Plasma Sprayed Tungsten-based Coatings and their Usage in Edge Plasma Region of Tokamaks

Czech Academy of Sciences Publication Activity Database

Matějíček, Jiří; Weinzettl, Vladimír; Dufková, Edita; Piffl, Vojtěch; Peřina, Vratislav

2006-01-01

Roč. 51, č. 2 (2006), s. 179-191 ISSN 0001-7043 Grant - others:Evropská unie EFDA Task TW-5-TVM-PSW (EU – Euratom) Institutional research plan: CEZ:AV0Z20430508; CEZ:AV0Z10480505 Keywords : plasma sprayed coatings * fusion * plasma facing components * tungsten * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics

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

International Nuclear Information System (INIS)

Hammond, Karl D.; Wirth, Brian D.

2014-01-01

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 below 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. 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.

Development of bonding techniques between tungsten and copper alloy for plasma facing components by HIP method. 1. Bonding between tungsten and oxygen free copper

International Nuclear Information System (INIS)

Saito, Shigeru; Fukaya, Kiyoshi; Ishiyama, Shintaro; Eto, Motokuni; Akiba, Masato

1999-08-01

In recent years, it has been considered that W (tungsten) is one of candidate materials for armor tiles of plasma facing components, like first wall or divertor, of fusion reactor. On the other hand, oxygen free high thermal conductivity (OFHC)-copper is proposed as heat sink materials behind the plasma facing materials because of its high thermal conductivity. However, plasma facing components are exposed to cyclic high heat load and heavily irradiated by 14 MeV neutron. Under these conditions, many unfavorable effects, for instance, thermal stresses of bonding interface, irradiation damage and He atom production by nuclear transmutation, will be decreased bonding strength between W and Cu alloys. Therefore, it is necessary to develop a reliable bonding techniques in order to make plasma facing components which can resist them. Then, we started the bonding technology development by hot isostatic press (HIP) method to bond W with Cu alloys. In this experiments, to optimize HIP bonding conditions, four point bending were performed for each bonded conditions at temperature from R.T. to 873 K and we could get the best HIP bonding conditions for W and OFHC-Cu as 1273 K x 2 hours x 147 MPa. To evaluate bonding strength of the specimen bonded at these conditions, tensile tests were also performed at same temperature range. The tensile strength was similar with OFHC-Cu which were treated at same conditions. (author)

Numerical simulation of the bubble growth due to hydrogen isotopes inventory processes in plasma-irradiated tungsten

International Nuclear Information System (INIS)

Sang, Chaofeng; Sun, Jizhong; Bonnin, Xavier; Liu, Shengguang; Wang, Dezhen

2013-01-01

Hydrogen isotopes (HI) inventory is a key issue for fusion devices like ITER. It is especially urgent to understand how HI are retained in tungsten since it currently is the most important candidate material for the plasma-facing wall. Bubble growth is an important experimental complication that yet prevents a full understanding of HI retention processes in tungsten walls and most critically the divertor elements. In this work, we develop a model based on rate equations, which includes the bubble growth in tungsten being exposed to a HI plasma. In the model, HI molecules can be produced through recombination processes on the inner surface of a bubble, and HI molecules can also dissociate themselves to solute atoms, and the latter diffuse into the bulk wall because of very high pressures inside the bubble. The present model is applied to simulate how HI are retained in plasma-irradiated tungsten in the form of molecules to explain the wall temperature, trap concentration, incident HI flux and fluence dependencies of bubble growth

Optimization of tungsten-steel joints for plasma facing components in fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Heuer, Simon; Linsmeier, Christian [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik, Juelich (Germany); Weber, Thomas; Linke, Jochen [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Werkstoffstruktur und -eigenschaften, Juelich (Germany); Matejicek, Jiri [Institute of Plasma Physics, Academy of Sciences of the Czech Republic, Prague (Czech Republic)

2015-07-01

Tungsten, joint to a martensitic-ferritic EUROFER97 structure, is a promising plasma facing material composite for fusion reactors. Due to the effect of mismatch in thermo-mechanical properties direct bonding is not feasible. Current research is therefore ongoing on interlayer systems. While the adhesion was already improved by the utilization of a discrete Cu, Ti or V interlayer, that is able to relax stresses by plastic deformation, joints still do not resist the expected load cycles in a fusion reactor. Therefore, alternatives for the interface are needed. This contribution presents research on functionally graded materials (FGM). The particular microstructure of a graded interlayer allows re-distributing macro stresses from a discrete interface to a greater volume while avoiding in particular Cu which tends to swell under neutron irradiation. A parameter study on the basis of finite element analysis will be presented as well as first results of several processing routes for FGM that shall be evaluated and benchmarked by mechanical as well as thermal testing.

Elastic–plastic adhesive impacts of tungsten dust with metal surfaces in plasma environments

Energy Technology Data Exchange (ETDEWEB)

Ratynskaia, S., E-mail: svetlana.ratynskaia@ee.kth.se [KTH Royal Institute of Technology, Association EUROfusion-VR, Stockholm (Sweden); Tolias, P. [KTH Royal Institute of Technology, Association EUROfusion-VR, Stockholm (Sweden); Shalpegin, A. [Université de Lorraine, Institut Jean Lamour, Vandoeuvre-lès-Nancy (France); Vignitchouk, L. [KTH Royal Institute of Technology, Association EUROfusion-VR, Stockholm (Sweden); De Angeli, M. [Istituto di Fisica del Plasma – Consiglio Nazionale delle Ricerche, Milan (Italy); Bykov, I. [KTH Royal Institute of Technology, Association EUROfusion-VR, Stockholm (Sweden); Bystrov, K.; Bardin, S. [FOM Institute DIFFER, Dutch Institute For Fundamental Energy Research, Edisonbaan 14, 3439MN Nieuwegein (Netherlands); Brochard, F. [Université de Lorraine, Institut Jean Lamour, Vandoeuvre-lès-Nancy (France); Ripamonti, D. [Istituto per l’Energetica e le Interfasi – Consiglio Nazionale delle Ricerche, Milan (Italy); Harder, N. den; De Temmerman, G. [FOM Institute DIFFER, Dutch Institute For Fundamental Energy Research, Edisonbaan 14, 3439MN Nieuwegein (Netherlands)

2015-08-15

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.

Hydrogen transport behavior of metal coatings for plasma-facing components

Energy Technology Data Exchange (ETDEWEB)

Anderl, R.A.; Holland, D.F.; Longhurst, G.R. (Idaho National Engineering Lab., Idaho Falls (USA))

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{sub 3}{sup +} 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 of tungsten, copper, and tungsten-coated copper specimens at temperatures ranging from 638 to 825 K and implanting particle fluxes of approximately 5x10{sup 19} D/m{sup 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. (orig.).

Hydrogen transport behavior of metal coatings for plasma-facing components

Science.gov (United States)

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.

Hydrogen transport behavior of metal coatings for plasma facing components

International Nuclear Information System (INIS)

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

1990-01-01

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 K to 825 K and implanting particle fluxes of approximately 5 x 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. 18 refs., 3 figs., 3 tabs

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

Science.gov (United States)

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.

Development of novel tungsten processing technologies for electro-chemical machining (ECM) of plasma facing components

International Nuclear Information System (INIS)

Holstein, Nils; Krauss, Wolfgang; Konys, Juergen

2011-01-01

Plasma facing components for fusion applications must exhibit long-term stability under extreme conditions, and therefore material imperfections cannot be tolerated due to a high risk of technical failures. To prevent or abolish defects in refractory metals components during the manufacturing process, some methods of electro-chemical machining as S-ECM and C-ECM were developed, enabling both the processing of smooth plain defect-free surfaces of different geometry and the removal of bulk material for the shaping of three-dimensional structures, also without cracks. It is discussed, that tungsten ablation with accurate electro-chemical molding is very sensitive to the kind of electric current, and therefore current investigations focused also on the effects of frequency profiles on the sharpness of edge rounding.

The WEST programme: Minimizing technology and operational risks of a full actively cooled tungsten divertor on ITER

Energy Technology Data Exchange (ETDEWEB)

Grosman, André, E-mail: andre.grosman@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Bucalossi, Jérôme; Doceul, Louis [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Escourbiac, Frédéric [ITER Organization, Cadarache, 13115 St. Paul-lez-Durance (France); Lipa, Manfred [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Merola, Mario [ITER Organization, Cadarache, 13115 St. Paul-lez-Durance (France); Missirlian, Marc [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Pitts, Richard A. [ITER Organization, Cadarache, 13115 St. Paul-lez-Durance (France); Samaille, Franck; Tsitrone, Emmanuelle [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France)

2013-10-15

Highlights: ► The WEST programme is a unique opportunity to experience the industrial scale manufacture of tungsten plasma-facing components similar to the ITER divertor ones. ► In Tore Supra, it will bring important know how for actively cooled W divertor operation. ► This can be done by a reasonable modification of the Tore Supra tokamak. ► A fast implementation of the project would make this information available in due time. ► This allows a significant contribution to the W ITER divertor risk minimization in its manufacturing and operation phase. -- Abstract: The WEST programme consists in transforming the Tore Supra tokamak into an X point divertor device, while taking advantage of its long discharge capability. This is obtained by inserting in vessel coils to create the X point while adapting the in-vessel elements to this new geometry. This will allow the full tungsten divertor technology to be used on ITER to be tested in anticipation of its use on ITER under relevant heat loading conditions and pulse duration. The early manufacturing of a significant industrial series of ITER-similar W plasma-facing units will contribute to the ITER divertor manufacturing risk mitigation and to that associated with early W divertor plasma operation on ITER.

Plasma etching of patterned tungsten

International Nuclear Information System (INIS)

Franssila, S.

1993-01-01

Plasma etching of tungsten is discussed from the viewpoint of thin film structure and integrated circuit process engineering. The emphasis is on patterned tungsten etching for silicon device and X-ray mask fabrication. After introducing tungsten etch chemistries and mechanisms, microstructural aspects of tungsten films (crystal structure, grain size, film density, defects, impurities) in relation to etching are discussed. Approaches to etch process optimization are presented, and the current state-of-the-art of patterned tungsten etching is reviewed. (orig.)

An equilibrium model for tungsten fuzz in an eroding plasma environment

International Nuclear Information System (INIS)

Doerner, R.P.; Baldwin, M.J.; Stangeby, P.C.

2011-01-01

A model equating the growth rate of tungsten fuzz on a plasma-exposed surface to the erosion rate of the fuzzy surface is developed to predict the likelihood of tungsten fuzz formation in the steady-state environment of toroidal confinement devices. To date this question has not been answered because the operational conditions in existing magnetic confinement machines do not necessarily replicate those expected in future fusion reactors (i.e. high-fluence operation, high temperature plasma-facing materials and edge plasma relatively free of condensable impurities). The model developed is validated by performing plasma exposure experiments at different incident ion energies (thereby varying the erosion rate) and measuring the resultant fuzz layer thickness. The results indicate that if the conditions exist for fuzz development in a steady-state plasma (surface temperature and energetic helium flux), then the erosion rate will determine the equilibrium thickness of the surface fuzz layer.

Materials for the plasma-facing components of steady state stellarators

International Nuclear Information System (INIS)

Bolt, H.; Boscary, J.; Greuner, H.; Grigull, P.; Maier, H.; Streibl, B.

2005-01-01

The specific advantage of current-free stellarators is their inherent capability for full steady-state operation. This will lead to long discharges and the corresponding stationary plasma exposure of the plasma-facing materials. Further to this, the absence of disruptions relaxes the requirements to the plasma-facing materials in terms of thermal shock stability, although ELM activity occurs also in stellarators and leads to fast transient surface loads on the ms-time scale. Another aspect regarding the plasma-material interactions in stellarators is the sensitivity to impurity accumulation in the core plasma. Thus, it is preferred to apply low-Z materials until operation scenarios are established which do not lead to this accumulation process. In the case of high-Z materials impurity accumulation will lead to a radiative plasma collapse. For the stellarator W7-X low-Z plasma-facing materials have been selected to protect the divertor and the wall surfaces. Due to the stationary operation, the plasma-facing materials have to be bonded or clamped to actively water-cooled substrates to remove the incident heat fluxes. The following materials have been selected to fulfil the operational requirements: 1. A three directionally carbon fibre reinforced carbon composite (CFC) with very high thermal conductivity bonded to a water cooled CuCrZr heat sink for the divertor which will be exposed to heat fluxes up to 10MW/m 2 . 2. Isotropic fine grain graphite tiles mechanically clamped to a CuCrZr heat sink which is brazed to a stainless steel cooling tube for the areas of moderate heat fluxes up to 0.5 MW/m 2 (baffles, inner wall). 3. Thick boron carbide coating on water cooled steel panels for the outer wall surfaces with low heat fluxes up to 0.2 MW/m 2 . This coating would be applied on most surfaces only after the initial operation. In the presentation the properties of these materials will be discussed with a view to the plasma-wall interaction in W7-X. In fusion reactors

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

Science.gov (United States)

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.

Advanced solutions for beryllium and tungsten plasma-facing components

International Nuclear Information System (INIS)

Ibbott, C.; Jakeman, R.; Ando, T.; Chiocchio, S.; Federici, G.; Heidl, H.; Tivey, R.; Falter, H.; Ciric, D.; Merola, M.; Vieider, G.; Ploechl, L.; Roedig, M.

1998-01-01

Beryllium and tungsten are candidate plasma-facing armour materials for the International Thermonuclear Experimental Reactor (ITER). These armours are proposed for areas with low heat flux (≤5 MW m -2 ); however, in the divertor, surface melting during abnormal events may occur. This paper reports the progress made in developing novel approaches to solving the difficulties posed in designing with these armours. A Be monoblock brazed to an OFHC 10 mm ID Cu tube using InCuSil 'ABA' braze alloy has survived 130 cycles of 10-11 MW m -2 for 6 s, with surface temperatures of 1250 C. No visible surface cracking occurred. The same monoblock was then exposed to several cycles of 20-22 MW m -2 for 8 s, creating a 2 mm deep molten layer. High cycle fatigue was then performed. The test results are detailed in this paper. Comparison between experimental and theoretical results are made. W and Cu have a large mismatch in their thermal expansion coefficients and two designs are proposed that minimise the interface stresses. These are: a 'brush'-like structure with rectangular fibres set in a Cu substrate using the 'active metal casting' (AMC) technique; and thin monoblocks (or lamellae) brazed or active metal cast onto a Cu tube. Analyses of the lamellae concept for steady-state heat loads of 5 MW m -2 are presented. Fatigue analyses show that both solutions are theoretically viable (∝10 4 cycles). A 'brush' mock-up has been manufactured and progress on its testing is reported. Results of all tests and their relevance to the ITER design are discussed. (orig.)

Interaction of plasma-facing materials with air and steam

International Nuclear Information System (INIS)

Druyts, F.; Fays, J.; Wu, C.H.

2002-01-01

In the design of ITER-FEAT, several candidate materials are foreseen for plasma-facing components of the divertor (tungsten, carbon fibre-reinforced composites (CFC), molybdenum) and the first wall (beryllium). In the view of accidental scenarios such as a loss of coolant accident or a loss of vacuum accident the reaction between these materials and steam or air remains a safety concern. To provide kinetic data, describing the chemical reactivity of plasma-facing materials in air and steam, we used coupled thermogravimetry/quadrupole mass spectrometry. In this paper we present the results of a screening investigation that compares the oxidation rates of tungsten, molybdenum, CFC and beryllium in the temperature range 300-700 deg. C. From the thermogravimetry and mass spectrometry results we obtained the reaction rates as a function of temperature. For the metals tungsten, molybdenum and beryllium, a transition is observed between protective oxidation at lower temperatures and non-protective oxidation at higher temperatures. This transition temperature lies in the range 500-550 deg. C for tungsten and molybdenum, which is lower than for beryllium. At above temperatures 550 deg. C, the oxides formed on molybdenum and tungsten volatilise. This increases the oxidation rate dramatically and can lead to mobilisation of activation products in a fusion reactor. We also performed experiments on both undoped CFC and CFC doped with 8-10% silicon. The influence of silicon doping on the chemical reactivity of CFC's in air is discussed

Hydrogen gas driven permeation through tungsten deposition layer formed by hydrogen plasma sputtering

International Nuclear Information System (INIS)

Uehara, Keiichiro; Katayama, Kazunari; Date, Hiroyuki; Fukada, Satoshi

2015-01-01

Highlights: • H permeation tests for W layer formed by H plasma sputtering are performed. • H permeation flux through W layer is larger than that through W bulk. • H diffusivity in W layer is smaller than that in W bulk. • The equilibrium H concentration in W layer is larger than that in W bulk. - Abstract: It is important to evaluate the influence of deposition layers formed on plasma facing wall on tritium permeation and tritium retention in the vessel of a fusion reactor from a viewpoint of safety. In this work, tungsten deposition layers having different thickness and porosity were formed on circular nickel plates by hydrogen RF plasma sputtering. Hydrogen permeation experiment was carried out at the temperature range from 250 °C to 500 °C and at hydrogen pressure range from 1013 Pa to 101,300 Pa. The hydrogen permeation flux through the nickel plate with tungsten deposition layer was significantly smaller than that through a bare nickel plate. This indicates that a rate-controlling step in hydrogen permeation was not permeation through the nickel plate but permeation though the deposition layer. The pressure dependence on the permeation flux differed by temperature. Hydrogen permeation flux through tungsten deposition layer is larger than that through tungsten bulk. From analysis of the permeation curves, it was indicated that hydrogen diffusivity in tungsten deposition layer is smaller than that in tungsten bulk and the equilibrium hydrogen concentration in tungsten deposition layer is enormously larger than that in tungsten bulk at same hydrogen pressure.

ITER tungsten divertor design development and qualification program

Energy Technology Data Exchange (ETDEWEB)

Hirai, T., E-mail: takeshi.hirai@iter.org [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint Paul lez Durance (France); Escourbiac, F.; Carpentier-Chouchana, S.; Fedosov, A.; Ferrand, L.; Jokinen, T.; Komarov, V.; Kukushkin, A.; Merola, M.; Mitteau, R.; Pitts, R.A.; Shu, W.; Sugihara, M. [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint Paul lez Durance (France); Riccardi, B. [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Suzuki, S. [JAEA, Fusion Research and Development Directorate JAEA, 801-1 Mukouyama, Naka, Ibaragi 311-0193 (Japan); Villari, R. [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, I-00044 Frascati, Rome (Italy)

2013-10-15

Highlights: • Detailed design development plan for the ITER tungsten divertor. • Latest status of the ITER tungsten divertor design. • Brief overview of qualification program for the ITER tungsten divertor and status of R and D activity. -- Abstract: In November 2011, the ITER Council has endorsed the recommendation that a period of up to 2 years be set to develop a full-tungsten divertor design and accelerate technology qualification in view of a possible decision to start operation with a divertor having a full-tungsten plasma-facing surface. To ensure a solid foundation for such a decision, a full tungsten divertor design, together with a demonstration of the necessary high performance tungsten monoblock technology should be completed within the required timescale. The status of both the design and technology R and D activity is summarized in this paper.

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

Science.gov (United States)

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.

A fracture mechanics study of tungsten failure under high heat flux loads

International Nuclear Information System (INIS)

Li, Muyuan

2015-01-01

The performance of fusion devices is highly dependent on plasma-facing components. Tungsten is the most promising candidate material for armors in plasma-facing components in ITER and DEMO. However, the brittleness of tungsten below the ductile-to-brittle transition temperature is very critical to the reliability of plasma-facing components. In this work, thermo-mechanical and fracture behaviors of tungsten are predicted numerically under fusion relevant thermal loadings.

Development of quantitative atomic modeling for tungsten transport study using LHD plasma with tungsten pellet injection

Science.gov (United States)

Murakami, I.; Sakaue, H. A.; Suzuki, C.; Kato, D.; Goto, M.; Tamura, N.; Sudo, S.; Morita, S.

2015-09-01

Quantitative tungsten study with reliable atomic modeling is important for successful achievement of ITER and fusion reactors. We have developed tungsten atomic modeling for understanding the tungsten behavior in fusion plasmas. The modeling is applied to the analysis of tungsten spectra observed from plasmas of the large helical device (LHD) with tungsten pellet injection. We found that extreme ultraviolet (EUV) emission of W24+ to W33+ ions at 1.5-3.5 nm are sensitive to electron temperature and useful to examine the tungsten behavior in edge plasmas. We can reproduce measured EUV spectra at 1.5-3.5 nm by calculated spectra with the tungsten atomic model and obtain charge state distributions of tungsten ions in LHD plasmas at different temperatures around 1 keV. Our model is applied to calculate the unresolved transition array (UTA) seen at 4.5-7 nm tungsten spectra. We analyze the effect of configuration interaction on population kinetics related to the UTA structure in detail and find the importance of two-electron-one-photon transitions between 4p54dn+1- 4p64dn-14f. Radiation power rate of tungsten due to line emissions is also estimated with the model and is consistent with other models within factor 2.

