-face teaching approaches and newer electronic learning activities and resources are utilised together. A new model of CPD for health care practitioners based on a blended learning approach was developed at a university in the south west of England. As part of the evaluation of the new modules, a qualitative study was conducted, in which 17 students who had experienced the modules were interviewed by telephone. Three main themes emerged from the interviews relating to the 'blended' nature of the blended learning modules. These were i) issues around the opportunities for discussion of online materials face-to-face; ii) issues of what material should be online versus face-to-face and iii) balancing online and face-to-face components. Teaching staff engaged in the development of blended learning courses need to pay particular attention to the ways in which they develop and integrate online and face-to-face materials. More attention needs to be paid to allowing opportunity for students to come together to create a 'community of inquiry'. Copyright © 2011 Elsevier Ltd. All rights reserved.

Determining the ion temperature and energy distribution in a lithium-plasma interaction test stand with a retarding field energy analyzer

NASA Astrophysics Data System (ADS)

Christenson, M.; Stemmley, S.; Jung, S.; Mettler, J.; Sang, X.; Martin, D.; Kalathiparambil, K.; Ruzic, D. N.

2017-08-01

The ThermoElectric-driven Liquid-metal plasma-facing Structures (TELS) experiment at the University of Illinois is a gas-puff driven, theta-pinch plasma source that is used as a test stand for off-normal plasma events incident on materials in the edge and divertor regions of a tokamak. The ion temperatures and resulting energy distributions are crucial for understanding how well a TELS pulse can simulate an extreme event in a larger, magnetic confinement device. A retarding field energy analyzer (RFEA) has been constructed for use with such a transient plasma due to its inexpensive and robust nature. The innovation surrounding the use of a control analyzer in conjunction with an actively sampling analyzer is presented and the conditions of RFEA operation are discussed, with results presented demonstrating successful performance under extreme conditions. Such extreme conditions are defined by heat fluxes on the order of 0.8 GW m-2 and on time scales of nearly 200 μs. Measurements from the RFEA indicate two primary features for a typical TELS discharge, following closely with the pre-ionizing coaxial gun discharge characteristics. For the case using the pre-ionization pulse (PiP) and the theta pinch, the measured ion signal showed an ion temperature of 23.3 ± 6.6 eV for the first peak and 17.6 ± 1.9 eV for the second peak. For the case using only the PiP, the measured signal showed an ion temperature of 7.9 ± 1.1 eV for the first peak and 6.6 ± 0.8 eV for the second peak. These differences illustrate the effectiveness of the theta pinch for imparting energy on the ions. This information also highlights the importance of TELS as being one of the few linear pulsed plasma sources whereby moderately energetic ions will strike targets without the need for sample biasing.

Determining the ion temperature and energy distribution in a lithium-plasma interaction test stand with a retarding field energy analyzer.

PubMed

Christenson, M; Stemmley, S; Jung, S; Mettler, J; Sang, X; Martin, D; Kalathiparambil, K; Ruzic, D N

2017-08-01

The ThermoElectric-driven Liquid-metal plasma-facing Structures (TELS) experiment at the University of Illinois is a gas-puff driven, theta-pinch plasma source that is used as a test stand for off-normal plasma events incident on materials in the edge and divertor regions of a tokamak. The ion temperatures and resulting energy distributions are crucial for understanding how well a TELS pulse can simulate an extreme event in a larger, magnetic confinement device. A retarding field energy analyzer (RFEA) has been constructed for use with such a transient plasma due to its inexpensive and robust nature. The innovation surrounding the use of a control analyzer in conjunction with an actively sampling analyzer is presented and the conditions of RFEA operation are discussed, with results presented demonstrating successful performance under extreme conditions. Such extreme conditions are defined by heat fluxes on the order of 0.8 GW m -2 and on time scales of nearly 200 μs. Measurements from the RFEA indicate two primary features for a typical TELS discharge, following closely with the pre-ionizing coaxial gun discharge characteristics. For the case using the pre-ionization pulse (PiP) and the theta pinch, the measured ion signal showed an ion temperature of 23.3 ± 6.6 eV for the first peak and 17.6 ± 1.9 eV for the second peak. For the case using only the PiP, the measured signal showed an ion temperature of 7.9 ± 1.1 eV for the first peak and 6.6 ± 0.8 eV for the second peak. These differences illustrate the effectiveness of the theta pinch for imparting energy on the ions. This information also highlights the importance of TELS as being one of the few linear pulsed plasma sources whereby moderately energetic ions will strike targets without the need for sample biasing.