Development of bonding techniques between tungsten and copper alloy for plasma facing components by HIP method (2). Bonding between tungsten and DS-copper

International Nuclear Information System (INIS)

Saito, Shigeru; Fukaya, Kiyoshi; Eto, Motokuni; Ishiyama, Shintaro; Akiba, Masato

2000-02-01

Recently, W (tungsten)-alloys are considered as plasma facing material (PFM) for ITER because of these many favorable properties such as high melting point (3655 K), relatively high thermal conductivity and higher resistivity for plasma sputtering. On the other hand, Cu-alloys, especially DS (dispersion strengthened)-Cu, are proposed as heat sink materials because of its high thermal conductivity and good mechanical properties at high temperature. Plasma facing components (PFC) are designed as the duplex structure where W armor tiles are bonded with Cu-alloy heat sink. Then, we started the bonding technology development by hot isostatic press (HIP) method to bond W with Cu-alloys because of its many advantages. Until now, it was reported that we could get the best HIP bonding conditions for W and OFHC-Cu and the tensile strength was similar with HIP treated OFHC-Cu. In this experiments, bonding tests of W and DS-Cu with insert material were performed. As insert material, OFHC-Cu was used with different thickness. Bonding conditions were selected as 1273 K x 2 hours x 147 MPa. Bonding tests with 0.3 to 1.8 mm thickness OFHC-Cu were successfully bonded but with 0.1 mm thickness was not bonded. From the results of tensile tests, the tensile strength of the specimens with 0.3 and 0.5 mm thickness were decreased at elevated temperature. It was shown that over 1.0 mm thickness OFHC-Cu insert may be needed and the tensile strength were a little higher than that of HIP treated OFHC-Cu. (author)

A study of plasma facing tungsten components with electrical discharge machined surface exposed to cyclic thermal loads

Energy Technology Data Exchange (ETDEWEB)

Seki, Yohji, E-mail: seki.yohji@jaea.go.jp; Ezato, Koichiro; Suzuki, Satoshi; Yokoyama, Kenji; Yamada, Hirokazu; Hirayama, Tomoyuki

2016-11-01

Through R&D for a plasma facing units (PFUs) in an outer vertical target of an ITER full-tungsten (W) divertor, Japan Atomic Energy Agency succeeded in demonstrating the durability of the W divertor shaped by an electrical discharge machining (EDM). To prevent melting of W armors in the PFUs, an adequate technology to meet requirements of a geometrical shape and a tolerance is one of the most important key issues in a manufacturing process. From the necessity, the EDM has been evaluated to control the final shape of the W armor. Though the EDM was known to be advantages such as an easy workability, a potential disadvantage of presence of micro-cracks on the W surface appeared. In order to examine a potential effect of the micro-crack on a heat removal durability, a high heat flux testing was carried out for the W divertor mock-up with the polish and the EDM. As the result, all of the W armors endured the repetitive heat load of 1000 cycles at an absorbed heat flux of more than 20 MW/m{sup 2}, which strongly encourages the realization of the PFUs of the ITER full-W divertor with the various geometrical shape and the high accuracy tolerance.

A study of plasma facing tungsten components with electrical discharge machined surface exposed to cyclic thermal loads

International Nuclear Information System (INIS)

Seki, Yohji; Ezato, Koichiro; Suzuki, Satoshi; Yokoyama, Kenji; Yamada, Hirokazu; Hirayama, Tomoyuki

2016-01-01

Through R&D for a plasma facing units (PFUs) in an outer vertical target of an ITER full-tungsten (W) divertor, Japan Atomic Energy Agency succeeded in demonstrating the durability of the W divertor shaped by an electrical discharge machining (EDM). To prevent melting of W armors in the PFUs, an adequate technology to meet requirements of a geometrical shape and a tolerance is one of the most important key issues in a manufacturing process. From the necessity, the EDM has been evaluated to control the final shape of the W armor. Though the EDM was known to be advantages such as an easy workability, a potential disadvantage of presence of micro-cracks on the W surface appeared. In order to examine a potential effect of the micro-crack on a heat removal durability, a high heat flux testing was carried out for the W divertor mock-up with the polish and the EDM. As the result, all of the W armors endured the repetitive heat load of 1000 cycles at an absorbed heat flux of more than 20 MW/m"2, which strongly encourages the realization of the PFUs of the ITER full-W divertor with the various geometrical shape and the high accuracy tolerance.

Development of quantitative atomic modeling for tungsten transport study Using LHD plasma with tungsten pellet injection

International Nuclear Information System (INIS)

Murakami, I.; Sakaue, H.A.; Suzuki, C.; Kato, D.; Goto, M.; Tamura, N.; Sudo, S.; Morita, S.

2014-10-01

Quantitative tungsten study with reliable atomic modeling is important for successful achievement of ITER and fusion reactors. We have developed tungsten atomic modeling for understanding the tungsten behavior in fusion plasmas. The modeling is applied to the analysis of tungsten spectra observed from currentless plasmas of the Large Helical Device (LHD) with tungsten pellet injection. We found that extreme ultraviolet (EUV) lines of W 24+ to W 33+ ions are very sensitive to electron temperature (Te) and useful to examine the tungsten behavior in edge plasmas. Based on the first quantitative analysis of measured spatial profile of W 44+ ion, the tungsten concentration is determined to be n(W 44+ )/n e = 1.4x10 -4 and the total radiation loss is estimated as ∼4 MW, of which the value is roughly half the total NBI power. (author)

Progress of research on plasma facing materials in University of Science and Technology Beijing

International Nuclear Information System (INIS)

Ge, Chang-Chun; Zhou, Zhang-Jian; Song, Shu-Xiang; Du, Juan; Zhong, Zhi-Hong

2007-01-01

In this paper, we report some new progress on plasma facing materials in University of Science and Technology Beijing (USTB), China. They include fabrication of tungsten coating with ultra-fine grain size by atmosphere plasma spraying; fabrication of tungsten with ultra-fine grain size by a newly developed method named as resistance sintering under ultra-high pressure; using the concept of functionally graded materials to join tungsten to copper based heat sink; joining silicon doped carbon to copper by brazing using a Ti based amorphous filler and direct casting

Heat load and deuterium plasma effects on SPS and WSP tungsten

Directory of Open Access Journals (Sweden)

Vilémová Monika

2015-06-01

Full Text Available Tungsten is a prime choice for armor material in future nuclear fusion devices. For the realization of fusion, it is necessary to address issues related to the plasma–armor interactions. In this work, several types of tungsten material were studied, i.e. tungsten prepared by spark plasma sintering (SPS and by water stabilized plasma spraying (WSP technique. An intended surface porosity was created in the samples to model hydrogen/helium bubbles. The samples were subjected to a laser heat loading and a radiation loading of deuterium plasma to simulate edge plasma conditions of a nuclear fusion device (power density of 108 W/cm2 and 107 W/cm2, respectively, in the pulse intervals up to 200 ns. Thermally induced changes in the morphology and the damage to the studied surfaces are described. Possible consequences for the fusion device operation are pointed out.

Toward Tungsten Plasma-Facing Components in KSTAR: Research on Plasma-Metal Wall Interaction

NARCIS (Netherlands)

Hong, S. H.; Kim, K. M.; Song, J. H.; Bang, E. N.; Kim, H. T.; Lee, K. S.; Litnovsky, A.; Hellwig, M.; Seo, D. C.; van den Berg, M. A.; Lee, H. H.; Kang, C. S.; Lee, H. Y.; Hong, J. H.; Bak, J. G.; Kim, H. S.; Juhn, J. W.; Son, S. H.; Kim, H. K.; Douai, D.; Grisolia, C.; Wu, J.; Luo, G. N.; Choe, W. H.; Komm, M.; De Temmerman, G.; Pitts, R.

2015-01-01

One of the main missions of KSTAR is to develop long-pulse operation capability relevant to the production of fusion energy. After a full metal wall configuration was decided for ITER, a major upgrade for KSTAR was planned, to a tungsten first wall similar to the JET ITER-like wall (coatings and

Advanced tungsten materials for plasma-facing components of DEMO and fusion power plants

International Nuclear Information System (INIS)

Neu, R.; Riesch, J.; Coenen, J.W.; Brinkmann, J.; Calvo, A.; Elgeti, S.; García-Rosales, C.; Greuner, H.; Hoeschen, T.; Holzner, G.; Klein, F.; Koch, F.

2016-01-01

Highlights: • Development of W-fibre enhanced W-composites incorporating extrinsic toughening mechanisms. • Production of a large sample (more than 2000 long fibres) for mechanical and thermal testing. • Even in a fully embrittled state, toughening mechanisms are still effective. • Emissions of volatile W-oxides can be suppressed by alloying W with elements forming stable oxides. • WCr10Ti2 has been successfully tested under accidental conditions and high heat fluxes. - Abstract: Tungsten is the major candidate material for the armour of plasma facing components in future fusion devices. To overcome the intrinsic brittleness of tungsten, which strongly limits its operational window, a W-fibre enhanced W-composite material (W_f/W) has been developed incorporating extrinsic toughening mechanisms. Small W_f/W samples show a large increase in toughness. Recently, a large sample (50 mm × 50 mm × 3 mm) with more than 2000 long fibres has been successfully produced allowing further mechanical and thermal testing. It could be shown that even in a fully embrittled state, toughening mechanisms as crack bridging by intact fibres, as well as the energy dissipation by fibre-matrix interface debonding and crack deflection are still effective. A potential problem with the use of pure W in a fusion reactor is the formation of radioactive and highly volatile WO_3 compounds and their potential release under accidental conditions. It has been shown that the oxidation of W can be strongly suppressed by alloying with elements forming stable oxides. WCr10Ti2 alloy has been produced on a technical scale and has been successfully tested in the high heat flux test facility GLADIS. Recently, W-Cr-Y alloys have been produced on a lab-scale. They seem to have even improved properties compared to the previously investigated W alloys.

Advanced tungsten materials for plasma-facing components of DEMO and fusion power plants

Energy Technology Data Exchange (ETDEWEB)

Neu, R., E-mail: Rudolf.Neu@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Fakultät für Maschinenbau, Technische Universität München, D-85748 Garching (Germany); Riesch, J. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Coenen, J.W. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Brinkmann, J. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Calvo, A. [CEIT and Tecnun (University of Navarra), E-20018 San Sebastian (Spain); Elgeti, S. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); García-Rosales, C. [CEIT and Tecnun (University of Navarra), E-20018 San Sebastian (Spain); Greuner, H.; Hoeschen, T.; Holzner, G. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Klein, F. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Koch, F. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); and others

2016-11-01

Highlights: • Development of W-fibre enhanced W-composites incorporating extrinsic toughening mechanisms. • Production of a large sample (more than 2000 long fibres) for mechanical and thermal testing. • Even in a fully embrittled state, toughening mechanisms are still effective. • Emissions of volatile W-oxides can be suppressed by alloying W with elements forming stable oxides. • WCr10Ti2 has been successfully tested under accidental conditions and high heat fluxes. - Abstract: Tungsten is the major candidate material for the armour of plasma facing components in future fusion devices. To overcome the intrinsic brittleness of tungsten, which strongly limits its operational window, a W-fibre enhanced W-composite material (W{sub f}/W) has been developed incorporating extrinsic toughening mechanisms. Small W{sub f}/W samples show a large increase in toughness. Recently, a large sample (50 mm × 50 mm × 3 mm) with more than 2000 long fibres has been successfully produced allowing further mechanical and thermal testing. It could be shown that even in a fully embrittled state, toughening mechanisms as crack bridging by intact fibres, as well as the energy dissipation by fibre-matrix interface debonding and crack deflection are still effective. A potential problem with the use of pure W in a fusion reactor is the formation of radioactive and highly volatile WO{sub 3} compounds and their potential release under accidental conditions. It has been shown that the oxidation of W can be strongly suppressed by alloying with elements forming stable oxides. WCr10Ti2 alloy has been produced on a technical scale and has been successfully tested in the high heat flux test facility GLADIS. Recently, W-Cr-Y alloys have been produced on a lab-scale. They seem to have even improved properties compared to the previously investigated W alloys.

Effects of fusion relevant transient energetic radiation, plasma and thermal load on PLANSEE double forged tungsten samples in a low-energy plasma focus device

Science.gov (United States)

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

Thermal conductivity reduction of tungsten plasma facing material due to helium plasma irradiation in PISCES using the improved 3-omega method

Energy Technology Data Exchange (ETDEWEB)

Cui, Shuang [Department of Mechanical and Aerospace Engineering, University of California, San Diego, La Jolla, CA 92093 (United States); Simmonds, Michael [Department of Physics, University of California, San Diego, La Jolla, CA 92093 (United States); Center for Energy Research, University of California, San Diego, La Jolla, CA 92093 (United States); Qin, Wenjing; Ren, Feng [School of Physics and Technology, Wuhan University, Wuhan, Hubei 430072 (China); Tynan, George R. [Department of Mechanical and Aerospace Engineering, University of California, San Diego, La Jolla, CA 92093 (United States); Center for Energy Research, University of California, San Diego, La Jolla, CA 92093 (United States); Doerner, Russell P. [Center for Energy Research, University of California, San Diego, La Jolla, CA 92093 (United States); Chen, Renkun, E-mail: rkchen@ucsd.edu [Department of Mechanical and Aerospace Engineering, University of California, San Diego, La Jolla, CA 92093 (United States); Center for Energy Research, University of California, San Diego, La Jolla, CA 92093 (United States)

2017-04-01

The near-surface region of plasma facing material (PFM) plays an important role in thermal management of fusion reactors. In this work, we measured thermal conductivity of tungsten (W) surface layers damaged by He plasma in PISCES at UCSD. We studied the damage effect on both bulk, and thin film, W. We observed that the surface morphology of both bulk and thin film was altered after exposure to He plasma with the fluence of 1 × 10{sup 26} m{sup −2} (bulk) and 2 × 10{sup 24} m{sup −2} (thin film). Transmission electron microscopy (TEM) analysis reveals that the depth of the irradiation damaged layer was approximately 20 nm on the bulk W exposed to He plasma at 773 K for 2000 s. In order to measure the thermal conductivity of this exceedingly thin damaged layer in the bulk W, we adopted the well-established ‘3-omega’ method and employed novel nanofabrication techniques to improve the measurement sensitivity. For the damaged W thin film sample, we measured the reduction in electrical conductivity and used the Wiedemann-Franz (W-F) law to extract the thermal conductivity. Results from both measurements show that thermal conductivity in the damaged layers was reduced by at least ∼80% compared to that of undamaged W. This large reduction in thermal conductivity can be attributed to the scattering of electrons, the dominant heat carriers in W, caused by defects introduced by He plasma irradiation.

Behavior of tungsten carbide in water stabilized plasma

Czech Academy of Sciences Publication Activity Database

Brožek, Vlastimil; Matějíček, Jiří; Neufuss, Karel

2007-01-01

Roč. 7, č. 4 (2007), s. 213-220 ISSN 1335-8987 R&D Projects: GA ČR(CZ) GA104/05/0540 Institutional research plan: CEZ:AV0Z20430508 Keywords : water stabilized plasma * tungsten carbide * tungsten hemicarbide * decarburization Subject RIV: BL - Plasma and Gas Discharge Physics

Overview of processing technologies for tungsten-steel composites and FGMs for fusion applications

Directory of Open Access Journals (Sweden)

Matějíček Jiří

2015-06-01

Full Text Available Tungsten is a prime candidate material for the plasma-facing components in future fusion devices, e.g. ITER and DEMO. Because of the harsh and complex loading conditions and the differences in material properties, joining of the tungsten armor to the underlying construction and/or cooling parts is a complicated issue. To alleviate the thermal stresses at the joint, a sharp interface may be replaced by a gradual one with a smoothly varying composition. In this paper, several techniques for the formation of tungsten-steel composites and graded layers are reviewed. These include plasma spraying, laser cladding, hot pressing and spark plasma sintering. Structure, composition and selected thermal and mechanical properties of representative layers produced by each of these techniques are presented. A summary of advantages and disadvantages of the techniques and an assessment of their suitability for the production of plasma-facing components is provided.

Developments toward the use of tungsten as armour material in plasma facing components promoted by Euratom-CEA Association

International Nuclear Information System (INIS)

Mitteau, R.; Missiaen, J.M.; Brustolin, P.

2006-01-01

Tungsten is increasingly considered as a prime candidate armour material facing the plasma in fusion experiments (ASDEX, JET, ITER). This material is, however, a challenge for the engineers due to its brittleness at room temperature. Its bonding to structural or cooled substrates is a critical issue. The Euratom-CEA Association promotes the development of evolutionary techniques aiming to produce high performance assemblies between tungsten and various substrates. These are 1) functionally graded tungsten to copper, 2) direct electron beam welding of tungsten to Mo-alloy TZM and 3) the characterisation of tungsten coatings deposited on carbon fibre composite by high energy deposition processes. 1) A functionally graded material eliminates the singular point which weakens the heterogeneous assembly, reducing the stresses and allowing a better behaviour. The sintering of submicronic W-Cu powders is investigated. The green shape is processed from W-CuO powder, which is reduced by a hydrogen flow. The compaction and sintering of layers of various compositions (10 to 30 % Cu) produces an assembly (density of ∼ 94%) with a good cohesion. However, the gradient is not effectively controlled, because of the migration of melt copper during the sintering. Future work aims to improve the process by using spark or microwave assisted sintering. 2) Electron beam welding of Mo-alloy TZM is investigated, to produce high temperature components required by radiation cooled PFCs. They require only mechanical properties and no vacuum sealing. The driving line is to use simple tungsten shapes to reduce the milling cost. In spite of low weldable properties of the refractory alloys, a good bonding up to a depth of 5 mm is obtained. Hardness measurements show that the melt area and the heat affected zone are harder than TZM, the weakest materials at 230 Hv. Quench tests in water from up to 2000 o C are done without apparent crack formation. 3) Finally, characterisation techniques are

Selection of plasma facing materials for ITER

International Nuclear Information System (INIS)

Ulrickson, M.; Barabash, V.; Chiocchio, S.

1996-01-01

ITER will be the first tokamak having long pulse operation using deuterium-tritium fuel. The problem of designing heat removal structures for steady state in a neutron environment is a major technical goal for the ITER Engineering Design Activity (EDA). The steady state heat flux specified for divertor components is 5 MW/m 2 for normal operation with transients to 15 MW/m 2 for up to 10 s. The selection of materials for plasma facing components is one of the major research activities. Three materials are being considered for the divertor; carbon fiber composites, beryllium, and tungsten. This paper discusses the relative advantages and disadvantages of these materials. The final section of plasma facing materials for the ITER divertor will not be made until the end of the EDA

Plasma exposure of different tungsten grades with plasma accelerators under ITER-relevant conditions

International Nuclear Information System (INIS)

Makhlaj, Vadym A; Garkusha, Igor E; Aksenov, Nikolay N; Byrka, Oleg V; Bazylev, Boris; Landman, Igor; Linke, Jochen; Wirtz, Marius; Malykhin, Sergey V; Pugachov, Anatoliy T; Sadowski, Marek J; Skladnik-Sadowska, Elzbieta

2014-01-01

This paper presents the results of tungsten irradiation experiments performed with three plasma facilities: the QSPA Kh-50 quasi-steady-state plasma accelerator, the PPA pulsed plasma gun and the magneto-plasma compressor. Targets made of different kinds of tungsten (sintered, rolled and deformed) were irradiated with powerful plasma streams at heat fluxes relevant to edge-localized modes in ITER. The irradiated targets were analyzed and two different meshes of cracks were identified. It has been shown that the major cracks do not depend on the tungsten grade. This has been attributed to ductile-to-brittle transition effects. Meshes of inter-granular micro-cracks were detected for energy loads above the melting threshold and these were probably caused by the re-solidification process. The blister-like and cellular-like structures were observed on sample surfaces exposed to helium and hydrogen plasmas. (paper)

PFMC14. 14th international conference on plasma-facing materials and components for fusion applications. Book of abstracts

International Nuclear Information System (INIS)

2013-01-01

The performance of fusion devices and of a future fusion power plant critically depends on the plasma facing materials and components. Resistance to local heat and particle loads, thermo-mechanical properties, as well as the response to neutron damage of the selected materials are critical parameters which need to be understood and tailored from atomistic to component levels. The 14th International Conference on Plasma-Facing Materials and Components for Fusion Applications addresses these issues. Among the topics of the joint conference recent developments and research results in the following fields are addressed: - Tungsten and tungsten alloys - Low-Z materials - Mixed materials - Erosion, redeposition and fuel retention - Materials under extreme thermal loads - Technology and testing of plasma-facing components - Neutron effects in plasma-facing materials - Advanced characterization of materials and components. Selected international speakers present overview lectures and treat detailed aspects of the given topics. Contributed papers to the subjects of the meeting are solicited for oral and poster presentations.