PREFACE: Theory of Fusion Plasmas, 13th Joint Varenna-Lausanne International Workshop (2012)

NASA Astrophysics Data System (ADS)

Garbet, Xavier; Sauter, Olivier

2012-12-01

The 2012 joint Varenna-Lausanne international workshop on the theory of fusion plasmas has been very fruitful. A broad variety of topics were addressed, as usual covering turbulence, MHD, edge physic, RF wave heating and a taste of astrophysics. Moreover the scope of the meeting was extended this year to include the physics of materials and diagnostics for burning plasmas. This evolution reflects the complexity of problems at hand in fusion, in particular in the context of ITER construction. Long-standing problems without immediate consequences have sometimes become an urgent matter in that context. One may quote for instance the choice of plasma facing components or the design of control systems. Another characteristic of the meeting is the interplay between various domains of plasma physics. For instance MHD modes are now currently investigated with gyrokinetic codes, kinetic effects are more and more included in MHD stability analysis, and turbulence is now accounted for in wave propagation problems. This is the proof of cross-fertilization and it is certainly a healthy sign in our community. Finally introducing some novelty in the programme does not prevent us from respecting the traditions of the meeting. As usual a good deal of the presentations were dedicated to numerical simulations. Combining advanced numerical techniques with elaborated analytical theory is certainly a trademark of the Varenna-Lausanne conference, which was respected again this year. The quality and size of the scientific production is illustrated by the 26 papers which appear in the present volume of Journal of Physics: Conference Series, all refereed. We would also like to mention another set of 20 papers to be published in Plasma Physics and Controlled Fusion. We hope the readers will enjoy this special issue of JPCS and the one to come in PPCF. Xavier Garbet and Olivier Sauter October 26, 2012

Measurement of He neutral temperature in detached plasmas using laser absorption spectroscopy

NASA Astrophysics Data System (ADS)

Aramaki, M.; Tsujihara, T.; Kajita, S.; Tanaka, H.; Ohno, N.

2018-01-01

The reduction of the heat load onto plasma-facing components by plasma detachment is an inevitable scheme in future nuclear fusion reactors. Since the control of the plasma and neutral temperatures is a key issue to the detached plasma generation, we have developed a laser absorption spectroscopy system for the metastable helium temperature measurements and used together with a previously developed laser Thomson scattering system for the electron temperature and density measurements. The thermal relaxation process between the neutral and the electron in the detached plasma generated in the linear plasma device, NAGDIS-II was studied. It is shown that the electron temperature gets close to the neutral temperature by increasing the electron density. On the other hand, the pressure dependence of electron and neutral temperatures shows the cooling effect by the neutrals. The possibility of the plasma fluctuation measurement using the fluctuation in the absorption signal is also shown.

29 CFR 1915.153 - Eye and face protection.

Code of Federal Regulations, 2010 CFR

2010-07-01

... flying objects. Detachable side protectors (e.g., a clip-on or slide-on side shield) meeting the... (Light) Less than 10 Arc cutting (Heavy) 500 11 500-1000 Plasma arc welding Less than 6 20 8 20− 10 100 11 100− 400 400− 800 Plasma arc cutting (light)** Less than 300 8 (medium)** 300-400 9 (heavy)** 400...

29 CFR 1915.153 - Eye and face protection.

Code of Federal Regulations, 2013 CFR

2013-07-01

... flying objects. Detachable side protectors (e.g., a clip-on or slide-on side shield) meeting the... (Light) Less than 10 Arc cutting (Heavy) 500 11 500-1000 Plasma arc welding Less than 6 20 8 20− 10 100 11 100− 400 400− 800 Plasma arc cutting (light)** Less than 300 8 (medium)** 300-400 9 (heavy)** 400...