Irradiation effects of hydrogen and helium plasma on different grade tungsten materials

Directory of Open Access Journals (Sweden)

X. Liu

2017-08-01

Full Text Available Fine-grain tungsten alloys could be one of the solutions for the plasma facing materials of future DEMO reactors. In order to evaluate the service performances of the newly developed W alloys under edge plasma irradiation and the synergetic effect of fusion plasma together with high heat flux, both low energy He ions and high energy H, H/He mixed neutral beam irradiation on W-ZrC, W-K, W-Y2O3, W-La2O3 and CVD-W coating were performed respectively at a liner plasma facility (Dalian Nationality University, China and the neutral beam facility GLADIS (IPP, Germany. Surface damages were characterized, and the crack formation and extension behaviors under ELM-like transient loading after H and H/He mixed beam irradiation were also investigated in the 60kW EMS-60 facility (Electron beam Materials testing Scenario at SWIP (Southwestern Institute of Physics, China. The experimental results indicated that surface damages induced by low or high energy H/He ion/neutral beam didn't closely correlate with the type of tungsten materials. However, H/He (6at% He concentration neutral beam induced more significant surface damages of the tested W materials than only H neutral beam irradiation under the similar irradiation conditions. Similarly, the mixed H/He pre-exposure remarkably reduced the critical power of crack initiation compared with the un-irradiated samples under 100 repetitive loads of 1ms pulse, while no significant degeneration for the case of only H beam irradiation was observed.

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

Energy Technology Data Exchange (ETDEWEB)

Grigorev, Petr, E-mail: grigorievpit@gmail.com [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, Mol, 2400 (Belgium); Ghent University, Applied Physics EA17 FUSION-DC, St. Pietersnieuwstraat, 41 B4, B-9000, Gent (Belgium); Department of Experimental Nuclear Physics K-89, Institute of Physics, Nanotechnologies, and Telecommunications, Peter the Great St. Petersburg Polytechnic University, St. Petersburg (Russian Federation); Matveev, Dmitry [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425, Jülich (Germany); Bakaeva, Anastasiia [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, Mol, 2400 (Belgium); Department of Applied Physics, Ghent University (Belgium); Terentyev, Dmitry [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, Mol, 2400 (Belgium); Zhurkin, Evgeny E. [Department of Experimental Nuclear Physics K-89, Institute of Physics, Nanotechnologies, and Telecommunications, Peter the Great St. Petersburg Polytechnic University, St. Petersburg (Russian Federation); Van Oost, Guido [Ghent University, Applied Physics EA17 FUSION-DC, St. Pietersnieuwstraat, 41 B4, B-9000, Gent (Belgium); Noterdaeme, Jean-Marie [Ghent University, Applied Physics EA17 FUSION-DC, St. Pietersnieuwstraat, 41 B4, B-9000, Gent (Belgium); Max-Planck-Institut für Plasmaphysik, Garching (Germany)

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 = 10{sup 24} D/m{sup 2}/s and fluence of 10{sup 26} D/m{sup 2}) 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{sup −5} at{sup −1}, while the bulk retention is about 4 × 10{sup −7} at{sup −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.

Neutron irradiation effects on plasma facing materials

Science.gov (United States)

Barabash, V.; Federici, G.; Rödig, M.; Snead, L. L.; Wu, C. H.

2000-12-01

This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed.

Neutron irradiation effects on plasma facing materials

International Nuclear Information System (INIS)

Barabash, V.; Federici, G.; Roedig, M.; Snead, L.L.; Wu, C.H.

2000-01-01

This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed

Operation of ASDEX Upgrade with tungsten coated walls

International Nuclear Information System (INIS)

Rohde, V.

2002-01-01

An alternative for low-Z materials in the main chamber of a future fusion device are high-Z materials, but the maximal tolerable concentration in the plasma core is restricted. A step by step approach to employ tungsten at the central column of ASDEX Upgrade was started in 1999. Meanwhile almost the whole central column is covered with tiles, which were coated by PVD with tungsten. Up to now 9000 s of plasma discharge covering all relevant scenarios were performed. Routine operation of ASDEX Upgrade was not affected by the tungsten. Typical concentrations below 10 -5 were found. The tungsten concentration is mostly connected to the transport into the core plasma, not to the tungsten erosion. It can be demonstrated, that additional central heating can eliminate the tungsten accumulation. These experiments demonstrate the compatibility of fusion plasmas with W plasma facing components under reactor relevant conditions. The erosion pattern found by post mortem analysis indicates that the main effect is ion sputtering. The main erosion of tungsten seems to occur during plasma ramp-up and ramp-down. (author)

Development of advanced high heat flux and plasma-facing materials

Science.gov (United States)

Linsmeier, Ch.; Rieth, M.; Aktaa, J.; Chikada, T.; Hoffmann, A.; Hoffmann, J.; Houben, A.; Kurishita, H.; Jin, X.; Li, M.; Litnovsky, A.; Matsuo, S.; von Müller, A.; Nikolic, V.; Palacios, T.; Pippan, R.; Qu, D.; Reiser, J.; Riesch, J.; Shikama, T.; Stieglitz, R.; Weber, T.; Wurster, S.; You, J.-H.; Zhou, Z.

2017-09-01

Plasma-facing materials and components in a fusion reactor are the interface between the plasma and the material part. The operational conditions in this environment are probably the most challenging parameters for any material: high power loads and large particle and neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak or stellarator reactor, given the proven geometries and technological solutions, requires an improvement of the thermo-mechanical capabilities of currently available materials. In its first part this article describes the requirements and needs for new, advanced materials for the plasma-facing components. Starting points are capabilities and limitations of tungsten-based alloys and structurally stabilized materials. Furthermore, material requirements from the fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, defining directions for the material development. Finally, safety requirements for a fusion reactor with its specific accident scenarios and their potential environmental impact lead to the definition of inherently passive materials, avoiding release of radioactive material through intrinsic material properties. The second part of this article demonstrates current material development lines answering the fusion-specific requirements for high heat flux materials. New composite materials, in particular fiber-reinforced and laminated structures, as well as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical operation space towards regions of extreme steady-state and transient loads. Self-passivating tungsten alloys, demonstrating favorable tungsten-like plasma-wall interaction behavior under normal operation conditions, are an intrinsic solution to otherwise catastrophic consequences of loss-of-coolant and air ingress events in a fusion reactor. Permeation barrier layers avoid the escape of tritium into structural and cooling

Tungsten and carbon surface change under high dose plasma exposure

International Nuclear Information System (INIS)

Martynenko, Y.V.; Khripunov, B.I.; Petrov, V.B.

2009-01-01

Study of surface composition dynamics has been made on the LENTA linear plasma simulator. Experiments have been made on tungsten and carbon materials subjected to steady-state plasma exposure. The achieved ion doses on the surface were 10 21 ion cm -2 . WL 10 tungsten containing 1% of La2O3 oxide and titanium-doped graphite RG-T were studied. The following experimental conditions were varied in these experiments: energy of ions, surface temperature, working gas. Irradiations of tungsten WL 10 were executed in deuterium plasma at low ion energies (about 20 eV) and at 200 eV for temperatures below 340 K. Graphite RG-T was exposed at 1300 K. Elevated surface temperature (about 1050K) was also characteristic of experiments on tungsten sample under nitrogen plasma impact (simulated inter-ELMs condition). Surface microstructure modification has been observed and surface composition changes were found on the materials showing influence of high dose plasma irradiations on element redistribution in the near surface layers. (author)

Self Passivating W-based Alloys as Plasma Facing Material

International Nuclear Information System (INIS)

Koch, F.; Koeppl, S.; Bolt, H.

2007-01-01

Full text of publication follows: Tungsten (W) is presently the main candidate material for the plasma-facing protection of future fusion power reactors due to the low sputter erosion under bombardment by energetic D, T and He ions. Thus a W-based protection material may provide a wall erosion lifetime of the order of five years which is a pre-requisite for economic fusion reactor operation. A potential problem with the use of pure W in a fusion reactor is the formation of radioactive and highly volatile WO 3 compounds and their potential release under accidental conditions. A loss-of-coolant event in a He-cooled reactor would lead to a temperature rise to 1100 deg. C after approx. 10 to 30 days due to the nuclear decay heat of the in-vessel components. In such a situation additional accidental intense air ingress into the reactor vessel would lead to the formation of WO 3 and subsequent evaporation of radioactive (WO 3 ) x -clusters. The use of self passivating W alloys either as bulk material or as thick coating on the steel wall may be a passively safe alternative for the plasma-facing protection. The use of this material would eliminate the above mentioned concern related to pure W. To enable the formation of a protective film in oxidizing atmosphere which seals the tungsten surface from further oxidation, different elements have been investigated as corrosion protection additives. Therefore binary and ternary tungsten alloys were synthesised using magnetron sputtering. The oxidation behaviour of films deposited on inert substrates was measured with a thermo-balance set up under synthetic air at temperatures up to 1000 deg. C. Binary alloys of W-Si showed good self passivation properties by forming a SiO 2 film at the surface. The oxidation rate of a compound containing 11 wt.% Si was reduced by a factor of 10 2 compared to pure tungsten between 800 deg. C and 1000 deg. C. Using ternary alloys the oxidation behaviour could be further improved. A compound of W

Deuterium inventory in tungsten after plasma exposure. A microstructural survey

International Nuclear Information System (INIS)

Manhard, Armin

2012-09-01

Tungsten is a promising material for armouring the plasma-facing wall of future nuclear fusion experiments and power plants. It has a very high melting point, good thermal conductivity and is highly resistant against physical sputtering by energetic particles from the plasma. It also has a very low solubility for hydrogen isotopes. This is important both for safety and also for economic reasons, in particular with regard to the radioactive fusion fuel tritium. Due to this low solubility, the retention of hydrogen isotopes in tungsten materials after exposure to a plasma is dominated by the trapping of hydrogen isotopes at tungsten lattice defects. Therefore, a strong dependence of the hydrogen isotope retention on the microstructure of the tungsten is to be expected. This work describes a survey study of tungsten with different microstructures exposed to deuterium plasmas under a wide range of different plasma exposure conditions. The isotope deuterium was used because its natural abundance is much smaller than that of hydrogen (i.e., 1 H). This allows detecting even very small amounts retained in the tungsten practically without background signal. Furthermore, the use of deuterium allows utilising the nuclear reaction 2 D( 3 He,p) 4 He for depth-resolved quantification of the deuterium inventory up to depths of several microns. In order to standardise the specimens as far as possible, they were all cut from the same initial material from a single manufacturing batch. After a chemo-mechanical polishing procedure, which produces a well-defined surface, the specimens were annealed at either of four different temperatures in order to modify the grain structure and the dislocation density. These were then characterised by scanning electron microscopy and scanning transmission electron microscopy. The specimens were subsequently exposed in a fully characterised deuterium plasma source at different specimen temperatures, ion energies and deuterium fluences. In addition

Studies of plasma interactions with tungsten targets in PF-1000U facility

Directory of Open Access Journals (Sweden)

Ladygina Maryna S.

2016-06-01

Full Text Available This paper presents results of experimental studies of tungsten samples of 99.95% purity, which were irradiated by intense plasma-ion streams. The behaviour of tungsten, and particularly its structural change induced by high plasma loads, is of great importance for fusion technology. The reported measurements were performed within a modified PF-1000U plasma-focus facility operated at the IFPiLM in Warsaw, Poland. The working gas was pure deuterium. In order to determine the main plasma parameters and to study the behaviour of impurities at different instants of the plasma discharge, the optical emission spectroscopy was used. The dependence of plasma parameters on the initial charging voltage (16, 19 and 21 kV was studied. Detailed optical measurements were performed during interactions of a plasma stream with the tungsten samples placed at the z-axis of the facility, at a distance of 6 cm from the electrode outlets. The recorded spectra showed distinct WI and WII spectral lines. Investigation of a target surface morphology, after its irradiation by intense plasma streams, was performed by means of an optical microscope. The observations revealed that some amounts of the electrodes material (mainly copper were deposited upon the irradiated sample surface. In all the cases, melted zones were observed upon the irradiated target surface, and in experiments performed at the highest charging voltage there were formed some cracks.

Tungsten Ions in Plasmas: Statistical Theory of Radiative-Collisional Processes

Directory of Open Access Journals (Sweden)

Alexander V. Demura

2015-05-01

Full Text Available The statistical model for calculations of the collisional-radiative processes in plasmas with tungsten impurity was developed. The electron structure of tungsten multielectron ions is considered in terms of both the Thomas-Fermi model and the Brandt-Lundquist model of collective oscillations of atomic electron density. The excitation or ionization of atomic electrons by plasma electron impacts are represented as photo-processes under the action of flux of equivalent photons introduced by E. Fermi. The total electron impact single ionization cross-sections of ions Wk+ with respective rates have been calculated and compared with the available experimental and modeling data (e.g., CADW. Plasma radiative losses on tungsten impurity were also calculated in a wide range of electron temperatures 1 eV–20 keV. The numerical code TFATOM was developed for calculations of radiative-collisional processes involving tungsten ions. The needed computational resources for TFATOM code are orders of magnitudes less than for the other conventional numerical codes. The transition from corona to Boltzmann limit was investigated in detail. The results of statistical approach have been tested by comparison with the vast experimental and conventional code data for a set of ions Wk+. It is shown that the universal statistical model accuracy for the ionization cross-sections and radiation losses is within the data scattering of significantly more complex quantum numerical codes, using different approximations for the calculation of atomic structure and the electronic cross-sections.

Tungsten as First Wall Material in Fusion Devices

International Nuclear Information System (INIS)

Kaufmann, M.

2006-01-01

In the PLT tokamak with a tungsten limiter strong cooling of the central plasma was observed. Since then mostly graphite has been used as limiter or target plate material. Only a few tokamaks (limiter: FTU, TEXTOR; divertor: Alcator C-Mod, ASDEX Upgrade) gained experience with high-Z-materials. With the observed strong co- deposition of tritium together with carbon in JET and as a result of design studies of fusion reactors, it became clear that in the long run tungsten is the favourite for the first-wall material. Tungsten as a plasma facing material requires intensive research in all areas, i.e. in plasma physics, plasma wall-interaction and material development. Tungsten as an impurity in the confined plasma reveals considerable differences to carbon. Strong radiation at high temperatures, in connection with mostly a pronounced inward drift forms a particular challenge. Turbulent transport plays a beneficial role in this regard. The inward drift is an additional problem in the pedestal region of H-mode plasmas in ITER-like configurations. The erosion by low energy hydrogen atoms is in contrast to carbon small. However, erosion by fast particles from heating measures and impurity ions, accelerated in the sheath potential, play an important role in the case of tungsten. Radiation by carbon in the plasma boundary reduces the load to the target plates. Neon or Argon as substitutes will increase the erosion of tungsten. So far experiments have demonstrated that in most scenarios the tungsten content in the central plasma can be kept sufficiently small. The material development is directed to the specific needs of existing or future devices. In ASDEX Upgrade, which will soon be a divertor experiment with a complete tungsten first-wall, graphite tiles are coated with tungsten layers. In ITER, the solid tungsten armour of the target plates has to be castellated because of its difference in thermal expansion compared to the cooling structure. In a reactor the technical

Effect of high-flux H/He plasma exposure on tungsten damage due to transient heat loads

Czech Academy of Sciences Publication Activity Database

De Temmerman, G.; Morgan, T.W.; van Eden, G.G.; de Kruif, T.; Wirtz, M.; Matějíček, Jiří; Chráska, Tomáš; Pitts, R.A.; Wright, G.M.

2015-01-01

Roč. 463, August (2015), s. 198-201 ISSN 0022-3115. [PLASMA-SURFACE INTERACTIONS 21: International Conference on Plasma-Surface Interactions in Controlled Fusion Devices. Kanazawa, 26.05.2014-30.05.2014] R&D Projects: GA ČR(CZ) GA14-12837S Institutional support: RVO:61389021 Keywords : plasma-facing components * tungsten * hydrogen * helium * ELM Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 2.199, year: 2015 http://www.sciencedirect.com/science/article/pii/S0022311514006758#

Irradiation effects in tungsten-copper laminate composite

Energy Technology Data Exchange (ETDEWEB)

Garrison, L.M., E-mail: garrisonlm@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Katoh, Y. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Snead, L.L. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Byun, T.S. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Reiser, J.; Rieth, M. [Karlsruhe Institute of Technology, Karlsruhe (Germany)

2016-12-01

Tungsten-copper laminate composite has shown promise as a structural plasma-facing component as compared to tungsten rod or plate. The present study evaluated the tungsten-copper composite after irradiation in the High Flux Isotope Reactor (HFIR) at temperatures of 410–780 °C and fast neutron fluences of 0.02–9.0 × 10{sup 25} n/m{sup 2}, E > 0.1 MeV, 0.0039–1.76 displacements per atom (dpa) in tungsten. Tensile tests were performed on the composites, and the fracture surfaces were analyzed with scanning electron microscopy. Before irradiation, the tungsten layers had brittle cleavage failure, but the overall composite had 15.5% elongation at 22 °C. After only 0.0039 dpa this was reduced to 7.7% elongation, and no ductility was observed after 0.2 dpa at all irradiation temperatures when tensile tested at 22 °C. For elevated temperature tensile tests after irradiation, the composite only had ductile failure at temperatures where the tungsten was delaminating or ductile. - Highlights: • Fusion reactors need a tough, ductile tungsten plasma-facing material. • The unirradiated tungsten-copper laminate is more ductile than tungsten alone. • After neutron irradiation, the composite has significantly less ductility. • The tungsten behavior appears to dominate the overall composite behavior.

A supercritical carbon dioxide plasma process for preparing tungsten oxide nanowires

International Nuclear Information System (INIS)

Kawashima, Ayato; Nomura, Shinfuku; Toyota, Hiromichi; Takemori, Toshihiko; Mukasa, Shinobu; Maehara, Tsunehiro

2007-01-01

A supercritical carbon dioxide (CO 2 ) plasma process for fabricating one-dimensional tungsten oxide nanowires coated with amorphous carbon is presented. High-frequency plasma was generated in supercritical carbon dioxide at 20 MPa by using tungsten electrodes mounted in a supercritical cell, and subsequently an organic solvent was introduced with supercritical carbon dioxide into the plasma. Electron microscopy and Raman spectroscopy investigations of the deposited materials showed the production of tungsten oxide nanowires with or without an outer layer. The nanowires with an outer layer exhibited a coaxial structure with an outer concentric layer of amorphous carbon and an inner layer of tungsten oxide with a thickness and diameter of 20-30 and 10-20 nm, respectively

On thermionic emission from plasma-facing components in tokamak-relevant conditions.

Czech Academy of Sciences Publication Activity Database

Komm, Michael; Ratynskaia, S.; Tolias, P.; Cavalier, Jordan; Dejarnac, Renaud; Gunn, J. P.; Podolník, Aleš

2017-01-01

Roč. 59, č. 9 (2017), č. článku 094002. ISSN 0741-3335 R&D Projects: GA ČR(CZ) GA16-14228S; GA MŠk(CZ) 8D15001 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : thermionic * PIC * tungsten * tokamak * thermionic emission * plasma facing components * particle-in-cell Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.392, year: 2016 http://iopscience.iop.org/article/10.1088/1361-6587/aa78c4/pdf

Tungsten Deposition on Graphite using Plasma Enhanced Chemical Vapour Deposition

International Nuclear Information System (INIS)

Sharma, Uttam; Chauhan, Sachin S; Sharma, Jayshree; Sanyasi, A K; Ghosh, J; Choudhary, K K; Ghosh, S K

2016-01-01

The tokamak concept is the frontrunner for achieving controlled thermonuclear reaction on earth, an environment friendly way to solve future energy crisis. Although much progress has been made in controlling the heated fusion plasmas (temperature ∼ 150 million degrees) in tokamaks, technological issues related to plasma wall interaction topic still need focused attention. In future, reactor grade tokamak operational scenarios, the reactor wall and target plates are expected to experience a heat load of 10 MW/m 2 and even more during the unfortunate events of ELM's and disruptions. Tungsten remains a suitable choice for the wall and target plates. It can withstand high temperatures, its ductile to brittle temperature is fairly low and it has low sputtering yield and low fuel retention capabilities. However, it is difficult to machine tungsten and hence usages of tungsten coated surfaces are mostly desirable. To produce tungsten coated graphite tiles for the above-mentioned purpose, a coating reactor has been designed, developed and made operational at the SVITS, Indore. Tungsten coating on graphite has been attempted and successfully carried out by using radio frequency induced plasma enhanced chemical vapour deposition (rf -PECVD) for the first time in India. Tungsten hexa-fluoride has been used as a pre-cursor gas. Energy Dispersive X-ray spectroscopy (EDS) clearly showed the presence of tungsten coating on the graphite samples. This paper presents the details of successful operation and achievement of tungsten coating in the reactor at SVITS. (paper)

Tungsten covered graphite and copper elements and ITER-like actively cooled tungsten divertor plasma facing units for the WEST project

International Nuclear Information System (INIS)

Guilhem, D; Bucalossi, J; Burles, S; Corre, Y; Ferlay, F; Firdaouss, M; Languille, P; Lipa, M; Martinez, A; Missirlian, M; Proust, M; Richou, M; Samaille, F; Tsitrone, E

2016-01-01

After a brief introduction giving some insight of the WEST project, we present the three types of plasma facing units (PFUs) developed for the WEST project taking into account the envisaged main scenarios: (1) high power short pulse scenario (a few seconds) where the objective is to maximize the power handling of the PFUs, up to 20 MW m −2 , (2) high fluence scenario (a few 100 s) on actively cooled ITER-like tungsten (W) PFUs, up to 10 MW m −2 during 1000 s. For the graphite PFUs, the high heat flux tests have been done at GLADIS (ion beam test facility), and for the CuCrZr PFUs on the JUDITH (electron beam test facility). The tests were successful, as no damage occurred for the different load cases. This confirms that the modelling done during the design phase is appropriate to describe these PFUs. Series productions are expected to be achieved by the end of 2015 for the graphite and CuCrZr PFUs, and few ITER-like W PFUs are expected at the beginning of 2016. The lower divertor will be complemented with ITER-like W PFUs as soon as available from our partners so that different fabrication procedures could be evaluated in a real industrial process and a real tokamak environment. (paper)

Influence of tungsten microstructure and ion flux on deuterium plasma-induced surface modifications and deuterium retention

Energy Technology Data Exchange (ETDEWEB)

Buzi, Luxherta [IEK - Plasmaphysik, Forschungszentrum Juelich GmbH, Association EURATOM-FZJ, Juelich (Germany); FOM Institute DIFFER-Dutch Institute for Fundamental Energy Research (Netherlands); Ghent University (Belgium); Temmerman, Greg de [FOM Institute DIFFER-Dutch Institute for Fundamental Energy Research (Netherlands); Reinhart, Michael; Matveev, Dmitry; Unterberg, Bernhard; Wienhold, Peter; Breuer, Uwe; Kreter, Arkadi [IEK - Plasmaphysik, Forschungszentrum Juelich GmbH, Association EURATOM-FZJ, Juelich (Germany); Oost, Guido van [Ghent University (Belgium)

2014-07-01

Tungsten is to be used as plasma-facing material for the ITER divertor due to its favourable thermal properties, low erosion and fuel retention. Bombardment of tungsten by low energy ions of hydrogen isotopes, at different surface temperature, can lead to surface modifications and influence the fuel accumulation in the material. This contribution will assess the impact of material microstructure and the correlation between the particle flux, surface modifications and deuterium retention in tungsten. Tungsten samples were exposed to deuterium plasma at a surface temperature of 510 K, 670 K and 870 K, ion energy of 40 eV and ion fluence of 10{sup 26} m{sup -2}. The high and low ion flux ranges were in the order 10{sup 24} m{sup -2}s{sup -1} and 10{sup 22} m{sup -2}s{sup -1}. Depth profiling of deuterium in all the samples was done by secondary ion mass spectroscopy technique and a scanning electron microscope was used to investigate the surface modifications. Modelling of the D desorption spectra with the coupled reaction diffusion system model will be also presented.

Mechanical characterization of W-armoured plasma-facing components after thermal fatigue

International Nuclear Information System (INIS)

Serret, D; Richou, M; Missirlian, M; Loarer, T

2011-01-01

The future fusion device ITER is aimed at demonstrating the scientific and technical feasibility of fusion power. Tens of thousands of W-armoured plasma-facing components (PFCs) will be installed in the vertical targets of the ITER divertor and subjected to a high heat flux. The purpose of this paper is to present the results of mechanical and microstructural characterization of tungsten PFCs after thermal fatigue tests. On each component, Vickers hardness measurements are made. In parallel, the mean grain diameter in the corresponding zone of tungsten material is determined. The empirical Hall-Petch relation was adapted to experimental data. However, due to the plateau effect on recrystallization hardness, this relation does not seem to be relevant once recrystallization is complete: a new approach is proposed for predicting the margin to the tungsten melting onset.

Consolidation of tungsten disilicide by plasma spraying

Czech Academy of Sciences Publication Activity Database

Brožek, Vlastimil; Ctibor, Pavel; Matějíček, Jiří; Rohan, Pavel; Janča, J.

2007-01-01

Roč. 52, č. 3 (2007), s. 311-320 ISSN 0001-7043 R&D Projects: GA ČR(CZ) GA104/05/0540 Institutional research plan: CEZ:AV0Z20430508 Keywords : Water stabilized plasma * tungsten disilicide * plasma deposition * thermal spray coatings Subject RIV: JJ - Other Materials

Characterization of thermomechanical damage on tungsten surfaces during long-duration plasma transients

Energy Technology Data Exchange (ETDEWEB)

Rivera, David, E-mail: david.rivera.ucla@gmail.com; Crosby, Tamer; Sheng, Andrew; Ghoniem, Nasr M.

2014-12-15

A new experimental facility constructed at UCLA for the simulation of high heat flux effects on plasma-facing materials is described. The High Energy Flux Test Facility (HEFTY) is equipped with a Praxair model SG-100 plasma gun, which is nominally rated at 80 kW of continuous operation, of which approximately 30 kW reaches the target due to thermal losses. The gun is used to impart high intermittent heat flux to metal samples mounted within a cylindrical chamber. The system is capable of delivering an instantaneous heat flux in the range of 30–300 MW/m{sup 2}, depending on sample proximity to the gun. The duration of the plasma heat flux is in the range of 1–1000 s, making it ideal for studies of mild plasma transients of relatively long duration. Tungsten and tungsten-copper alloy metal samples are tested in these transient heat flux conditions, and the surface is characterized for damage evaluation using optical, SEM, XRD, and micro-fabrication techniques. Results from a Finite Element (FE) thermo-elastoplasticity model indicate that during the heat-up phase of a plasma transient pulse, the majority of the sample surface is under compressive stresses leading to plastic deformation of the surface. Upon sample cooling, the recovered elastic strain of cooler parts of the sample exceeds that from parts that deformed plastically, resulting in a tensile surface self-stress (residual surface stress). The intensity of the residual tensile surface stress is experimentally correlated with the onset of complex surface fracture morphology on the tungsten surface, and extending below the surface region. Micro-compression mechanical tests of W micro-pillars show that the material has significant plasticity, failing by a “barreling” mode before plasma exposure, and by normal dislocation slip and localized shear after plasma exposure. Ongoing modeling of the complex thermo-fracture process, coupled with elasto-plasticity is based on a phase field approach for distributed

Design and development of a LIBS system on linear plasma device PSI-2 for in situ real-time diagnostics of plasma-facing materials

Directory of Open Access Journals (Sweden)

X. Jiang

2017-08-01

Full Text Available Laser induced breakdown spectroscopy (LIBS is a strong candidate for detecting and monitoring the H/D/T content on the surface of plasma facing components (PFCs due to its capability of fast direct in situ measurement in extreme environment (e.g., vacuum, magnetic field, long distance, complex geometry. To study the feasibilities and encounter the challenges of LIBS on plasma devices, a LIBS system has been set up on the linear plasma device PSI-2. A number of key parameters including laser energy, the influence of magnetic field and the persistence of laser induced plasma are studied. Real-time measurements of deuterium outgassing on tungsten samples exposed to deuterium plasma of 1025 D/m2 are performed in the first 40–130 min after plasma exposure. The experimental results are compared to the calculations in the literature.

Operation of ICRF antennas in a full tungsten environment in ASDEX Upgrade

Science.gov (United States)

Bobkov, Vl.; Braun, F.; Dux, R.; Giannone, L.; Herrmann, A.; Kallenbach, A.; Müller, H. W.; Neu, R.; Noterdaeme, J.-M.; Pütterich, Th.; Rohde, V.; ASDEX Upgrade Team

2009-06-01

In the 2007 and early part of 2008 experimental campaigns, ASDEX Upgrade operated with full tungsten (W) wall without boronization. Use of ICRF power results in a significant increase of W source. Low temperature conditions at the plasma facing components, achieved by a large clearance between the separatrix and the antenna (>6 cm) and by elevated gas puff rates (>5×1021 s) help to lower W sputtering yield during ICRF. Operation of neighboring ICRF antennas at the phase difference close to -90° can lead to a reduction in the W source. However, a reduction of parallel near-fields by antenna design is needed to further minimize the W source. A relation has been established between the HFSS code calculations predicting a dominant role of box currents in the formation of parallel antenna near-fields and the experiment. The shapes of the measured vertical profile of effective sputtering yields and the calculated sheath driving voltages show a qualitative agreement. This confirms that the existing tools are a good basis to design an improved antenna.

Operation of ICRF antennas in a full tungsten environment in ASDEX Upgrade

International Nuclear Information System (INIS)

Bobkov, Vl.; Braun, F.; Dux, R.; Giannone, L.; Herrmann, A.; Kallenbach, A.; Mueller, H.W.; Neu, R.; Noterdaeme, J.-M.; Puetterich, Th.; Rohde, V.

2009-01-01

In the 2007 and early part of 2008 experimental campaigns, ASDEX Upgrade operated with full tungsten (W) wall without boronization. Use of ICRF power results in a significant increase of W source. Low temperature conditions at the plasma facing components, achieved by a large clearance between the separatrix and the antenna (>6 cm) and by elevated gas puff rates (>5x10 21 s -1 ) help to lower W sputtering yield during ICRF. Operation of neighboring ICRF antennas at the phase difference close to -90 deg. can lead to a reduction in the W source. However, a reduction of parallel near-fields by antenna design is needed to further minimize the W source. A relation has been established between the HFSS code calculations predicting a dominant role of box currents in the formation of parallel antenna near-fields and the experiment. The shapes of the measured vertical profile of effective sputtering yields and the calculated sheath driving voltages show a qualitative agreement. This confirms that the existing tools are a good basis to design an improved antenna.

Impact of microstructure on the plasma performance of industrial and high-end tungsten grades

Energy Technology Data Exchange (ETDEWEB)

Pintsuk, G., E-mail: g.pintsuk@fz-juelich.de [Forschungszentrum Jülich, EURATOM Association, 52428 Jülich (Germany); Loewenhoff, Th. [Forschungszentrum Jülich, EURATOM Association, 52428 Jülich (Germany)

2013-07-15

Tungsten and tungsten alloys are actually the primary choice as plasma facing materials for future fusion reactors. Thereby, the material’s response to the different loading conditions occurring in a tokamak is strongly depending on the material properties and therefore the material’s microstructure. This is on the one hand controlled via the manufacturing process and/or the material’s composition and on the other hand by the operational conditions causing recrystallization and melting, and subsequently not only a modified microstructure but also locally a modified composition. The influence of the variation in microstructure is addressed and the pros and cons for using the respective materials and tungsten in general in a fusion environment with steady state and transient thermal loads are outlined. While roughening and the related cracking can hardly be avoided, melting will thwart all efforts to establish a high end microstructure with defined directional properties.

Impact of microstructure on the plasma performance of industrial and high-end tungsten grades

International Nuclear Information System (INIS)

Pintsuk, G.; Loewenhoff, Th.

2013-01-01

Tungsten and tungsten alloys are actually the primary choice as plasma facing materials for future fusion reactors. Thereby, the material’s response to the different loading conditions occurring in a tokamak is strongly depending on the material properties and therefore the material’s microstructure. This is on the one hand controlled via the manufacturing process and/or the material’s composition and on the other hand by the operational conditions causing recrystallization and melting, and subsequently not only a modified microstructure but also locally a modified composition. The influence of the variation in microstructure is addressed and the pros and cons for using the respective materials and tungsten in general in a fusion environment with steady state and transient thermal loads are outlined. While roughening and the related cracking can hardly be avoided, melting will thwart all efforts to establish a high end microstructure with defined directional properties

Ultrafine tungsten as a plasma-facing component in fusion devices: effect of high flux, high fluence low energy helium irradiation

International Nuclear Information System (INIS)

El-Atwani, O.; Gonderman, Sean; Allain, J.P.; Efe, Mert; Klenosky, Daniel; Qiu, Tian; De Temmerman, Gregory; Morgan, Thomas; Bystrov, Kirill

2014-01-01

This work discusses the response of ultrafine-grained tungsten materials to high-flux, high-fluence, low energy pure He irradiation. Ultrafine-grained tungsten samples were exposed in the Pilot-PSI (Westerhout et al 2007 Phys. Scr. T128 18) linear plasma device at the Dutch Institute for Fundamental Energy Research (DIFFER) in Nieuwegein, the Netherlands. The He flux on the tungsten samples ranged from 1.0 × 10 23 –2.0 × 10 24  ions m −2  s −1 , the sample bias ranged from a negative (20–65) V, and the sample temperatures ranged from 600–1500 °C. SEM analysis of the exposed samples clearly shows that ultrafine-grained tungsten materials have a greater fluence threshold to the formation of fuzz by an order or magnitude or more, supporting the conjecture that grain boundaries play a major role in the mechanisms of radiation damage. Pre-fuzz damage analysis is addressed, as in the role of grain orientation on structure formation. Grains of (1 1 0) and (1 1 1) orientation showed only pore formation, while (0 0 1) oriented grains showed ripples (higher structures) decorated with pores. Blistering at the grain boundaries is also observed in this case. In situ TEM analysis during irradiation revealed facetted bubble formation at the grain boundaries likely responsible for blistering at this location. The results could have significant implications for future plasma-burning fusion devices given the He-induced damage could lead to macroscopic dust emission into the fusion plasma. (paper)

Erosion and migration of tungsten employed at the central column heat shield of ASDEX Upgrade

International Nuclear Information System (INIS)

Krieger, K.; Gong, X.; Balden, M.; Hildebrandt, D.; Maier, H.; Rohde, V.; Roth, J.; Schneider, W.

2002-01-01

In ASDEX Upgrade, tungsten was employed as plasma facing material at the central column heat shield in the plasma main chamber. The campaign averaged tungsten erosion flux was determined by measuring the difference of the W-layer thickness before and after the experimental campaign using ion beam analysis methods. The observed lateral variation and the total amount of eroded tungsten are attributed to erosion by impact of ions from the scrape-off layer plasma. Migration and redeposition of eroded tungsten were investigated by quantitative analysis of deposited tungsten on collector probes and wall samples. The obtained results, as well as the spectroscopically observed low tungsten plasma penetration probability, indicate that a major fraction of the eroded tungsten migrates predominantly through direct transport channels in the outer plasma scrape-off layer without entering the confined plasma

HRP facility for fabrication of ITER vertical target divertor full scale plasma facing units

International Nuclear Information System (INIS)

Visca, Eliseo; Roccella, S.; Candura, D.; Palermo, M.; Rossi, P.; Pizzuto, A.; Sanguinetti, G.P.; Mancini, A.; Verdini, L.; Cacciotti, E.; Cerri, V.; Mugnaini, G.; Reale, A.; Giacomi, G.

2015-01-01

Highlights: • R&D activities for the manufacturing of ITER divertor high heat flux plasma-facing components (HHFC). • ENEA and Ansaldo have jointly manufactured several actively cooled monoblock mock-ups and prototypical components. • ENEA and ANSALDO NUCLEARE jointly participate to the European program for the qualification of the manufacturing technology for the ITER divertor IVT. • Successful manufacturing by HRP (Hot Radial Pressing) of first full-scale full-W armored IVT qualification prototype. - Abstract: ENEA and Ansaldo Nucleare S.p.A. (ANN) have being deeply involved in the European development activities for the manufacturing of the ITER Divertor Inner Vertical Target (IVT) plasma-facing components. During normal operation the heat flux deposited on the bottom segment of divertor is 5–10 MW/m 2 but the capability to remove up to 20 MW/m 2 during transient events of 10 s must also be demonstrated. In order to fulfill ITER requirements, ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP). The last challenge is now to fabricate full-scale prototypes of the IVT, aimed to be qualified for the next step, i.e. the series production. On the basis of the experience of manufacturing hundreds of small mock-ups, ENEA designed and installed a new suitable HRP facility. The objective of getting a final shaped plasma facing unit (PFU) that satisfies these requirements is an ambitious target because tolerances set by ITER/F4E are very tight. The setting-up of the equipment started with the fabrication of full scale and representative ‘dummies’ in which stainless steel instead of CFC or W was used for monoblocks. The results confirmed that dimensions were compliant with the required tolerances. The paper reports a brief description of the innovative HRP equipment and the dimensional check results after HRP of the first full-scale full-W PFU.

HRP facility for fabrication of ITER vertical target divertor full scale plasma facing units

Energy Technology Data Exchange (ETDEWEB)

Visca, Eliseo, E-mail: eliseo.visca@enea.it [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Roccella, S. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Candura, D.; Palermo, M. [Ansaldo Nucleare S.p.A., Corso Perrone 25, IT-16152 Genova (Italy); Rossi, P.; Pizzuto, A. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Sanguinetti, G.P. [Ansaldo Nucleare S.p.A., Corso Perrone 25, IT-16152 Genova (Italy); Mancini, A.; Verdini, L.; Cacciotti, E.; Cerri, V.; Mugnaini, G.; Reale, A.; Giacomi, G. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy)

2015-10-15

Highlights: • R&D activities for the manufacturing of ITER divertor high heat flux plasma-facing components (HHFC). • ENEA and Ansaldo have jointly manufactured several actively cooled monoblock mock-ups and prototypical components. • ENEA and ANSALDO NUCLEARE jointly participate to the European program for the qualification of the manufacturing technology for the ITER divertor IVT. • Successful manufacturing by HRP (Hot Radial Pressing) of first full-scale full-W armored IVT qualification prototype. - Abstract: ENEA and Ansaldo Nucleare S.p.A. (ANN) have being deeply involved in the European development activities for the manufacturing of the ITER Divertor Inner Vertical Target (IVT) plasma-facing components. During normal operation the heat flux deposited on the bottom segment of divertor is 5–10 MW/m{sup 2} but the capability to remove up to 20 MW/m{sup 2} during transient events of 10 s must also be demonstrated. In order to fulfill ITER requirements, ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP). The last challenge is now to fabricate full-scale prototypes of the IVT, aimed to be qualified for the next step, i.e. the series production. On the basis of the experience of manufacturing hundreds of small mock-ups, ENEA designed and installed a new suitable HRP facility. The objective of getting a final shaped plasma facing unit (PFU) that satisfies these requirements is an ambitious target because tolerances set by ITER/F4E are very tight. The setting-up of the equipment started with the fabrication of full scale and representative ‘dummies’ in which stainless steel instead of CFC or W was used for monoblocks. The results confirmed that dimensions were compliant with the required tolerances. The paper reports a brief description of the innovative HRP equipment and the dimensional check results after HRP of the first full-scale full-W PFU.

Radiative Recombination and Photoionization Data for Tungsten Ions. Electron Structure of Ions in Plasmas

Directory of Open Access Journals (Sweden)

Malvina B. Trzhaskovskaya

2015-05-01

Full Text Available Theoretical studies of tungsten ions in plasmas are presented. New calculations of the radiative recombination and photoionization cross-sections, as well as radiative recombination and radiated power loss rate coefficients have been performed for 54 tungsten ions for the range W6+–W71+. The data are of importance for fusion investigations at the reactor ITER, as well as devices ASDEX Upgrade and EBIT. Calculations are fully relativistic. Electron wave functions are found by the Dirac–Fock method with proper consideration of the electron exchange. All significant multipoles of the radiative field are taken into account. The radiative recombination rates and the radiated power loss rates are determined provided the continuum electron velocity is described by the relativistic Maxwell–Jüttner distribution. The impact of the core electron polarization on the radiative recombination cross-section is estimated for the Ne-like iron ion and for highly-charged tungsten ions within an analytical approximation using the Dirac–Fock electron wave functions. The effect is shown to enhance the radiative recombination cross-sections by ≲20%. The enhancement depends on the photon energy, the principal quantum number of polarized shells and the ion charge. The influence of plasma temperature and density on the electron structure of ions in local thermodynamic equilibrium plasmas is investigated. Results for the iron and uranium ions in dense plasmas are in good agreement with previous calculations. New calculations were performed for the tungsten ion in dense plasmas on the basis of the average-atom model, as well as for the impurity tungsten ion in fusion plasmas using the non-linear self-consistent field screening model. The temperature and density dependence of the ion charge, level energies and populations are considered.