29 CFR 1915.153 - Eye and face protection.

Code of Federal Regulations, 2014 CFR

2014-07-01

... flying objects. Detachable side protectors (e.g., a clip-on or slide-on side shield) meeting the... (Light) Less than 10 Arc cutting (Heavy) 500 11 500-1000 Plasma arc welding Less than 6 20 8 20− 10 100 11 100− 400 400− 800 Plasma arc cutting (light)** Less than 300 8 (medium)** 300-400 9 (heavy)** 400...

29 CFR 1915.153 - Eye and face protection.

Code of Federal Regulations, 2012 CFR

2012-07-01

... flying objects. Detachable side protectors (e.g., a clip-on or slide-on side shield) meeting the... (Light) Less than 10 Arc cutting (Heavy) 500 11 500-1000 Plasma arc welding Less than 6 20 8 20− 10 100 11 100− 400 400− 800 Plasma arc cutting (light)** Less than 300 8 (medium)** 300-400 9 (heavy)** 400...

29 CFR 1915.153 - Eye and face protection.

Code of Federal Regulations, 2011 CFR

2011-07-01

... flying objects. Detachable side protectors (e.g., a clip-on or slide-on side shield) meeting the... (Light) Less than 10 Arc cutting (Heavy) 500 11 500-1000 Plasma arc welding Less than 6 20 8 20− 10 100 11 100− 400 400− 800 Plasma arc cutting (light)** Less than 300 8 (medium)** 300-400 9 (heavy)** 400...

BCA-kMC Hybrid Simulation for Hydrogen and Helium Implantation in Material under Plasma Irradiation

NASA Astrophysics Data System (ADS)

Kato, Shuichi; Ito, Atsushi; Sasao, Mamiko; Nakamura, Hiroaki; Wada, Motoi

2015-09-01

Ion implantation by plasma irradiation into materials achieves the very high concentration of impurity. The high concentration of impurity causes the deformation and the destruction of the material. This is the peculiar phenomena in the plasma-material interaction (PMI). The injection process of plasma particles are generally simulated by using the binary collision approximation (BCA) and the molecular dynamics (MD), while the diffusion of implanted atoms have been traditionally solved by the diffusion equation, in which the implanted atoms is replaced by the continuous concentration field. However, the diffusion equation has insufficient accuracy in the case of low concentration, and in the case of local high concentration such as the hydrogen blistering and the helium bubble. The above problem is overcome by kinetic Monte Carlo (kMC) which represents the diffusion of the implanted atoms as jumps on interstitial sites in a material. In this paper, we propose the new approach ``BCA-kMC hybrid simulation'' for the hydrogen and helium implantation under the plasma irradiation.

Deuterium retention and surface modification of tungsten macrobrush samples exposed in FTU Tokamak

NASA Astrophysics Data System (ADS)

Maddaluno, G.; Giacomi, G.; Rufoloni, A.; Verdini, L.

2007-06-01

The effect of discrete structures such as macrobrush or castellated surfaces on power handling and deuterium retention of plasma facing components is to be assessed since such geometrical configurations are needed for increasing the lifetime of the armour to heat-sink joint. Four small macrobrush W and W + 1%La2O3 samples have been exposed in the Frascati Tokamak Upgrade (FTU) scrape-off layer up to the last closed flux surface by means of the Sample Introduction System. FTU is an all metal machine with no carbon source inside vacuum vessel; it exhibits ITER relevant energy and particle fluxes on the plasma facing components. Here, results on morphological surface changes (SEM), chemical composition (EDX) and deuterium retention (TDS) are reported.

Measurement of the surface morphology of plasma facing components on the EAST tokamak by a laser speckle interferometry approach

NASA Astrophysics Data System (ADS)

Hongbei, WANG; Xiaoqian, CUI; Yuanbo, LI; Mengge, ZHAO; Shuhua, LI; Guangnan, LUO; Hongbin, DING

2018-03-01

The laser speckle interferometry approach provides the possibility of an in situ optical non-contacted measurement for the surface morphology of plasma facing components (PFCs), and the reconstruction image of the PFC surface morphology is computed by a numerical model based on a phase unwrapping algorithm. A remote speckle interferometry measurement at a distance of three meters for real divertor tiles retired from EAST was carried out in the laboratory to simulate a real detection condition on EAST. The preliminary surface morphology of the divertor tiles was well reproduced by the reconstructed geometric image. The feasibility and reliability of this approach for the real-time measurement of PFCs have been demonstrated.

Autologous Platelet-Poor Plasma Gel for Injection Laryngoplasty

PubMed Central

Woo, Seung Hoon; Kim, Jin Pyeong; Park, Jung Je; Chung, Phil-Sang

2013-01-01

Purpose To overcome the potential disadvantages of the use of foreign materials and autologous fat or collagen, we introduce here an autologous plasma gel for injection laryngoplasty. The purpose of this study was to present a new injection material, a plasma gel, and to discuss its clinical effectiveness. Materials and Methods From 2 mL of blood, the platelet poor serum layer was collected and heated at 100℃ for 12 min to form a plasma gel. The plasma gel was then injected into a targeted site; the safety and efficacy thereof were evaluated in 30 rats. We also conducted a phase I/II clinical study of plasma gel injection laryngoplasty in 11 unilateral vocal fold paralysis patients. Results The plasma gel was semi-solid and an easily injectable material. Of note, plasma gel maintains the same consistency for up to 1 year in a sealed bottle. However, exposure to room air causes the plasma gel to disappear within 1 month. In our animal study, the autologous plasma gel remained in situ for 6 months in animals with minimal inflammation. Clinical study showed that vocal cord palsy was well compensated for with the plasma gel in all patients at two months after injection with no significant complications. Jitter, shimmer, maximum, maximum phonation time (MPT) and mean voice handicap index (VHI) also improved significantly after plasma gel injection. However, because the injected plasma gel was gradually absorbed, 6 patients needed another injection, while the gel remained in place in 2 patients. Conclusion Injection laryngoplasty with autologous plasma gel may be a useful and safe treatment option for temporary vocal cord palsy. PMID:24142660

Liquid injection plasma deposition method and apparatus

DOEpatents

Kong, Peter C.; Watkins, Arthur D.

1999-01-01

A liquid injection plasma torch deposition apparatus for depositing material onto a surface of a substrate may comprise a plasma torch for producing a jet of plasma from an outlet nozzle. A plasma confinement tube having an inlet end and an outlet end and a central bore therethrough is aligned with the outlet nozzle of the plasma torch so that the plasma jet is directed into the inlet end of the plasma confinement tube and emerges from the outlet end of the plasma confinement tube. The plasma confinement tube also includes an injection port transverse to the central bore. A liquid injection device connected to the injection port of the plasma confinement tube injects a liquid reactant mixture containing the material to be deposited onto the surface of the substrate through the injection port and into the central bore of the plasma confinement tube.

Review on plasmas in extraordinary media: plasmas in cryogenic conditions and plasmas in supercritical fluids

NASA Astrophysics Data System (ADS)

Stauss, Sven; Muneoka, Hitoshi; Terashima, Kazuo

2018-02-01

Plasma science and technology has enabled advances in very diverse fields: micro- and nanotechnology, chemical synthesis, materials fabrication and, more recently, biotechnology and medicine. While many of the currently employed plasma tools and technologies are very advanced, the types of plasmas used in micro- and nanofabrication pose certain limits, for example, in treating heat-sensitive materials in plasma biotechnology and plasma medicine. Moreover, many physical properties of plasmas encountered in nature, and especially outer space, i.e. very-low-temperature plasmas or plasmas that occur in high-density media, are not very well understood. The present review gives a short account of laboratory plasmas generated under ’extreme’ conditions: at cryogenic temperatures and in supercritical fluids. The fundamental characteristics of these cryogenic plasmas and cryoplasmas, and plasmas in supercritical fluids, especially supercritical fluid plasmas, are presented with their main applications. The research on such exotic plasmas is expected to lead to further understanding of plasma physics and, at the same time, enable new applications in various technological fields.