A fatigue lifetime assessment of WEST ITER Like Plasma Facing Unit

International Nuclear Information System (INIS)

Languille, P.; Missirlian, M.; Guilhem, D.; Ferlay, F.; Batal, T.; Bucalossi, J.; Firdaouss, M.; Larroque, S.; Martinez, A.; Richou, M.

2016-01-01

Highlights: • ITER plasma facing component divertor technology is integrated in WEST. • ITER Like attachments in WEST has been optimised. • The ITER Like PFU is compatible with a wide range of plasma scenarios. - Abstract: Based on a monoblock concept (e.g. a tube-in-tile concept), each elementary tungsten plasma facing component (called Plasma-Facing Unit PFU) of the WEST lower divertor follows as closely as possible the same monoblock geometry, materials and bonding technology that is envisaged for ITER. A fatigue simulation of W PFU was used to validate its specific integration into WEST. The complex design, the material heterogeneities and the usage outside operational load design envelope are all possible causes of fatigue failure. This paper shows how the ITER like monoblocks and its U-shaped attachments technology are integrated into the WEST divertor by performing finite element analysis. The WEST lower divertor is designed to withstand 15 MW steady-state of injected power, with peaked heat fluxes up to 20 MW/m 2 . The integration and the design choices of a WEST ITER Like Plasma Facing Unit inside the WEST vacuum chamber is valid for an “expected life time” of repeated inter ELMs thermal steady state (>10 s) cycles and for 300 off-normal vertical displacement events.

A fatigue lifetime assessment of WEST ITER Like Plasma Facing Unit

Energy Technology Data Exchange (ETDEWEB)

Languille, P., E-mail: pascal.languille@gmail.com; Missirlian, M.; Guilhem, D.; Ferlay, F.; Batal, T.; Bucalossi, J.; Firdaouss, M.; Larroque, S.; Martinez, A.; Richou, M.

2016-11-01

Highlights: • ITER plasma facing component divertor technology is integrated in WEST. • ITER Like attachments in WEST has been optimised. • The ITER Like PFU is compatible with a wide range of plasma scenarios. - Abstract: Based on a monoblock concept (e.g. a tube-in-tile concept), each elementary tungsten plasma facing component (called Plasma-Facing Unit PFU) of the WEST lower divertor follows as closely as possible the same monoblock geometry, materials and bonding technology that is envisaged for ITER. A fatigue simulation of W PFU was used to validate its specific integration into WEST. The complex design, the material heterogeneities and the usage outside operational load design envelope are all possible causes of fatigue failure. This paper shows how the ITER like monoblocks and its U-shaped attachments technology are integrated into the WEST divertor by performing finite element analysis. The WEST lower divertor is designed to withstand 15 MW steady-state of injected power, with peaked heat fluxes up to 20 MW/m{sup 2}. The integration and the design choices of a WEST ITER Like Plasma Facing Unit inside the WEST vacuum chamber is valid for an “expected life time” of repeated inter ELMs thermal steady state (>10 s) cycles and for 300 off-normal vertical displacement events.

Experimental study of tungsten transport properties in T-10 plasma

Science.gov (United States)

Krupin, V. A.; Nurgaliev, M. R.; Klyuchnikov, L. A.; Nemets, A. R.; Zemtsov, I. A.; Dnestrovskij, A. Yu.; Sarychev, D. V.; Lisitsa, V. S.; Shurygin, V. A.; Leontiev, D. S.; Borschegovskij, A. A.; Grashin, S. A.; Ryjakov, D. V.; Sergeev, D. S.; Mustafin, N. A.; Trukhin, V. M.; Solomatin, R. Yu.; Tugarinov, S. N.; Naumenko, N. N.

2017-06-01

First experimental results of tungsten transport investigation in OH and ECRH plasmas in the T-10 tokamak with W-limiter and movable Li-limiter are presented. It is shown that tungsten tends to accumulate (a joint process of cumulation and peaking) near the plasma axis in ohmic regimes. The cumulation of W is enhanced in discharges with high values of the parameter γ ={{\\bar{n}}\\text{e}}\\centerdot {{\\bar{Z}}\\text{eff}}\\centerdot I\\text{pl}-1.5 that coincides with accumulation conditions of light and medium impurities in T-10 plasmas. Experiments with Li-limiter show the immeasurable level of Li3+ (0.3-0.5% of n e) of T-10 CXRS diagnostics because of the low inflow of Li with respect to other light impurities. Nevertheless, the strong influence of lithium on inflow of light and tungsten impurities is observed. In discharges with lithized walls, vanishing of light impurities occurs and values of {{Z}\\text{eff}}≈ 1 are obtained. It is also shown that the tungsten density in the plasma center decreases by 15 to 20 times while the W inflow reduces only by 2 to 4 times. In lithized discharges with high γ, the flattening of the tungsten density profile occurs and its central concentration decreases up to 10 times during the on-axis ECRH. This effect is observed together with the increase of the W inflow by 3 to 4 times at the ECRH stage.

Joining of Tungsten Armor Using Functional Gradients

International Nuclear Information System (INIS)

John Scott O'Dell

2006-01-01

The joining of low thermal expansion armor materials such as tungsten to high thermal expansion heat sink materials has been a major problem in plasma facing component (PFC) development. Conventional planar bonding techniques have been unable to withstand the high thermal induced stresses resulting from fabrication and high heat flux testing. During this investigation, innovative functional gradient joints produced using vacuum plasma spray forming techniques have been developed for joining tungsten armor to copper alloy heat sinks. A model was developed to select the optimum gradient architecture. Based on the modeling effort, a 2mm copper rich gradient was selected. Vacuum plasma pray parameters and procedures were then developed to produce the functional gradient joint. Using these techniques, dual cooling channel, medium scale mockups (32mm wide x 400mm length) were produced with vacuum plasma spray formed tungsten armor. The thickness of the tungsten armor was up to 5mm thick. No evidence of debonding at the interface between the heat sink and the vacuum plasma sprayed material was observed.

Characterization of plasma coated tungsten heavy alloy

International Nuclear Information System (INIS)

Bose, A.; Kapoor, D.; Lankford, J. Jr.; Nicholls, A.E.

1996-01-01

The detrimental environmental impact of Depleted Uranium-based penetrators have led to tremendous development efforts in the area of tungsten heavy alloy based penetrators. One line of investigation involves the coating of tungsten heavy alloys with materials that are prone to shear localization. Plasma spraying of Inconel 718 and 4340 steel have been used to deposit dense coatings on tungsten heavy alloy substrates. The aim of the investigation was to characterize the coating primarily in terms of its microstructure and a special push-out test. The paper describes the results of the push-out tests and analyzes some of the possible failure mechanisms by carrying out microstructural characterization of the failed rings obtained from the push out tests

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

Directory of Open Access Journals (Sweden)

C. N. Taylor

2017-05-01

Full Text Available 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.

Displacement disorder and reconstruction of the (001) face of tungsten

International Nuclear Information System (INIS)

Egorushkin, V.E.; Kul'ment'ev, A.I.; Savushkin, E.V.

1992-01-01

The reconstruction of the (001) border of tungsten is examined taking into consideration random static displacements of surface atoms in the high-temperature (1 x 1) phase. A microscopic model is proposed, in which the creation of c(2 x 2) phase is described as a transition of the Jahn-Teller type and an ordering of static displacements. It is shown that displacement disorder induces instability of (001) tungsten with respect to reconstruction. The effect of a uniform electric field on a disordered reconstructing surface is examined. A possible reason is given for pronounced differences in the results of investigations of the structural conversion of the (001) face in tungsten when different experimental methods are used

Experimental mechanistic investigation of the nanostructuring of tungsten with low energy helium plasmas

Energy Technology Data Exchange (ETDEWEB)

Fiflis, P., E-mail: fiflis1@illinois.edu; Connolly, N.; Ruzic, D.N.

2016-12-15

Helium ion bombardment of tungsten at temperatures between approximately one third and one half of its melting point has shown growth of nanostructures colloquially referred to as “fuzz”. The nanostructures take the form of thin tendrils of diameter about 30 nm and grow out of the bulk material. Tungsten will and does compose one of the key materials for plasma facing components (PFCs) in fusion reactors. The formation of nanostructured fuzz layers on PFCs would be detrimental to the performance of the reactor, and must therefore be avoided. Previous experiments have shown evidence that tungsten fuzz is initially grown by loop punching of helium bubbles created in the bulk. However, once the tendrils grow to sufficient length, the tendrils should intercept the entire helium flux, halting the production of fuzz. Fuzz continues to grow though. To increase the understanding of the mechanisms of tungsten fuzz formation, and thereby aid the avoidance of its production, a series of tests were performed to examine the validity of several theories regarding later stage tungsten fuzz growth. Tests showed that the fuzz formation was dependent solely on the bombardment of helium ions, and not on electric fields, or adatom diffusion. Experiments employing a tungsten coated molybdenum sample indicate the presence of a strong mixing layer and strongly suggest that tungsten fuzz growth continues to occur from the bottom up even as the tendrils grow in size. Tests also show a similarity between different metals exposed to helium ion fluxes where the ratio of bubble diameter to tendril diameter is constant.

Experimental mechanistic investigation of the nanostructuring of tungsten with low energy helium plasmas

International Nuclear Information System (INIS)

Fiflis, P.; Connolly, N.; Ruzic, D.N.

2016-01-01

Helium ion bombardment of tungsten at temperatures between approximately one third and one half of its melting point has shown growth of nanostructures colloquially referred to as “fuzz”. The nanostructures take the form of thin tendrils of diameter about 30 nm and grow out of the bulk material. Tungsten will and does compose one of the key materials for plasma facing components (PFCs) in fusion reactors. The formation of nanostructured fuzz layers on PFCs would be detrimental to the performance of the reactor, and must therefore be avoided. Previous experiments have shown evidence that tungsten fuzz is initially grown by loop punching of helium bubbles created in the bulk. However, once the tendrils grow to sufficient length, the tendrils should intercept the entire helium flux, halting the production of fuzz. Fuzz continues to grow though. To increase the understanding of the mechanisms of tungsten fuzz formation, and thereby aid the avoidance of its production, a series of tests were performed to examine the validity of several theories regarding later stage tungsten fuzz growth. Tests showed that the fuzz formation was dependent solely on the bombardment of helium ions, and not on electric fields, or adatom diffusion. Experiments employing a tungsten coated molybdenum sample indicate the presence of a strong mixing layer and strongly suggest that tungsten fuzz growth continues to occur from the bottom up even as the tendrils grow in size. Tests also show a similarity between different metals exposed to helium ion fluxes where the ratio of bubble diameter to tendril diameter is constant.

Manufacturing and testing in reactor relevant conditions of brazed plasma facing components of the ITER divertor

International Nuclear Information System (INIS)

Bisio, M.; Branca, V.; Marco, M. Di; Federici, A.; Grattarola, M.; Gualco, G.; Guarnone, P.; Luconi, U.; Merola, M.; Ozzano, C.; Pasquale, G.; Poggi, P.; Rizzo, S.; Varone, F.

2005-01-01

A fabrication route based on brazing technology has been developed for the realization of the high heat flux components for the ITER vertical target and Dome-Liner. The divertor vertical target is armoured with carbon fiber reinforced carbon and tungsten in the lower straight part and in the upper curved part, respectively. The armour material is joined to heat sinks made of precipitation hardened copper-chromium-zirconium alloy. The plasma facing units of the dome component are based on a tungsten flat tile design with hypervapotron cooling. An innovative brazing technique based on the addition of carbon fibers to the active brazing alloy, developed by Ansaldo Ricerche for applications in the field of the energy production, has been used for the carbon fiber composite to copper joint to reduce residual stresses. The tungsten-copper joint has been realized by direct casting. A proper brazing thermal cycle has been studied to guarantee the required mechanical properties of the precipitation hardened alloy after brazing. The fabrication route of plasma facing components for the ITER vertical target and dome based on the brazing technology has been proved by means of thermal fatigue tests performed on mock-ups in reactor relevant conditions

Tungsten transport and sources control in JET ITER-like wall H-mode plasmas

Energy Technology Data Exchange (ETDEWEB)

Fedorczak, N., E-mail: nicolas.fedorczak@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Monier-Garbet, P. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Pütterich, T. [MPI für Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, 85748 Garching (Germany); Brezinsek, S. [Institute of Energy and Climate Research, Forschungszentrum Jlich, Assoc EURATOM-FZJ, Jlich (Germany); Devynck, P.; Dumont, R.; Goniche, M.; Joffrin, E. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Lerche, E. [Association EURATOM-Belgian State, LPP-ERM-KMS, TEC partner, Brussels (Belgium); Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Lipschultz, B. [York Plasma Institute, University of York, Heslington, York YO10 5DD (United Kingdom); Luna, E. de la [Laboratorio Nacional de Fusin, Asociacin EURATOM/CIEMAT, 28040 Madrid (Spain); Maddison, G. [Culham Centre for Fusion Energy, EURATOM-CCFE Association, Abingdon (United Kingdom); Maggi, C. [MPI für Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, 85748 Garching (Germany); Matthews, G. [Culham Centre for Fusion Energy, EURATOM-CCFE Association, Abingdon (United Kingdom); Nunes, I. [Istituto de plasmas e fusao nuclear, Lisboa (Portugal); Rimini, F. [Culham Centre for Fusion Energy, EURATOM-CCFE Association, Abingdon (United Kingdom); Solano, E.R. [Laboratorio Nacional de Fusin, Asociacin EURATOM/CIEMAT, 28040 Madrid (Spain); Tamain, P. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Tsalas, M. [Association EURATOM-Hellenic Republic, NCSR Demokritos 153 10, Attica (Greece); Vries, P. de [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

2015-08-15

A set of discharges performed with the JET ITER-like wall is investigated with respect to control capabilities on tungsten sources and transport. In attached divertor regimes, increasing fueling by gas puff results in higher divertor recycling ion flux, lower divertor tungsten source, higher ELM frequency and lower core plasma radiation, dominated by tungsten ions. Both pedestal flushing by ELMs and divertor screening (including redeposition) are possibly responsible. For specific scenarios, kicks in plasma vertical position can be employed to increase the ELM frequency, which results in slightly lower core radiation. The application of ion cyclotron radio frequency heating at the very center of the plasma is efficient to increase the core electron temperature gradient and flatten electron density profile, resulting in a significantly lower central tungsten peaking. Beryllium evaporation in the main chamber did not reduce the local divertor tungsten source whereas core radiation was reduced by approximately 50%.

Hydrogen retention in carbon-tungsten co-deposition layer formed by hydrogen RF plasma

International Nuclear Information System (INIS)

Katayama, K.; Kawasaki, T.; Manabe, Y.; Nagase, H.; Takeishi, T.; Nishikawa, M.

2006-01-01

Carbon-tungsten co-deposition layers (C-W layers) were formed by sputtering method using hydrogen or deuterium RF plasma. The deposition rate of the C-W layer by deuterium plasma was faster than that by hydrogen plasma, where the increase of deposition rate of tungsten was larger than that of carbon. This indicates that the isotope effect on sputtering-depositing process for tungsten is larger than that for carbon. The release curve of hydrogen from the C-W layer showed two peaks at 400 deg. C and 700 deg. C. Comparing the hydrogen release from the carbon deposition layer and the tungsten deposition layer, it is considered that the increase of the release rate at 400 deg. C is affected by tungsten and that at 700 deg. C is affected by carbon. The obtained hydrogen retention in the C-W layers which have over 60 at.% of carbon was in the range between 0.45 and 0.16 as H/(C + W)

Near-surface thermal characterization of plasma facing components using the 3-omega method

International Nuclear Information System (INIS)

Dechaumphai, Edward; Barton, Joseph L.; Tesmer, Joseph R.; Moon, Jaeyun; Wang, Yongqiang; Tynan, George R.; Doerner, Russell P.; Chen, Renkun

2014-01-01

Near-surface regime plays an important role in thermal management of plasma facing components in fusion reactors. Here, we applied a technique referred to as the ‘3ω’ method to measure the thermal conductivity of near-surface regimes damaged by ion irradiation. By modulating the frequency of the heating current in a micro-fabricated heater strip, the technique enables the probing of near-surface thermal properties. The technique was applied to measure the thermal conductivity of a thin ion-irradiated layer on a tungsten substrate, which was found to decrease by nearly 60% relative to pristine tungsten for a Cu ion dosage of 0.2 dpa

Review on the explosive consolidation methods to fabricate tungsten based PFMs

Energy Technology Data Exchange (ETDEWEB)

Wang, Shuming, E-mail: wangshuming@ustb.edu.cn; Sun, Chongxiao; Guo, Wenhao; Yan, Qingzhi; Zhou, Zhangjian; Zhang, Yingchun; Shen, Weiping; Ge, Changchun

2014-12-15

Tungsten is one of the best candidates for plasma-facing materials in the fusion reactors, owing to its many unique properties. In the development of tungsten-based Plasma Facing Materials/Components (PFMs/PFCs), materials scientists have explored many different, innovative preparation and processing routes to meet the requirement of International Thermonuclear Experimental Reactor (ITER). Some explosive consolidation technology intrinsic characteristics, which make it suitable for powder metallurgy (powders consolidation) and PFMs production, are the high pressure processing, highly short heating time and can be considered as a highly competitive green technology. In this work, an overview of explosive consolidation techniques applied to fabricate tungsten-based PFMs is presented. Emphasis is given to describe the main characteristics and potentialities of the explosive sintering, explosive consolidation techniques. The aspects presented and discussed in this paper indicate the explosive consolidation processes as a promising and competitive technology for tungsten-based PFMs processing.

RF induction plasma spheroidization of tungsten powders

International Nuclear Information System (INIS)

Gu Zhogntao; Ye Gaoying; Liu Chuandong; Tong Honghui

2009-01-01

Irregularly-shaped tungsten powders (average granular sizes of 512 μm) have been spheroidized by radio frequency (RF)induction plasma. The effects of feed rate, mode of material dispersion, particle size on spheroidization efficiency are investigated. Experimental results show that the spheroidization efficiency decreases rapidly when the feed rate increases to more than 95 g/min. Only 30% spheroidization efficiency is gained at the feed rate of 135.75 g/min. The spheroidization efficiency is also affected by the flow rate of carrier gas. When the flow rate of carrier gas is 0.12 m 3 /h, the dispersion effect is the best, and the spheroidization efficiency is almost 100%. The apparent density of tungsten powders increases a bit with the increase of spheroidization efficiency. And the particle size uniformity of spheroidized tungsten powders is in accordance with that of original powders. (authors)

Evidences of trapping in tungsten and implications for plasma-facing components

Science.gov (United States)

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

Trapping effects that include significant delays in permeation saturation, abrupt changes in permeation rate associated with temperature changes, and larger than expected inventories of hydrogen isotopes in the material, were seen in implantation-driven permeation experiments using 25- and 50-micron thick tungsten foils at temperatures of 638 to 825 K. Computer models that simulate permeation transients reproduce the steady-state permeation and reemission behavior of these experiments with expected values of material parameters. However, the transient time characteristics were not successfully simulated without the assumption of traps of substantial trap energy and concentration. An analytical model based on the assumptions of thermodynamic equilibrium between trapped hydrogen atoms and a comparatively low mobile atom concentration successfully accounts for the observed behavior. Using steady-state and transient permeation data from experiments at different temperatures, the effective trap binding energy may be inferred. We analyze a tungsten coated divertor plate design representative of those proposed for ITER and ARIES and consider the implications for tritium permeation and retention if the same trapping we observed was present in that tungsten. Inventory increases of several orders of magnitude may result.

Tungsten: An option for divertor and main chamber plasma facing components in future fusion devices

International Nuclear Information System (INIS)

Neu, R.; Dux, R.; Kallenbach, A.; Maggi, C.F.; Puetterich, T.; Balden, M.; Eich, T.; Fuchs, J.C.; Gruber, O.; Herrmann, A.; Maier, H.; Mueller, H.W.; Pugno, R.; Radivojevic, I.; Rohde, V.; Sips, A.C.C.; Suttrop, W.; Ye, M.Y.; O'Mullane, M.; Whiteford, A.