Material for electrodes of low temperature plasma generators

DOEpatents

Caplan, Malcolm; Vinogradov, Sergel Evge'evich; Ribin, Valeri Vasil'evich; Shekalov, Valentin Ivanovich; Rutberg, Philip Grigor'evich; Safronov, Alexi Anatol'evich

2008-12-09

Material for electrodes of low temperature plasma generators. The material contains a porous metal matrix impregnated with a material emitting electrons. The material uses a mixture of copper and iron powders as a porous metal matrix and a Group IIIB metal component such as Y.sub.2O.sub.3 is used as a material emitting electrons at, for example, the proportion of the components, mass %: iron: 3-30; Y.sub.2O.sub.3:0.05-1; copper: the remainder. Copper provides a high level of heat conduction and electric conductance, iron decreases intensity of copper evaporation in the process of plasma creation providing increased strength and lifetime, Y.sub.2O.sub.3 provides decreasing of electronic work function and stability of arc burning. The material can be used for producing the electrodes of low temperature AC plasma generators used for destruction of liquid organic wastes, medical wastes, and municipal wastes as well as for decontamination of low level radioactive waste, the destruction of chemical weapons, warfare toxic agents, etc.

Material for electrodes of low temperature plasma generators

DOEpatents

Caplan, Malcolm; Vinogradov, Sergel Evge'evich; Ribin, Valeri Vasil'evich; Shekalov, Valentin Ivanovich; Rutberg, Philip Grigor'evich; Safronov, Alexi Anatol'evich; Shiryaev, Vasili Nikolaevich

2010-03-02

Material for electrodes of low temperature plasma generators. The material contains a porous metal matrix impregnated with a material emitting electrons. The material uses a mixture of copper and iron powders as a porous metal matrix and a Group IIIB metal component such as Y.sub.2O.sub.3 is used as a material emitting electrons at, for example, the proportion of the components, mass %: iron:3-30; Y.sub.2O.sub.3:0.05-1; copper: the remainder. Copper provides a high level of heat conduction and electric conductance, iron decreases intensity of copper evaporation in the process of plasma creation providing increased strength and lifetime, Y.sub.2O.sub.3 provides decreasing of electronic work function and stability of arc burning. The material can be used for producing the electrodes of low temperature AC plasma generators used for destruction of liquid organic wastes, medical wastes, municipal wastes as well as for decontamination of low level radioactive waste, the destruction of chemical weapons, warfare toxic agents, etc.

Investigation of some process parameters using microwave plasma technology for the treatment of radioactive waste

NASA Astrophysics Data System (ADS)

Trnovcevic, J.; Schneider, F.; Scherer, U. W.

2017-02-01

The production of nuclear energy and the application of other nuclear technologies produce large volumes of low- and intermediate-level radioactive wastes. To investigate a novel means of treating such wastes, plasma is investigated for its efficacy. Plasma treatment promises to simultaneously treat all waste types without any previous sorting or pre-treatment. Microwave-driven plasma torches have the advantage of high-energy efficiency and low-electrode wear. In small-scale experiments, several design variations of an open plasma oven were assembled in order to investigate constraints caused by the materials and oven geometry. The experimental set-up was modified several times in order to test the design characteristics and the variation of plasma-specific proprieties related to the radioactive waste treatment and in order to find a suitable solution with the minimum complexity that allows a representative reproducibility of the results obtained. A plasma torch controlled by a 2.45 GHz microwave signal of up to 200 W was used, employing air as the primary plasma gas with a flow rate of ∼2 L/min. Different organic and inorganic materials in different shapes and sizes were treated besides a standardized mixture resembling mixed wastes from nuclear plants. The results prove that the chosen microwave plasma torch is suitable for a combined combustion and melting of organic and in-organic materials. Investigation of the specimen size to be treated is influential in this process: the power is still too low to melt larger samples, but the temperature is sufficient to treat all kinds of material. When glass particles are added, materials melt together to form an amorphous substance, proving the possibility to vitrify material with this plasma torch. By optimization of the oven configuration, the time needed to combust 25 g of standard sample was reduced by ∼50%. Typical energy efficiencies were found in the range of 8-20% for melting of metal chipping, and ∼90% for melting of zinc powder.

Methods and system for controlled laser-driven explosive bonding

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rubenchik, Alexander M.; Farmer, Joseph C.; Hackel, Lloyd

A technique for bonding two dissimilar materials includes positioning a second material over a first material at an oblique angle and applying a tamping layer over the second martial. A laser beam is directed at the second material that generates a plasma at the location of impact on the second material. The plasma generates pressure that accelerates a portion of the second material to a very high velocity and towards the first material. The second material impacts the first material causing bonding of the two materials.