2005-01-01

The tungsten programme in ASDEX Upgrade is pursued towards a full high-Z device. The spectroscopic diagnostic and the cooling factor of W have been extended and refined. The W-coated surfaces represent now a fraction of 65% (24.8 m2). The only two major components which are not yet coated are the strikepoint region of the lower divertor as well as the limiters at the low field side. While extending the W surfaces, the W concentration and the discharge behaviour have changed gradually pointing to critical issues when operating with a W wall: anomalous transport in the plasma centre should not be too low, otherwise neoclassical accumulation can occur. A very successful remedy is the addition of central RF heating at the 20-30% level. Regimes with low ELM activity show increased impurity concentration over the whole plasma radius. These discharges can be cured by increasing the ELM frequency through pellet ELM pacemaking or by higher heating power. Moderate gas puffing also mitigates the impurity influx and penetration, however at the expense of lower confinement. The erosion yield at the low field side guard limiter can be as high as 10 -3 and fast particle losses from NBI were identified to contribute a significant part to the W sputtering. Discharges run in the upper, W coated divertor do not show higher W concentrations than comparable discharges in the lower C-based divertor. (author)

Thermal response of plasma sprayed tungsten coating to high heat flux

International Nuclear Information System (INIS)

Liu, X.; Yang, L.; Tamura, S.; Tokunaga, K.; Yoshida, N.; Noda, N.; Xu, Z.

2004-01-01

In order to investigate the thermal response of tungsten coating on carbon and copper substrates by vacuum plasma spray (VPS) or inert gas plasma spray (IPS), annealing and cyclic heat load experiments of these coatings were conducted. It is indicated that the multi-layered tungsten and rhenium interface of VPS-W/CFC failed to act as a diffusion barrier at elevated temperature and tungsten carbides were developed after 1 h incubation time when annealing temperature was higher than 1600 deg. C. IPS-W/Cu and W/C without an intermediate bonding layer were failed by the detachment of the tungsten coating at 900 and 1200 deg. C annealing for several hours, respectively. Cyclic heat load of electron beam with 35 MW/m 2 and 3-s pulse duration indicated that IPS-W/Cu samples failed with local detachment of the tungsten coating within 200 cycles and IPS-W/C showed local cracks by 300 cycles, but VPS-W/CFC withstood 1000 cycles without visible damages. However, crack creation and propagation in VPS-W/CFC were also observed under higher heat load

Experimental setup for producing tungsten coated graphite tiles using plasma enhanced chemical vapor deposition technique for fusion plasma applications

International Nuclear Information System (INIS)

Chauhan, Sachin Singh; Sharma, Uttam; Choudhary, K.K.; Sanyasi, A.K.; Ghosh, J.; Sharma, Jayshree

2013-01-01

Plasma wall interaction (PWI) in fusion grade machines puts stringent demands on the choice of materials in terms of high heat load handling capabilities and low sputtering yields. Choice of suitable material still remains a challenge and open topic of research for the PWI community. Carbon fibre composites (CFC), Beryllium (Be), and Tungsten (W) are now being considered as first runners for the first wall components of future fusion machines. Tungsten is considered to be one of the suitable materials for the job because of its superior properties than carbon like low physical sputtering yield and high sputter energy threshold, high melting point, fairly high re-crystallization temperature, low fuel retention capabilities, low chemical sputtering with hydrogen and its isotopes and most importantly the reparability with various plasma techniques both ex-situ and in-situ. Plasma assisted chemical vapour deposition is considered among various techniques as the most preferable technique for fabricating tungsten coated graphite tiles to be used as tokamak first wall and target components. These coated tiles are more favourable compared to pure tungsten due to their light weight and easier machining. A system has been designed, fabricated and installed at SVITS, Indore for producing tungsten coated graphite tiles using Plasma Enhanced Chemical Vapor Deposition (PE-CVD) technique for Fusion plasma applications. The system contains a vacuum chamber, a turbo-molecular pump, two electrodes, vacuum gauges, mass analyzer, mass flow controllers and a RF power supply for producing the plasma using hydrogen gas. The graphite tiles will be put on one of the electrodes and WF6 gas will be inserted in a controlled manner in the hydrogen plasma to achieve the tungsten-coating with WF6 dissociation. The system is integrated at SVITS, Indore and a vacuum of the order of 3*10 -6 is achieved and glow discharge plasma has been created to test all the sub-systems. The system design with

Generation and analysis of plasmas with centrally reduced helicity in full-tungsten ASDEX upgrade

Energy Technology Data Exchange (ETDEWEB)

Bock, Alexander

2016-03-01

The most promising concepts for harnessing nuclear fusion are toroidal devices like tokamaks, where a plasma is confined by helically twisted magnetic field lines. To provide the twisting of the field lines, a tokamak relies on a toroidal current in the plasma, which is largely generated by a transformer. As such, conventional tokamaks are limited to pulsed operation. Moreover, this current makes tokamak plasmas prone to numerous confinement degrading magnetohydrodynamic (MHD) instabilities that can emerge at locations where the field line helicity q takes on rational values like 1/1, 3/2 or 2/1, i.e. sawteeth or neoclassical tearing modes (NTMs). This thesis presents studies of plasmas with centrally elevated q-profiles created by external electron-cyclotron and neutral beam current drive (ECCD/NBCD) under steady-state conditions in the full-tungsten tokamak ASDEX Upgrade. Without the usually monotonic q-profile, instabilities of low helicity disappear, thereby improving the plasma stability. Furthermore, elevating q increases the amount of so-called (toroidal) bootstrap current, which the plasma drives by itself in the presence of pressure gradients, thereby reducing the reliance on the transformer. In the best case, an advanced tokamak (AT) could thus run in steady state. Additionally, an elevated and thus flat/slightly reversed q-profile is thought to improve confinement by impeding turbulent transport. Reconstruction of the tailored q-profile is accomplished with the new integrated data equilibrium (IDE) code and information from a key diagnostic that is based on the Motional Stark Effect (MSE). During the course of this work it was discovered that the MSE diagnostic suffers from interference from polarised background light. A prototype mitigation system was successfully tested. Also, non-linearities in the diagnostic's optical relay system were found and a calibration scheme devised to take them into account. Both the conventional approach of AT

Deuterium Depth Profile in Neutron-Irradiated Tungsten Exposed to Plasma

International Nuclear Information System (INIS)

Shimada, Masashi; Cao, G.; Hatano, Y.; Oda, T.; Oya, Y.; Hara, M.; Calderoni, P.

2011-01-01

The effect of radiation damage has been mainly simulated using high-energy ion bombardment. The ions, however, are limited in range to only a few microns into the surface. Hence, some uncertainty remains about the increase of trapping at radiation damage produced by 14 MeV fusion neutrons, which penetrate much farther into the bulk material. With the Japan-US joint research project: Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), the tungsten samples (99.99 % pure from A.L.M.T., 6mm in diameter, 0.2mm in thickness) were irradiated to high flux neutrons at 50 C and to 0.025 dpa in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL). Subsequently, the neutron-irradiated tungsten samples were exposed to a high-flux deuterium plasma (ion flux: 1021-1022 m-2s-1, ion fluence: 1025-1026 m-2) in the Tritium Plasma Experiment (TPE) at the Idaho National Laboratory (INL). First results of deuterium retention in neutron-irradiated tungsten exposed in TPE have been reported previously. This paper presents the latest results in our on-going work of deuterium depth profiling in neutron-irradiated tungsten via nuclear reaction analysis. The experimental data is compared with the result from non neutron-irradiated tungsten, and is analyzed with the Tritium Migration Analysis Program (TMAP) to elucidate the hydrogen isotope behavior such as retention and depth distribution in neutron-irradiated and non neutron-irradiated tungsten.

Tungsten coatings electro-deposited on CFC substrates from oxide molten salt

Energy Technology Data Exchange (ETDEWEB)

Sun, Ningbo; Zhang, Yingchun, E-mail: zycustb@163.com; Lang, Shaoting; Jiang, Fan; Wang, Lili

2014-12-15

Tungsten is considered as plasma facing material in fusion devices because of its high melting point, its good thermal conductivity, its low erosion rate and its benign neutron activation properties. On the other hand, carbon based materials like C/C fiber composites (CFC) have been used for plasma facing materials (PFMs) due to their high thermal shock resistance, light weight and high strength. Tungsten coatings on CFC substrates are used in the JET divertor in the frame of the JET ITER-like wall project, and have been prepared by plasma spray (PS) and other techniques. In this study, tungsten coatings were electro-deposited on CFC from Na{sub 2}WO{sub 4}–WO{sub 3} molten salt under various deposition parameters at 900 °C in air. In order to obtain tungsten coatings with excellent performance, the effects of pulse duration ratio and pulse current density on microstructures and crystal structures of tungsten coatings were investigated by X-ray diffraction (XRD, Rigaku Industrial Co., Ltd., D/MAX-RB) and a scanning electron microscope (SEM, JSM 6480LV). It is found that the pulsed duration ratio and pulse current density had a significant influence on tungsten nucleation and electro-crystallization phenomena. SEM observation revealed that intact, uniform and dense tungsten coatings formed on the CFC substrates. Both the average grain size and thickness of the coating increased with the pulsed current density. The XRD results showed that the coatings consisted of a single phase of tungsten with the body centered cubic (BCC) structure. The oxygen content of electro-deposited tungsten coatings was lower than 0.05%, and the micro-hardness was about 400 HV.

Arc erosion of full metal plasma facing components at the inner baffle region of ASDEX Upgrade

Directory of Open Access Journals (Sweden)

V. Rohde

2016-12-01

Full Text Available At the inner baffle of the AUG divertor massive polished inserts of tungsten and P92 steel were installed to measure the erosion by arcing. For tungsten most of the traces are less than 0.4µm deep and a similar amount of tungsten is deposited close to the traces. Few craters up to 4µm resulting in an average erosion rate of 2×1013 at cm−2s−1 are observed. The behaviour for P92 steel is quite different: most of the traces are 4µm deep, up to 80µm were observed. An average erosion rate of 400×1013 at cm−2s−1, i.e. more than a factor of hundred higher compared to tungsten, is found. Therefore, erosion by arcing has to be taken into account to determine the optimal material mix for future fusion devices.

Advanced smart tungsten alloys for a future fusion power plant

Science.gov (United States)

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.

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

Science.gov (United States)

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.

Surface hardening induced by high flux plasma in tungsten revealed by nano-indentation

Energy Technology Data Exchange (ETDEWEB)

Terentyev, D., E-mail: dterenty@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium); Bakaeva, A. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium); Department of Applied Physics, Ghent University, St. Pietersnieuwstraat 41, 9000 Ghent (Belgium); Pardoen, T.; Favache, A. [Institute of Mechanics, Materials and Civil Engineering, Université catholique de Louvain, Place Sainte Barbe 2 L5.02.02, 1348 Louvain-la-Neuve (Belgium); Zhurkin, E.E. [Department of Experimental Nuclear Physics K-89, Faculty of Physics and Mechanics, St. Petersburg State Polytechnical University, 29 Polytekhnicheskaya str., 195251 St. Petersburg (Russian Federation)

2016-08-01

Surface hardness of tungsten after high flux deuterium plasma exposure has been characterized by nanoindentation. The effect of plasma exposure was rationalized on the basis of available theoretical models. Resistance to plastic penetration is enhanced within the 100 nm sub-surface region, attributed to the pinning of geometrically necessary dislocations on nanometric deuterium cavities – signature of plasma-induced defects and deuterium retention. Sub-surface extension of thereby registered plasma-induced damage is in excellent agreement with the results of alternative measurements. The study demonstrates suitability of nano-indentation to probe the impact of deposition of plasma-induced defects in tungsten on near surface plasticity under ITER-relevant plasma exposure conditions.

Mechanical properties of tungsten following rhenium ion and helium plasma exposure

Directory of Open Access Journals (Sweden)

C.S. Corr

2017-08-01

Full Text Available Mechanical properties of Tungsten (W samples irradiated with 2 MeV Rhenium (Re ions and helium (He plasma were investigated using nanoindentation. It was found that there was an increase in hardness for all samples following separate irradiation with both Re ion and He plasma. A slight increase in hardness was obtained for combined exposures. A comparable increase in hardness was observed for a pure He plasma with a sample temperature of 473 K and 1273 K. Optical interferometry was employed to compare surface modification of the samples. Grazing incidence small angle x-ray scattering confirmed He nano-bubble formation of approximately 1 nm diameter in the higher temperature sample, which was not observed with samples at the lower temperatures.

Armour Materials for the ITER Plasma Facing Components

Science.gov (United States)

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.

Armour materials for the ITER plasma facing components

International Nuclear Information System (INIS)

Barabash, V.; Federici, G.; Matera, R.; Raffray, A.R.

1999-01-01

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-a-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 and D. (orig.)

Adsorption of lithium-lanthanum films on the (100) tungsten face

International Nuclear Information System (INIS)

Gupalo, M.S.; Smereka, T.P.; Babkin, G.V.; Palyukh, B.M.

1982-01-01

The method of contact potential difference is used to investigate combined adsorption of lithium-lanthanum on the (100) tungsten face. The data on work functions and thermal stability of mixed lithium-lanthanum films are obtained. The presence of lanthanum on the W(100) surface leads to appearance of minimum of work functions unobserved for the Li-W(100) system, minimum work functions and optimum lithium concentration in a mixed film are decreased at initial lanthanum coating increase. The presence of lanthanum on the W(100) face leads to lithium adsorption heat decrease

Transient induced tungsten melting at the Joint European Torus (JET).

Czech Academy of Sciences Publication Activity Database

Coenen, J.W.; Matthews, G.F.; Krieger, K.; Iglesias, D.; Bunting, P.; Corre, Y.; Silburn, S.; Balboa, I.; Bazylev, B.; Conway, N.; Coffey, I.; Dejarnac, Renaud; Gauthier, E.; Gaspar, J.; Jachmich, S.; Jepu, I.; Makepeace, C.; Scannell, R.; Stamp, M.; Petersson, P.; Pitts, R.A.; Wiesen, S.; Widdowson, A.; Heinola, K.; Baron-Wiechec, A.

T170, December (2017), č. článku 014013. ISSN 0031-8949. [PFMC 2017: 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications. Düsseldorf, 16.05.2017-19.05.2017] EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : fusion * melting * plasma wall interaction * tungsten * plasma facing components Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 1.3 Physical sciences Impact factor: 1.280, year: 2016 http://iopscience.iop.org/article/10.1088/1402-4896/aa8789/meta

OEDGE Modeling of Collector Probe measurements in L-mode from the DIII-D tungsten ring campaign

Science.gov (United States)

Elder, J. D.; Stangeby, P. C.; Unterberg, Z.; Donovan, D.; Wampler, W. R.; Watkins, J.; Abrams, T.; McLean, A. G.

2017-10-01

During the tungsten ring campaign on DIII-D, a collector probe system with multiple diameter, dual-facing collector rods was inserted into the far scrape off layer (SOL) near the outer midplane to measure the plasma tungsten content. For most probes more tungsten was observed on the side connected along field lines to the inner divertor, with the larger probes showing largest divertor-facing asymmetries The OEDGE code is used to model the tungsten erosion, transport and deposition. It has been enhanced with (i) a peripheral particle transport and deposition model to record the impurity content in the peripheral region outside the regular mesh, and (ii) a collector probe model. The OEDGE results approximately reproduce both the divertor-facing asymmetries and the radial decay of each collector probe profile. The effect of changing impurity transport assumptions and wall location are examined. The measured divertor-facing asymmetries imply a higher tungsten density in the plasma upstream of the probe; this was expected theoretically from the effect of the parallel ion temperature gradient force driving upstream transport of tungsten from the outer divertor and was also found in the code analysis. Work supported by the US Department of Energy under DE-FC02-04ER54698, DE-NA0003525, DE-AC05-00OR22725, and DE-AC52-07NA27344.

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

Science.gov (United States)

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.

Investigation of Plasma Facing Components in Plasma Focus Operation

Science.gov (United States)

Roshan, M. V.; Babazadeh, A. R.; Kiai, S. M. Sadat; Habibi, H.; Mamarzadeh, M.

2007-09-01

Both aspects of the plasma-wall interactions, counter effect of plasma and materials, have been considered in our experiments. The AEOI plasma focus, Dena, has Filippov-type electrodes. The experimental results verify that neutron production increases using tungsten as an anode insert material, compared to the copper one. The experiments show decrement of the hardness of Aluminum targets outward the sides, from 135 to 78 in Vickers scale. The sputtering yield is about 0.0065 for deuteron energy of 50 keV.

Research status and issues of tungsten plasma facing materials for ITER and beyond

International Nuclear Information System (INIS)

Ueda, Y.; Coenen, J.W.; De Temmerman, G.; Doerner, R.P.; Linke, J.; Philipps, V.; Tsitrone, E.

2014-01-01

This review summarizes surface morphology changes of tungsten caused by heat and particle loadings from edge plasmas, and their effects on enhanced erosion and material lifetime in ITER and beyond. Pulsed heat loadings by transients (disruption and ELM) are the largest concerns due to surface melting, cracking, and dust formation. Hydrogen induced blistering is unlikely to be an issue of ITER. Helium bombardment would cause surface morphology changes such as W fuzz, He holes, and nanometric bubble layers, which could lead to enhanced erosion (e.g. unipolar arcing of W fuzz). Particle loadings could enhance pulsed heat effects (cracking and erosion) due to surface layer embrittlement by nanometric bubbles and solute atoms. But pulsed heat loadings alleviate surfaces morphology changes in some cases (He holes by ELM-like heat pulses). Effects of extremely high fluence (∼10 30 m −2 ), mixed materials, and neutron irradiation are important issues to be pursued for ITER and beyond. In addition, surface refurbishment to prolong material lifetime is also an important issue

Plasma spraying of zirconium carbide – hafnium carbide – tungsten cermets

Czech Academy of Sciences Publication Activity Database

Brožek, Vlastimil; Ctibor, Pavel; Cheong, D.-I.; Yang, S.-H.

2009-01-01

Roč. 9, č. 1 (2009), s. 49-64 ISSN 1335-8987 Institutional research plan: CEZ:AV0Z20430508 Keywords : Plasma spraying * cermet coatings * microhardness * zirconium carbide * hafnium carbide * tungsten * water stabilized plasma Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass

Development of positron annihilation spectroscopy for investigating deuterium decorated voids in neutron-irradiated tungsten

Energy Technology Data Exchange (ETDEWEB)

Taylor, C.N., E-mail: chase.taylor@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Shimada, M.; Merrill, B.J. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Akers, D.W. [Experimental Programs, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Hatano, Y. [Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555 (Japan)

2015-08-15

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

Development of positron annihilation spectroscopy for investigating deuterium decorated voids in neutron-irradiated tungsten

Science.gov (United States)

Taylor, C. N.; Shimada, M.; Merrill, B. J.; Akers, D. W.; Hatano, Y.

2015-08-01

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

Chemically deposited tungsten fibre-reinforced tungsten – The way to a mock-up for divertor applications

Directory of Open Access Journals (Sweden)

J. Riesch

2016-12-01

Full Text Available The development of advanced materials is essential for sophisticated energy systems like a future fusion reactor. Tungsten fibre-reinforced tungsten composites (Wf/W utilize extrinsic toughening mechanisms and therefore overcome the intrinsic brittleness of tungsten at low temperature and its sensitivity to operational embrittlement. This material has been successfully produced and tested during the last years and the focus is now put on the technological realisation for the use in plasma facing components of fusion devices. In this contribution, we present a way to utilize Wf/W composites for divertor applications by a fabrication route based on the chemical vapour deposition (CVD of tungsten. Mock-ups based on the ITER typical design can be realized by the implementation of Wf/W tiles. A concept based on a layered deposition approach allows the production of such tiles in the required geometry. One fibre layer after the other is positioned and ingrown into the W-matrix until the final sample size is reached. Charpy impact tests on these samples showed an increased fracture energy mainly due to the ductile deformation of the tungsten fibres. The use of Wf/W could broaden the operation temperature window of tungsten significantly and mitigate problems of deep cracking occurring typically in cyclic high heat flux loading. Textile techniques are utilized to optimise the tungsten wire positioning and process speed of preform production. A new device dedicated to the chemical deposition of W enhances significantly, the available machine time for processing and optimisation. Modelling shows that good deposition results are achievable by the use of a convectional flow and a directed temperature profile in an infiltration process.

Deuterium-induced nanostructure formation on tungsten exposed to high-flux plasma

NARCIS (Netherlands)

Xu, H.Y.; De Temmerman, G.C.; Luo, G.-N.; Jia, Y.Z.; Yuan, Y.; Fu, B.Q.; Godfrey, A.; Liu, W.