Entire plasmas can be restructured when electrons are emitted from the boundaries

DOE PAGES

Campanell, M. D.

2015-04-14

It is well known that electron emission can restructure the thin sheaths at plasma-facing surfaces. But conventional models assume that the plasma's structure negligibly changes (the “presheath” is still thought to be governed by ion acceleration to the Bohm speed). Here, it is shown by theory and simulation that the presheath can take a fundamentally different structure where the emitted electrons entering the quasineutral region cause numerous changes. As a result, gradients of total plasma density, ion and electron pressures, and electric potential throughout the “inverted” presheath can carry different magnitudes, and opposite signs, from Bohm presheaths.

Ti film deposition process of a plasma focus: Study by an experimental design

NASA Astrophysics Data System (ADS)

Inestrosa-Izurieta, M. J.; Moreno, J.; Davis, S.; Soto, L.

2017-10-01

The plasma generated by plasma focus (PF) devices have substantially different physical characteristics from another plasma, energetic ions and electrons, compared with conventional plasma devices used for plasma nanofabrication, offering new and unique opportunities in the processing and synthesis of Nanomaterials. This article presents the use of a plasma focus of tens of joules, PF-50J, for the deposition of materials sprayed from the anode by the plasma dynamics in the axial direction. This work focuses on the determination of the most significant effects of the technological parameters of the system on the obtained depositions through the use of a statistical experimental design. The results allow us to give a qualitative understanding of the Ti film deposition process in our PF device depending on four different events provoked by the plasma dynamics: i) an electric erosion of the outer material of the anode; ii) substrate ablation generating an interlayer; iii) electron beam deposition of material from the center of the anode; iv) heat load provoking clustering or even melting of the deposition surface.

Selective Plasma Etching of Polymeric Substrates for Advanced Applications

PubMed Central

Puliyalil, Harinarayanan; Cvelbar, Uroš

2016-01-01

In today’s nanoworld, there is a strong need to manipulate and process materials on an atom-by-atom scale with new tools such as reactive plasma, which in some states enables high selectivity of interaction between plasma species and materials. These interactions first involve preferential interactions with precise bonds in materials and later cause etching. This typically occurs based on material stability, which leads to preferential etching of one material over other. This process is especially interesting for polymeric substrates with increasing complexity and a “zoo” of bonds, which are used in numerous applications. In this comprehensive summary, we encompass the complete selective etching of polymers and polymer matrix micro-/nanocomposites with plasma and unravel the mechanisms behind the scenes, which ultimately leads to the enhancement of surface properties and device performance. PMID:28335238

Early Career. Harnessing nanotechnology for fusion plasma-material interface research in an in-situ particle-surface interaction facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Allain, Jean Paul

2014-08-08

This project consisted of fundamental and applied research of advanced in-situ particle-beam interactions with surfaces/interfaces to discover novel materials able to tolerate intense conditions at the plasma-material interface (PMI) in future fusion burning plasma devices. The project established a novel facility that is capable of not only characterizing new fusion nanomaterials but, more importantly probing and manipulating materials at the nanoscale while performing subsequent single-effect in-situ testing of their performance under simulated environments in fusion PMI.

A review of low density porous materials used in laser plasma experiments

NASA Astrophysics Data System (ADS)

Nagai, Keiji; Musgrave, Christopher S. A.; Nazarov, Wigen

2018-03-01

This review describes and categorizes the synthesis and properties of low density porous materials, which are commonly referred to as foams and are utilized for laser plasma experiments. By focusing a high-power laser on a small target composed of these materials, high energy and density states can be produced. In the past decade or so, various new target fabrication techniques have been developed by many laboratories that use high energy lasers and consequently, many publications and reviews followed these developments. However, the emphasis so far has been on targets that did not utilize low density porous materials. This review therefore, attempts to redress this balance and endeavors to review low density materials used in laser plasma experiments in recent years. The emphasis of this review will be on aspects of low density materials that are of relevance to high energy laser plasma experiments. Aspects of low density materials such as densities, elemental compositions, macroscopic structures, nanostructures, and characterization of these materials will be covered. Also, there will be a brief mention of how these aspects affect the results in laser plasma experiments and the constrictions that these requirements put on the fabrication of low density materials relevant to this field. This review is written from the chemists' point of view to aid physicists and the new comers to this field.

21 CFR 640.100 - Immune Globulin (Human).

Code of Federal Regulations, 2010 CFR

2010-04-01

... (Human). The product is defined as a sterile solution containing antibodies derived from human plasma. (b) Source material. The source material of Immune Globulin (Human) shall be plasma recovered from Whole Blood prepared as prescribed in §§ 640.1 through 640.5, or Source Plasma prepared as prescribed in...

21 CFR 640.100 - Immune Globulin (Human).

Code of Federal Regulations, 2012 CFR

2012-04-01

... (Human). The product is defined as a sterile solution containing antibodies derived from human plasma. (b) Source material. The source material of Immune Globulin (Human) shall be plasma recovered from Whole Blood prepared as prescribed in §§ 640.1 through 640.5, or Source Plasma prepared as prescribed in...

21 CFR 640.100 - Immune Globulin (Human).

Code of Federal Regulations, 2011 CFR

2011-04-01

... (Human). The product is defined as a sterile solution containing antibodies derived from human plasma. (b) Source material. The source material of Immune Globulin (Human) shall be plasma recovered from Whole Blood prepared as prescribed in §§ 640.1 through 640.5, or Source Plasma prepared as prescribed in...

21 CFR 640.100 - Immune Globulin (Human).

Code of Federal Regulations, 2013 CFR

2013-04-01

... (Human). The product is defined as a sterile solution containing antibodies derived from human plasma. (b) Source material. The source material of Immune Globulin (Human) shall be plasma recovered from Whole Blood prepared as prescribed in §§ 640.1 through 640.5, or Source Plasma prepared as prescribed in...

21 CFR 640.100 - Immune Globulin (Human).

Code of Federal Regulations, 2014 CFR

2014-04-01

... (Human). The product is defined as a sterile solution containing antibodies derived from human plasma. (b) Source material. The source material of Immune Globulin (Human) shall be plasma recovered from Whole Blood prepared as prescribed in §§ 640.1 through 640.5, or Source Plasma prepared as prescribed in...

Chapter 8: Plasma operation and control

NASA Astrophysics Data System (ADS)

ITER Physics Expert Group on Disruptions, Control, Plasma, and MHD; ITER Physics Expert Group on Energetic Particles, Heating, Current and Drive; ITER Physics Expert Group on Diagnostics; ITER Physics Basis Editors

1999-12-01

Wall conditioning of fusion devices involves removal of desorbable hydrogen isotopes and impurities from interior device surfaces to permit reliable plasma operation. Techniques used in present devices include baking, metal film gettering, deposition of thin films of low-Z material, pulse discharge cleaning, glow discharge cleaning, radio frequency discharge cleaning, and in situ limiter and divertor pumping. Although wall conditioning techniques have become increasingly sophisticated, a reactor scale facility will involve significant new challenges, including the development of techniques applicable in the presence of a magnetic field and of methods for efficient removal of tritium incorporated into co-deposited layers on plasma facing components and their support structures. The current status of various approaches is reviewed, and the implications for reactor scale devices are summarized. Creation and magnetic control of shaped and vertically unstable elongated plasmas have been mastered in many present tokamaks. The physics of equilibrium control for reactor scale plasmas will rely on the same principles, but will face additional challenges, exemplified by the ITER/FDR design. The absolute positioning of outermost flux surface and divertor strike points will have to be precise and reliable in view of the high heat fluxes at the separatrix. Long pulses will require minimal control actions, to reduce accumulation of AC losses in superconducting PF and TF coils. To this end, more complex feedback controllers are envisaged, and the experimental validation of the plasma equilibrium response models on which such controllers are designed is encouraging. Present simulation codes provide an adequate platform on which equilibrium response techniques can be validated. Burning plasmas require kinetic control in addition to traditional magnetic shape and position control. Kinetic control refers to measures controlling density, rotation and temperature in the plasma core as well as in plasma periphery and divertor. The planned diagnostics (Chapter 7) serve as sensors for kinetic control, while gas and pellet fuelling, auxiliary power and angular momentum input, impurity injection, and non-inductive current drive constitute the control actuators. For example, in an ignited plasma, core density controls fusion power output. Kinetic control algorithms vary according to the plasma state, e.g. H- or L-mode. Generally, present facilities have demonstrated the kinetic control methods required for a reactor scale device. Plasma initiation - breakdown, burnthrough and initial current ramp - in reactor scale tokamaks will not involve physics differing from that found in present day devices. For ITER, the induced electric field in the chamber will be ~0.3V· m-1 - comparable to that required by breakdown theory but somewhat smaller than in present devices. Thus, a start-up 3MW electron cyclotron heating system will be employed to assure burnthrough. Simulations show that plasma current ramp up and termination in a reactor scale device can follow procedures developed to avoid disruption in present devices. In particular, simulations remain in the stable area of the li-q plane. For design purposes, the resistive V·s consumed during initiation is found, by experiments, to follow the Ejima expression, 0.45μ0 RIp. Advanced tokamak control has two distinct goals. First, control of density, auxiliary power, and inductive current ramping to attain reverse shear q profiles and internal transport barriers, which persist until dissipated by magnetic flux diffusion. Such internal transport barriers can lead to transient ignition. Second, combined use poloidal field shape control with non-inductive current drive and NBI angular momentum injection to create and control steady state, high bootstrap fraction, reverse shear discharges. Active n = 1 magnetic feedback and/or driven rotation will be required to suppress resistive wall modes for steady state plasmas that must operate in the wall stabilized regime for reactor levels of β >= 0.03.