2015-01-01

PLASMA-SURFACE INTERACTIONS 21 — Proceedings of the 21st International Conference on Plasma-Surface Interactions in Controlled Fusion Devices Kanazawa, Japan May 26-30, 2014 Surface topography of polycrystalline tungsten (W) have been examined after exposure to a low-energy (38 eV/D), high-flux

ELM-induced melting: assessment of shallow melt layer damage and the power handling capability of tungsten in a linear plasma device

Czech Academy of Sciences Publication Activity Database

Morgan, T.W.; van Eden, G.G.; de Kruif, T.M.; van den Berg, A.; Matějíček, Jiří; Chráska, Tomáš; De Temmerman, G.

-, T159 (2014), 014022-014022 ISSN 0031-8949. [International Conference on Plasma-Facing Materials and Components for Fusion Applications/14./. Jülich, 13.05.2013-17.05.2013] Institutional support: RVO:61389021 Keywords : melting * tungsten * ELMs * divertor * ITER * DEMO Subject RIV: JG - Metallurgy Impact factor: 1.126, year: 2014 http://iopscience.iop.org/1402-4896/2014/T159/014022/pdf/1402-4896_2014_T159_014022.pdf

Incorporation of tungsten metal fibers in a metal and ceramic matrix

Directory of Open Access Journals (Sweden)

V. Brozek

2017-01-01

Full Text Available Tungsten fibers have high tensile strength but a poor oxidation resistance at elevated temperatures. Using this first characteristic and to prevent oxidation of tungsten coated composite materials in which the primary requirement: reinforcement against destruction or deformation, was studied on tungsten fibers and tungsten wires which were coated by applying the metal and ceramic powders via plasma spraying device in plasma generator WSP®. Deposition took place in an atmosphere of Ar + 7 % H2, sufficient to reduce the oxidized trace amounts of tungsten.

Tritium decay helium-3 effects in tungsten

Directory of Open Access Journals (Sweden)

M. Shimada

2017-08-01

Full Text Available Tritium (T implanted by plasmas diffuses into bulk material, especially rapidly at elevated temperatures, and becomes trapped in neutron radiation-induced defects in materials that act as trapping sites for the tritium. The trapped tritium atoms will decay to produce helium-3 (3He atoms at a half-life of 12.3 years. 3He has a large cross section for absorbing thermal neutrons, which after absorbing a neutron produces hydrogen (H and tritium ions with a combined kinetic energy of 0.76 MeV through the 3He(n,HT nuclear reaction. The purpose of this paper is to quantify the 3He produced in tungsten by tritium decay compared to the neutron-induced helium-4 (4He produced in tungsten. This is important given the fact that helium in materials not only creates microstructural damage in the bulk of the material but alters surface morphology of the material effecting plasma-surface interaction process (e.g. material evolution, erosion and tritium behavior of plasma-facing component materials. Effects of tritium decay 3He in tungsten are investigated here with a simple model that predicts quantity of 3He produced in a fusion DEMO FW based on a neutron energy spectrum found in literature. This study reveals that: (1 helium-3 concentration was equilibrated to ∼6% of initial/trapped tritium concentration, (2 tritium concentration remained approximately constant (94% of initial tritium concentration, and (3 displacement damage from 3He(n,HT nuclear reaction became >1 dpa/year in DEMO FW.

Advanced Electrochemical Machining (ECM) for tungsten surface micro-structuring in blanket applications

International Nuclear Information System (INIS)

Holstein, Nils; Krauss, Wolfgang; Konys, Jürgen; Heuer, Simon; Weber, Thomas

2016-01-01

Highlights: • Electrochemical Machining is an appropriate tool for tungsten shaping. • Progress in shaping achieved by combination of ECM with advanced micro-lithography. • Application in First Wall for connection of plasma facing material to breeder blanket. • Successful development of adhesion promotors by ECM for plasma spraying interlayers. • Microstructure electrochemical manufacturing of tungsten in sizes of 100 μm achieved. - Abstract: Plasma facing components for fusion applications must have to exhibit long-term stability under extreme physical conditions, and therefore any material imperfections caused by mechanical and/or thermal stresses in the shaping processes cannot be tolerated due to a high risk of possible technical failures under fusion conditions. To avoid such defects, the method of Electrochemical Machining (ECM) enables a complete defect-free processing of removal of tungsten material during the desired shaping, also for high penetration depths. Furthermore, supported by lithographic mask pretreatment, three-dimensional distinct geometric structures can be positive-imaged via the directional galvanic dissolution applying M-ECM process into the tungsten bulk material. New required applications for tungsten components, e.g. as adhesion promotors in W-surfaces to enable sure grip and bonding of thick plasma-spraying layers for blanket components, will define the way of further miniaturization of well-established millimeter dimensioned M-ECM shaping processes to dimensions of 100 μm and furthermore down to 50 μm. Besides current M-ECM limits the article describes inevitable needs of further developments for mask resists, mask materials and the resulting ECM parameters, to reach the needed accuracy in tungsten microstructure. The achieved progress and observed correlations of processing parameters will be manifested by produced demonstrators made by the new “μM”-ECM process.

Advanced Electrochemical Machining (ECM) for tungsten surface micro-structuring in blanket applications

Energy Technology Data Exchange (ETDEWEB)

Holstein, Nils, E-mail: nils.holstein@kit.edu [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen, Baden-Württemberg (Germany); Krauss, Wolfgang; Konys, Jürgen [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen, Baden-Württemberg (Germany); Heuer, Simon; Weber, Thomas [Research Center Jülich, Institute of Energy- and Climate Research – Plasma Physics (IEK-4), D-52425 Jülich (Germany)

2016-11-01

Highlights: • Electrochemical Machining is an appropriate tool for tungsten shaping. • Progress in shaping achieved by combination of ECM with advanced micro-lithography. • Application in First Wall for connection of plasma facing material to breeder blanket. • Successful development of adhesion promotors by ECM for plasma spraying interlayers. • Microstructure electrochemical manufacturing of tungsten in sizes of 100 μm achieved. - Abstract: Plasma facing components for fusion applications must have to exhibit long-term stability under extreme physical conditions, and therefore any material imperfections caused by mechanical and/or thermal stresses in the shaping processes cannot be tolerated due to a high risk of possible technical failures under fusion conditions. To avoid such defects, the method of Electrochemical Machining (ECM) enables a complete defect-free processing of removal of tungsten material during the desired shaping, also for high penetration depths. Furthermore, supported by lithographic mask pretreatment, three-dimensional distinct geometric structures can be positive-imaged via the directional galvanic dissolution applying M-ECM process into the tungsten bulk material. New required applications for tungsten components, e.g. as adhesion promotors in W-surfaces to enable sure grip and bonding of thick plasma-spraying layers for blanket components, will define the way of further miniaturization of well-established millimeter dimensioned M-ECM shaping processes to dimensions of 100 μm and furthermore down to 50 μm. Besides current M-ECM limits the article describes inevitable needs of further developments for mask resists, mask materials and the resulting ECM parameters, to reach the needed accuracy in tungsten microstructure. The achieved progress and observed correlations of processing parameters will be manifested by produced demonstrators made by the new “μM”-ECM process.

Plasma Wall Interaction Phenomena on Tungsten Armour Materials for Fusion Applications

Energy Technology Data Exchange (ETDEWEB)

Uytdenhouwen, I. [SCK.CEN - The Belgian Nuclear Research Centre, Institute for Nuclear Materials Science, Boeretang 200, 2400 Mol (Belgium); Forschungszentrum Juelich GmbH, EURATOM-association, D-52425 Juelich (Germany); Department of Applied Physics, Ghent University, Rozier 44, 9000 Ghent (Belgium); Massaut, V. [Department of Applied Physics, Ghent University, Rozier 44, 9000 Ghent (Belgium); Linke, J. [Forschungszentrum Juelich GmbH, EURATOM-association, D-52425 Juelich (Germany); Van Oost, G. [Department of Applied Physics, Ghent University, Rozier 44, 9000 Ghent (Belgium)

2008-07-01

One of the most attractive future complements to present energy sources is nuclear fusion. A large progress was made throughout the last decade from both the physical as the technological area leading to the construction of the ITER machine. One of the key issues that recently received a large interest at international level is focused on the Plasma Wall Interaction (PWI). One of the promising Plasma Facing Materials (PFM) are Tungsten (W) and Tungsten alloys. However, despite the worldwide use and industrial availability of W, the database of physical and mechanical properties is very limited. Especially after fusion relevant neutron irradiation and PWI phenomena, most of the properties are still unknown. The plasma fuel consists out of deuterium (D) and tritium (T). Tritium is radio-active and therefore an issue from the safety point of view. During steady-state plasma operation of future fusion power plants, the PFM need to extract a power density of {approx}10-20 MW/m{sup 2}. On top of this heat, transient events will deposit an additional non-negligible amount of energy (Disruptions, Vertical Displacement Events, Edge Localized Modes) during short durations. These severe heat loads cause cracking and even melting of the surface resulting in a reduced lifetime and the creation of dust. A contribution to the understanding of cracking phenomena under the severe thermal loads is described as well as the properties degradation under neutron irradiation. Several W grades were irradiated in the BR2 reactor (SCK.CEN) and the thermal loads were simulated with the electron-beam facility JUDITH (FZJ). Since knowledge should be gained about the Tritium retention in the PFM for safety and licensing reasons, a unique test facility at SCK.CEN is being set-up. The plasmatron VISION-I will simulate steady state plasmas for Tritium retention studies. The formation of surface cracks and dust, the initial porosity, neutron induced traps, re-deposited material - change the Tritium

Plasma Wall Interaction Phenomena on Tungsten Armour Materials for Fusion Applications

International Nuclear Information System (INIS)

Uytdenhouwen, I.; Massaut, V.; Linke, J.; Van Oost, G.

2008-01-01

One of the most attractive future complements to present energy sources is nuclear fusion. A large progress was made throughout the last decade from both the physical as the technological area leading to the construction of the ITER machine. One of the key issues that recently received a large interest at international level is focused on the Plasma Wall Interaction (PWI). One of the promising Plasma Facing Materials (PFM) are Tungsten (W) and Tungsten alloys. However, despite the worldwide use and industrial availability of W, the database of physical and mechanical properties is very limited. Especially after fusion relevant neutron irradiation and PWI phenomena, most of the properties are still unknown. The plasma fuel consists out of deuterium (D) and tritium (T). Tritium is radio-active and therefore an issue from the safety point of view. During steady-state plasma operation of future fusion power plants, the PFM need to extract a power density of ∼10-20 MW/m 2 . On top of this heat, transient events will deposit an additional non-negligible amount of energy (Disruptions, Vertical Displacement Events, Edge Localized Modes) during short durations. These severe heat loads cause cracking and even melting of the surface resulting in a reduced lifetime and the creation of dust. A contribution to the understanding of cracking phenomena under the severe thermal loads is described as well as the properties degradation under neutron irradiation. Several W grades were irradiated in the BR2 reactor (SCK.CEN) and the thermal loads were simulated with the electron-beam facility JUDITH (FZJ). Since knowledge should be gained about the Tritium retention in the PFM for safety and licensing reasons, a unique test facility at SCK.CEN is being set-up. The plasmatron VISION-I will simulate steady state plasmas for Tritium retention studies. The formation of surface cracks and dust, the initial porosity, neutron induced traps, re-deposited material - change the Tritium

Analysis of cobalt, tantalum, titanium, vanadium and chromium in tungsten carbide by inductively coupled plasma-optical emission spectrometry

CSIR Research Space (South Africa)

Archer, M

2003-12-01

Full Text Available Inductively coupled plasma optical emission spectroscopy (ICP-OES) was used to measure the concentrations of cobalt, tantalum, titanium, vanadium and chromium in solutions of tungsten carbide. The main advantage of the method described here lies...

Traditional Technology of Chromium-Tungsten Steels Facing, its Disadvantages and Suggestions for their Eliminations

OpenAIRE

Valuev, Denis Viktorovich; Malushin, N. N.; Valueva, Anna Vladimirovna; Dariev, R. S.; Mamadaliev, R. A.

2016-01-01

To reveal the disadvantages of the traditional technology of facing with chromium-tungsten steels analysis of the given technology was completed. The analysis showed that the main disadvantages of the technology are high-temperature heating and underutilization of high-alloyed metal properties. To eliminate the disadvantages we developed the methods of facing allowing obtaining faced metal which state is close to that of the hardened one without cracks.

Demonstration of multi-generational growth of tungsten nanoparticles in hydrogen plasma using in situ laser extinction method

Science.gov (United States)

Ouaras, K.; Lombardi, G.; Hassouni, K.

2018-03-01

For the first time, we demonstrate that tungsten (W) nanoparticles (NPs) are created when a tungsten target is exposed to low-pressure, high density hydrogen plasma. The plasma was generated using a novel dual plasma system combining a microwave discharge and a pulsed direct-current (DC) discharge. The tungsten surface originates in the multi-generational formation of a significant population of 30-70 nm diameter particles when the W cathode is biased at ~  -1 kV and submitted to ~1020 m2 s-1 H+/H2+ /H3+ ions flux. The evidenced NPs formation should be taking into account as one of the consequence of the plasma surface interaction outcomes, especially for fusion applications.

Post-examination of helium-cooled tungsten components exposed to DEMO specific cyclic thermal loads

International Nuclear Information System (INIS)

Ritz, G.; Hirai, T.; Linke, J.; Norajitra, P.; Giniyatulin, R.; Singheiser, L.

2009-01-01

A concept of helium-cooled tungsten finger module was developed for the European DEMO divertor. The concept was realized and tested under DEMO specific cyclic thermal loads up to 10 MW/m 2 . The modules were examined carefully before and after loading by metallography and microstructural analyses. While before loading mainly discrete and shallow cracks were found on the tungsten surface due to the manufacturing process, dense crack networks were observed at the loaded surfaces due to the thermal stress. In addition, cracks occurred in the structural, heat sink part and propagated along the grains orientation of the deformed tungsten material. Facilitated by cracking, the molten brazing metal between the tungsten plasma facing material and the W-La 2 O 3 heat sink, that could not withstand the operational temperatures, infiltrated the tungsten components and, due to capillary forces, even reached the plasma facing surface through the cracks. The formed cavity in the brazed layer reduced the heat conduction and the modules were further damaged due to overheating during the applied heat loads. Based on this detailed characterization and possible improvements of the design and of the manufacturing routes are discussed.

Sub-surface microstructure of single and polycrystalline tungsten after high flux plasma exposure studied by TEM

Energy Technology Data Exchange (ETDEWEB)

Dubinko, A., E-mail: adubinko@sckcen.be [Institute for Nuclear Material Sciences, SCK-CEN, 2400 Mol (Belgium); Department of Applied Physics, Ghent University, 9000 Ghent (Belgium); Terentyev, D. [Institute for Nuclear Material Sciences, SCK-CEN, 2400 Mol (Belgium); Bakaeva, A. [Institute for Nuclear Material Sciences, SCK-CEN, 2400 Mol (Belgium); Department of Applied Physics, Ghent University, 9000 Ghent (Belgium); Hernández-Mayoral, M. [Division of Materials, CIEMAT, 28040 Madrid (Spain); De Temmerman, G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul-lez-Durance Cedex (France); Buzi, L. [Forschungszentrum Julich, Inst. Energie & Klimaforsch Plasmaphys, D-52425 Julich (Germany); Noterdaeme, J.-M. [Department of Applied Physics, Ghent University, 9000 Ghent (Belgium); Unterberg, B. [Forschungszentrum Julich, Inst. Energie & Klimaforsch Plasmaphys, D-52425 Julich (Germany)

2017-01-30

Highlights: • Plasma exposure induces dislocation-dominated microstructure as indicated by TEM. • Plasma exposure increases surface dislocation density by an order of magnitude in the polycrystalline tungsten. • Intensive dislocation-grain boundary interaction observed in polycrystalline tungsten. • Dislocation loops are observed in both polycrystalline and single crystal tungsten. - Abstract: We have performed high flux plasma exposure of tungsten and subsequent microstructural characterization using transmission electron microscopy (TEM) techniques. The aim was to reveal the nanometric features in the sub-surface region as well as to compare the microstructural evolution in tungsten single crystal and ITER-relevant specification. In both types of samples, TEM examination revealed the formation of a dense dislocation network and dislocation tangles. The estimated dislocation density in the sub-surface region was of the order of 10{sup 14} m{sup −2} and it gradually decreased with a depth position of the examined sample. Besides individual dislocation lines, networks and tangles, the interstitial dislocation loops have been observed in all examined samples only after the exposure. Contrary to that, examination of the pristine single crystal W and backside of the plasma-exposed samples did not reveal the presence of dislocation loops and tangles. This clearly proves that high flux plasma exposure induces severe plastic deformation in the sub-surface region irrespective of the presence of initial dislocations and sub-grains, and the formation of dislocation tangles, networks and interstitial loops is a co-product of thermal stress and intensive plasma particles uptake.

Synthesis of tungsten oxide, silver, and gold nanoparticles by radio frequency plasma in water

International Nuclear Information System (INIS)

Hattori, Yoshiaki; Nomura, Shinfuku; Mukasa, Shinobu; Toyota, Hiromichi; Inoue, Toru; Usui, Tomoya

2013-01-01

Highlights: •RF plasma in water was used for nanoparticle synthesis. •Nanoparticles were produced from erosion of metallic electrode. •Rectangular and spherical tungsten oxide nanoparticles were produced. •No oxidations of the silver and gold spherical nanoparticles were produced. -- Abstract: A process for synthesis of nanoparticles using plasma in water generated by a radio frequency of 27.12 MHz is proposed. Tungsten oxide, silver, and gold nanoparticles were produced at 20 kPa through erosion of a metallic electrode exposed to plasma. Characterization of the produced nanoparticles was carried out by XRD, absorption spectrum, and TEM. The nanoparticle sizes were compared with those produced by a similar technique using plasma in liquid

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

Directory of Open Access Journals (Sweden)

Kuroda Yusuke

2013-11-01

Full Text Available 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.

Investigation of transient melting of tungsten by ELMs in ASDEX Upgrade.

Czech Academy of Sciences Publication Activity Database

Krieger, K.; Sieglin, B.; Balden, M.; Coenen, J.W.; Göths, B.; Laggner, F.; de Marné, P.; Matthews, G.F.; Nille, D.; Rohde, V.; Dejarnac, Renaud; Faitsch, M.; Giannone, L.; Herrmann, A.; Horáček, Jan; Komm, Michael; Pitts, R.A.; Ratynskaia, S.; Thorén, E.; Tolias, P.

T170, December (2017), č. článku 014030. ISSN 0031-8949. [PFMC 2017: 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications. Düsseldorf, 16.05.2017-19.05.2017] EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : plasma-facing components * tungsten * melting * edge-localized modes Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 1.3 Physical sciences Impact factor: 1.280, year: 2016 http://iopscience.iop.org/article/10.1088/1402-4896/aa8be8

Recent Advances on Hydrogenic Retention in ITER's Plasma-Facing Materials: BE, C, W

International Nuclear Information System (INIS)

Skinner, C.H.; Haasz, A.A.; Alimov, V.Kh.; Bekris, N.; Causey, R.A.; Clark, R.E.H.; Coad, J.P.; Davis, J.W.; Doerner, R.P.; Mayer, M.; Pisarev, A.; Roth, J.; Tanabe, T.

2008-01-01

Management of tritium inventory remains one of the grand challenges in the development of fusion energy and the choice of plasma-facing materials is a key factor for in-vessel tritium retention. The Atomic and Molecular Data Unit of the International Atomic Energy Agency organized a Coordinated Research Project (CRP) on the overall topic of tritium inventory in fusion reactors during the period 2001-2006. This dealt with hydrogenic retention in ITER's plasma-facing materials, Be, C, W, and in compounds (mixed materials) of these elements as well as tritium removal techniques. The results of the CRP are summarized in this article together with recommendations for ITER. Basic parameters of diffusivity, solubility and trapping in Be, C and W are reviewed. For Be, the development of open porosity can account for transient hydrogenic pumping but long term retention will be dominated by codeposition. Codeposition is also the dominant retention mechanism for carbon and remains a serious concern for both Be and C containing layers. Hydrogenic trapping in unirradiated tungsten is low but will increase with ion and neutron damage. Mixed materials will be formed in a tokamak and these can also retain significant amounts of hydrogen isotopes. Oxidative and photon-based techniques for detritiation of plasma-facing components are described

Recent Advances on Hydrogenic Retention in ITER's Plasma-Facing Materials: BE, C, W.