Fine structure of synapses of the central nervous system in resinless sections.

PubMed

Cohen, R S; Wolosewick, J J; Becker, R P; Pappas, G D

1983-10-01

The cytoskeleton has been implicated in neuronal function, particularly in axonal transport, excitability at axonal membranes, and movement of synaptic vesicles at preganglionic endings. The present study demonstrates the presence of a pre- and postsynaptic cytoskeleton in resinless sections of CNS tissue by use of the polyethylene glycol (PEG) technique of Wolosewick (1980) viewed by conventional transmission EM, scanning transmission EM, and surface scanning EM. The PEG technique permits visualization of the cytoskeletal network unobscured by the electron scattering properties of epoxy embedment. In the presynaptic process, synaptic vesicles appear to be suspended in a filamentous network that is contiguous with the synaptic vesicle membrane and with the presynaptic plasma membrane and its dense material. In the postsynaptic process, the postsynaptic density (PSD) is seen in intimate contact with the postsynaptic membrane. En face images of the PSD in some synapses appear as a torus. Emanating from the filamentous web of the PSD are filaments which extend to the adjacent plasma membrane. We conclude that membranous synaptic elements are contiguous with a three-dimensional lattice network that is similar to that described in whole unembedded cells (Wolosewick and Porter, 1976). Moreover, the synaptic densities represent a specialized elaboration of the cytoskeleton.

Plasma nitriding induced growth of Pt-nanowire arrays as high performance electrocatalysts for fuel cells

NASA Astrophysics Data System (ADS)

Du, Shangfeng; Lin, Kaijie; Malladi, Sairam K.; Lu, Yaxiang; Sun, Shuhui; Xu, Qiang; Steinberger-Wilckens, Robert; Dong, Hanshan

2014-09-01

In this work, we demonstrate an innovative approach, combing a novel active screen plasma (ASP) technique with green chemical synthesis, for a direct fabrication of uniform Pt nanowire arrays on large-area supports. The ASP treatment enables in-situ N-doping and surface modification to the support surface, significantly promoting the uniform growth of tiny Pt nuclei which directs the growth of ultrathin single-crystal Pt nanowire (2.5-3 nm in diameter) arrays, forming a three-dimensional (3D) nano-architecture. Pt nanowire arrays in-situ grown on the large-area gas diffusion layer (GDL) (5 cm2) can be directly used as the catalyst electrode in fuel cells. The unique design brings in an extremely thin electrocatalyst layer, facilitating the charge transfer and mass transfer properties, leading to over two times higher power density than the conventional Pt nanoparticle catalyst electrode in real fuel cell environment. Due to the similar challenges faced with other nanostructures and the high availability of ASP for other material surfaces, this work will provide valuable insights and guidance towards the development of other new nano-architectures for various practical applications.