Energy Technology Data Exchange (ETDEWEB)

Skinner, C H; Alimov, Kh; Bekris, N; Causey, R A; Clark, R.E.H.; Coad, J P; Davis, J W; Doerner, R P; Mayer, M; Pisarev, A; Roth, J

2008-03-29

Management of tritium inventory remains one of the grand challenges in the development of fusion energy and the choice of plasma-facing materials is a key factor for in-vessel tritium retention. The Atomic and Molecular Data Unit of the International Atomic Energy Agency organized a Coordinated Research Project (CRP) on the overall topic of tritium inventory in fusion reactors during the period 2001-2006. This dealt with hydrogenic retention in ITER's plasma-facing materials, Be, C, W, and in compounds (mixed materials) of these elements as well as tritium removal techniques. The results of the CRP are summarized in this article together with recommendations for ITER. Basic parameters of diffusivity, solubility and trapping in Be, C and W are reviewed. For Be, the development of open porosity can account for transient hydrogenic pumping but long term retention will be dominated by codeposition. Codeposition is also the dominant retention mechanism for carbon and remains a serious concern for both Be and C containing layers. Hydrogenic trapping in unirradiated tungsten is low but will increase with ion and neutron damage. Mixed materials will be formed in a tokamak and these can also retain significant amounts of hydrogen isotopes. Oxidative and photon-based techniques for detritiation of plasma-facing components are described.

Performance of tungsten-based materials and components under ITER and DEMO relevant steady-state thermal loads

Energy Technology Data Exchange (ETDEWEB)

Ritz, Guillaume Henri

2011-07-01

In nuclear fusion devices the surfaces directly facing the plasma are irradiated with high energy fluxes. The most intense loads are deposited on the divertor located at the bottom of the plasma chamber, which has to withstand continuous heat loads with a power density of several MW . m{sup -2} as well as transient events. These are much shorter (in the millisecond and sub-millisecond regime) but deposit a higher power densities of a few GW . m{sup -2}. The search for materials that can survive to those severe loading conditions led to the choice of tungsten which possesses advantageous attributes such as a high melting point, high thermal conductivity, low thermal expansion and an acceptable activation rate. These properties made it an attractive and promising candidate as armor material for divertors of future fusion devices such as ITER and DEMO. For the DEMO divertor, conceptual studies on helium-cooled tungsten plasma-facing components were performed. The concept was realized and tested under DEMO specific cyclic thermal loads. The examination of the plasma-facing components by microstructural analyses before and after thermal loading enabled to determine the mechanisms for components failure. Among others, it clearly showed the impact of the tungsten grade and the thermal stress induced crack formation on the performance of the armor material and in general of the plasma-facing component under high heat loads. A tungsten qualification program was launched to study the behaviour of various tungsten grades, in particular the crack formation, under fusion relevant steady-state thermal loads. In total, seven commercially available materials from two industrial suppliers were investigated. As the material's thermal response is strongly related to its microstructure, this program comprised different material geometries and manufacturing technologies. It also included the utilization of an actively cooled specimen holder which has been designed to perform

Performance of tungsten-based materials and components under ITER and DEMO relevant steady-state thermal loads

International Nuclear Information System (INIS)

Ritz, Guillaume Henri

2011-01-01

In nuclear fusion devices the surfaces directly facing the plasma are irradiated with high energy fluxes. The most intense loads are deposited on the divertor located at the bottom of the plasma chamber, which has to withstand continuous heat loads with a power density of several MW . m -2 as well as transient events. These are much shorter (in the millisecond and sub-millisecond regime) but deposit a higher power densities of a few GW . m -2 . The search for materials that can survive to those severe loading conditions led to the choice of tungsten which possesses advantageous attributes such as a high melting point, high thermal conductivity, low thermal expansion and an acceptable activation rate. These properties made it an attractive and promising candidate as armor material for divertors of future fusion devices such as ITER and DEMO. For the DEMO divertor, conceptual studies on helium-cooled tungsten plasma-facing components were performed. The concept was realized and tested under DEMO specific cyclic thermal loads. The examination of the plasma-facing components by microstructural analyses before and after thermal loading enabled to determine the mechanisms for components failure. Among others, it clearly showed the impact of the tungsten grade and the thermal stress induced crack formation on the performance of the armor material and in general of the plasma-facing component under high heat loads. A tungsten qualification program was launched to study the behaviour of various tungsten grades, in particular the crack formation, under fusion relevant steady-state thermal loads. In total, seven commercially available materials from two industrial suppliers were investigated. As the material's thermal response is strongly related to its microstructure, this program comprised different material geometries and manufacturing technologies. It also included the utilization of an actively cooled specimen holder which has been designed to perform sophisticated

Numerical simulation of strong evaporation and condensation for plasma-facing materials

International Nuclear Information System (INIS)

Kunugi, T.; Yasuda, H.

1994-01-01

The thermal response of the divertor plate to the hard plasma disruptions had been analyzed numerically by the two dimensional transient heat transfer code. There are several studies of the vapor shielding effects on the thermal response to the plasma disruption. However, it was pointed out some discrepancies among the numerical results calculated by U.S., EC and Japan for the same disruption conditions by van der Laan. One of the authors studied the sensitivity of some parameters (i.e., the temperature dependency of the thermal properties, an evaporation coefficient and a saturated condensation ratio) of disruption erosion analysis. Though the authors expected that the variations in evaporation models lead to the large variety of the erosion, they gave no significant effects on the surface temperature, the evaporation and melt-layer thickness. In this paper, the authors will describe the development of the numerical simulation codes for the strong evaporation and condensation from the plasma facing materials (PFMs) such as carbon, tungsten and beryllium

Spectroscopic Investigations of Highly Charged Tungsten Ions - Atomic Spectroscopy and Fusion Plasma Diagnostics

Energy Technology Data Exchange (ETDEWEB)

Clementson, Joel [Lund Univ. (Sweden)

2010-05-01

The spectra of highly charged tungsten ions have been investigated using x-ray and extreme ultraviolet spectroscopy. These heavy ions are of interest in relativistic atomic structure theory, where high-precision wavelength measurements benchmark theoretical approaches, and in magnetic fusion research, where the ions may serve to diagnose high-temperature plasmas. The work details spectroscopic investigations of highly charged tungsten ions measured at the Livermore electron beam ion trap (EBIT) facility. Here, the EBIT-I and SuperEBIT electron beam ion traps have been employed to create, trap, and excite tungsten ions of M- and L-shell charge states. The emitted spectra have been studied in high resolution using crystal, grating, and x-ray calorimeter spectrometers. In particular, wavelengths of n = 0 M-shell transitions in K-like W⁵⁵⁺ through Ne-like W⁶⁴⁺, and intershell transitions in Zn-like W⁴⁴⁺ through Co-like W⁴⁷⁺ have been measured. Special attention is given to the Ni-like W46+ ion, which has two strong electric-dipole forbidden transitions that are of interest for plasma diagnostics. The EBIT measurements are complemented by spectral modeling using the Flexible Atomic Code (FAC), and predictions for tokamak spectra are presented. The L-shell tungsten ions have been studied at electron-beam energies of up to 122 keV and transition energies measured in Ne-like W⁶⁴⁺ through Li-like W⁷¹⁺. These spectra constitute the physics basis in the design of the ion-temperature crystal spectrometer for the ITER tokamak. Tungsten particles have furthermore been introduced into the Sustained Spheromak Physics Experiment (SSPX) spheromak in Livermore in order to investigate diagnostic possibilities of extreme ultraviolet tungsten spectra for the ITER divertor. The spheromak measurement and spectral modeling using FAC suggest that tungsten ions in charge states around Er-like W⁶⁺ could be useful for

Damage to Preheated Tungsten Targets after Multiple Plasma Impacts Simulating ITER ELMs

Energy Technology Data Exchange (ETDEWEB)

Garkusha, I.E.; Bandura, A.N.; Byrka, O.V.; Chebotarev, V.V.; Makhlay, V.A.; Tereshin, V.I. [Kharkov Inst. of Physics and Technology, Inst. of Plasma Physics of National Science Center, Akademicheskaya street, 1, 61108 Kharkov (Ukraine); Landman, I.; Pestchanyi, S. [FZK-Forschungszentrum Karlsruhe, Association Euratom-FZK, Technik und Umwelt, Postfach 3640, D-7602 1 Karlsruhe (Germany)

2007-07-01

Full text of publication follows: The energy loads onto ITER divertor surfaces associated with the Type I ELMs are expected to be up to 1 MJ/m{sup 2} during 0.1-0.5 ms, with the number of pulses about 103 per discharge. Tungsten is a candidate material for major part of the surface, but its brittleness can result in substantial macroscopic erosion after the repetitive heat loads. To minimize the brittle destruction, tungsten may be preheated above the ductile-to-brittle transition temperature. In this work the behavior of preheated tungsten targets under repetitive ELM-like plasma pulses is studied in simulation experiments with the quasi-stationary plasma accelerator QSPA Kh-50. The targets have been exposed up to 450 pulses of the duration 0.25 ms and the heat loads either 0.45 MJ/m{sup 2} or 0.75 MJ/m{sup 2}, which is respectively below and above the melting threshold. During the exposures the targets were permanently kept preheated at 650 deg. C by a heater at target backside. In the course of exposures the irradiated surfaces were examined after regular numbers of pulses using the SEM and the optical microscopy. The profilometry, XRD, microhardness and weight loss measurements have been performed, as well as comparisons of surface damages after the heat loads both below and above the melting threshold. It is obtained that macro-cracks do not develop on the preheated surface. After the impacts with surface melting, a fine mesh of intergranular microcracks has appeared. The width of fine intergranular cracks grows with pulse number, achieving 1-1.5 microns after 100 pulses, and after 210 pulses the crack width increases up to 20 microns, which is comparable with grain sizes. Threshold changes in surface morphology resulting in corrugation structures and pits on the surface as well as importance of surface tension in resulted 'micro-brush' structures are discussed. Further evolution of the surface pattern is caused by loss of separated grains on exposed

Deuterium-induced nanostructure formation on tungsten exposed to high-flux plasma

NARCIS (Netherlands)

Xu, H.Y.; De Temmerman, G.; Luo, G. N.; Jia, Y. Z.; Yuan, Y.; Fu, B. Q.; Godfrey, A.; Liu, W.

2015-01-01

Surface topography of polycrystalline tungsten (W) have been examined after exposure to a low-energy (38 eV/D), high-flux (∼1.1–1.5 × 1024 m−2 s−1) deuterium plasma in the Pilot-PSI linear plasma device. The methods used were scanning electron microscopy

Simulations of thermionic suppression during tungsten transient melting experiments.

Czech Academy of Sciences Publication Activity Database

Komm, Michael; Tolias, P.; Ratynskaia, S.; Dejarnac, Renaud; Gunn, J. P.; Krieger, K.; Podolník, Aleš; Pitts, R.A.; Pánek, Radomír

T170, December (2017), č. článku 014069. ISSN 0031-8949. [PFMC 2017: 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications. Düsseldorf, 16.05.2017-19.05.2017] R&D Projects: GA ČR(CZ) GA16-14228S; GA MŠk(CZ) 8D15001 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : tokamak * thermionic emission * tungsten * melt * plasma-facing component Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 1.3 Physical sciences Impact factor: 1.280, year: 2016 http://iopscience.iop.org/article/10.1088/1402-4896/aa9209

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

International Nuclear Information System (INIS)

Shimada, Masashi; Hatano, Y.; Calderoni, P.; Oda, T.; Oya, Y.; Sokolov, M.; Zhang, K.; Cao, G.; Kolasinski, R.; Sharpe, J.P.

2011-01-01

With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 x 10 21 m -2 s -1 , ion fluence: 4 x 10 25 m -2 ) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

Science.gov (United States)

Shimada, Masashi; Hatano, Y.; Calderoni, P.; Oda, T.; Oya, Y.; Sokolov, M.; Zhang, K.; Cao, G.; Kolasinski, R.; Sharpe, J. P.

2011-08-01

With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 × 1021 m-2 s-1, ion fluence: 4 × 1025 m-2) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

Internal reflection of interstitial atoms from close-packed tungsten faces

International Nuclear Information System (INIS)

Dranova, Zh.I.; Mikhajlovskij, I.M.

1981-01-01

Use of field-ion microscopy methods has shown that changes in microtopography of tungsten specimens irradiated with 2-5 keV helium atoms are mainly related to the liberation of interstitial atoms on the surface. It is established that the atom liberation on the surface is considerably anisotropic: maximum quantity of atoms is observed in the vicinity of faces (100), (111) and (211) along the sections of zone lines (110) oriented along the edge of the first Brillouin zone. The atom liberation on plane sections of the most dense-packed face (110) was not observed as a rule; atomic steps of the face are interstitial atom sinks. It is concluded on the basis of the results obtained that there is the predominant inner reflection of interstitial atoms from the dense-packed faces and a possible contribution of inner reflection to the surface migration processes activated with the ion bombardment as well as material swelling have been analyzed [ru

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

Science.gov (United States)

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.

Modeling of hydrogen desorption from tungsten surface

Energy Technology Data Exchange (ETDEWEB)

Guterl, J., E-mail: jguterl@ucsd.edu [University of California, San Diego, La Jolla, CA 92093 (United States); Smirnov, R.D. [University of California, San Diego, La Jolla, CA 92093 (United States); Krasheninnikov, S.I. [University of California, San Diego, La Jolla, CA 92093 (United States); Nuclear Research National University MEPhI, Moscow 115409 (Russian Federation); Uberuaga, B.; Voter, A.F.; Perez, D. [Los Alamos National Laboratory, Los Alamos, NM 8754 (United States)

2015-08-15

Hydrogen retention in metallic plasma-facing components is among key-issues for future fusion devices. For tungsten, which has been chosen as divertor material in ITER, hydrogen desorption parameters experimentally measured for fusion-related conditions show large discrepancies. In this paper, we therefore investigate hydrogen recombination and desorption on tungsten surfaces using molecular dynamics simulations and accelerated molecular dynamics simulations to analyze adsorption states, diffusion, hydrogen recombination into molecules, and clustering of hydrogen on tungsten surfaces. The quality of tungsten hydrogen interatomic potential is discussed in the light of MD simulations results, showing that three body interactions in current interatomic potential do not allow to reproduce hydrogen molecular recombination and desorption. Effects of surface hydrogen clustering on hydrogen desorption are analyzed by introducing a kinetic model describing the competition between surface diffusion, clustering and recombination. Different desorption regimes are identified and reproduce some aspects of desorption regimes experimentally observed.

Status of R and D of the plasma facing components for the ITER divertor

International Nuclear Information System (INIS)

Mazul, I.V.; Akiba, M.; Arkhipov, I.

2001-01-01

The paper reports the progress made by the ITER Home Teams in the development of robust carbon and tungsten armoured plasma facing components for the ITER divertor. The activities on the development and study of armour materials, joining technologies, non-destructive evaluation techniques, high heat flux testing of manufactured components and neutron irradiation resistance studies are presented. The results of these activities confirm the feasibility of the main divertor components. Examples of the fruitful collaboration between Parties and future R and D needs are also described. (author)

Conceptual Design for a Bulk Tungsten Divertor Tile in JET

International Nuclear Information System (INIS)

Mertens, P.; Neubauer, O.; Philipps, V.; Schweer, B.; Samm, U.; Hirai, T.; Sadakov, S.

2006-01-01

With ITER on the verge of being build, the ITER-like Wall project (ILW) for JET aims at providing the plasma chamber of the tokamak with an environment of mixed materials which will be relevant to the support of decisions to the first wall construction and, from the point of view of plasma physics, to the corresponding investigations of possible plasma configuration and plasma-wall interaction. In both respects, tungsten plays a key role in the divertor cladding whereas beryllium will be used for the vessel's first wall. For the central tile, also called LB-SRP for '' Load-Bearing Septum Replacement Plate '', resort to bulk tungsten is envisaged in order to cope with the high loads expected (up to 10 MW/m 2 for about 10 s). This is indeed the preferred plasma-facing component for positioning the outer strike-point in the divertor. Forschungszentrum Juelich has developed a conceptual design for this tile, based on an assembly of tungsten blades or lamellae. It was selected in the frame of an extensive R-and-D study in search of a suitable, inertially cooled component(T. Hirai et al., R-and-D on full tungsten divertor and beryllium wall for JET ITER-like Wall Project: this conference). As reported elsewhere, the design is actually driven by electromagnetic considerations in the first place(S. Sadakov et al., Detailed electromagnetic analysis for optimisation of a tungsten divertor plate for JET: this conference). The lamellae are grouped in four stacks per tile which are independently attached to an equally re-designed supporting structure. A so-called adapter plate, also a new design, takes care of an appropriate interface to the base carrier of JET, onto which modules of two tiles are positioned and screwed by remote handling (RH) procedures. The compatibility of the design on the whole with RH requirements is another essential ingredient which was duly taken into account throughout. The concept and the underlying philosophy will be presented along with important

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

Energy Technology Data Exchange (ETDEWEB)

Shimada, Masashi, E-mail: Masashi.Shimada@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Hatano, Y. [Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555 (Japan); Calderoni, P. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Oda, T. [Department of Nuclear Engineering and Management, The University of Tokyo, Tokyo 113-8656 (Japan); Oya, Y. [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Sokolov, M. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Zhang, K. [Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555 (Japan); Cao, G. [Department of Engineering Physics, University of Wisconsin-Madison, Madison, WI 53706 (United States); Kolasinski, R. [Hydrogen and Metallurgical Science Department, Sandia National Laboratories, Livermore, CA 94551 (United States); Sharpe, J.P. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

2011-08-01

With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 x 10{sup 21} m{sup -2} s{sup -1}, ion fluence: 4 x 10{sup 25} m{sup -2}) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

Tungsten dust remobilization under steady-state and transient plasma conditions.

Czech Academy of Sciences Publication Activity Database

Ratynskaia, S.; Tolias, P.; De Angeli, M.; Weinzettl, Vladimír; Matějíček, Jiří; Bykov, I.; Rudakov, D.L.; Vignitchouk, L.; Thorén, E.; Riva, G.; Ripamonti, D.; Morgan, T.; Pánek, Radomír; De Temmerman, G.

2017-01-01

Roč. 12, August (2017), s. 569-574 ISSN 2352-1791. [PSI 2016 - 22nd International Conference on Plasma Surface Interactions in Controlled Fusion Devices/22./. Roma, 30.05.2016-03.06.2016] R&D Projects: GA ČR(CZ) GA14-12837S; GA MŠk(CZ) LM2015045 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : Tungsten * Dust * Remobilization * Tokamak Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.sciencedirect.com/science/article/pii/S2352179116301168

Tritium Decay Helium-3 Effects in Tungsten

Energy Technology Data Exchange (ETDEWEB)

Shimada, M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Merrill, B. J. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-06-01

A critical challenge for long-term operation of ITER and beyond to a Demonstration reactor (DEMO) and future fusion reactor will be the development of plasma-facing components (PFCs) that demonstrate erosion resistance to steady-state/transient heat fluxes and intense neutral/ion particle fluxes under the extreme fusion nuclear environment, while at the same time minimizing in-vessel tritium inventories and permeation fluxes into the PFC’s coolant. Tritium will diffuse in bulk tungsten at elevated temperatures, and can be trapped in radiation-induced trap site (up to 1 at. % T/W) in tungsten [1,2]. Tritium decay into helium-3 may also play a major role in microstructural evolution (e.g. helium embrittlement) in tungsten due to relatively low helium-4 production (e.g. He/dpa ratio of 0.4-0.7 appm [3]) in tungsten. Tritium-decay helium-3 effect on tungsten is hardly understood, and its database is very limited. Two tungsten samples (99.99 at. % purity from A.L.M.T. Co., Japan) were exposed to high flux (ion flux of 1.0x1022 m-2s-1 and ion fluence of 1.0x1026 m-2) 0.5%T2/D2 plasma at two different temperatures (200, and 500°C) in Tritium Plasma Experiment (TPE) at Idaho National Laboratory. Tritium implanted samples were stored at ambient temperature in air for more than 3 years to investigate tritium decay helium-3 effect in tungsten. The tritium distributions on plasma-exposed was monitored by a tritium imaging plate technique during storage period [4]. Thermal desorption spectroscopy was performed with a ramp rate of 10°C/min up to 900°C to outgas residual deuterium and tritium but keep helium-3 in tungsten. These helium-3 implanted samples were exposed to deuterium plasma in TPE to investigate helium-3 effect on deuterium behavior in tungsten. The results show that tritium surface concentration in 200°C sample decreased to 30 %, but tritium surface concentration in 500°C sample did not alter over the 3 years storage period, indicating possible tritium

Binary-collision-approximation simulation for noble gas irradiation onto plasma facing materials

International Nuclear Information System (INIS)

Saito, Seiki; Nakamura, Hiroaki; Takayama, Arimichi; Ito, Atsushi M

2014-01-01

A number of experiments show that helium plasma constructs filament (fuzz) structures whose diameter is in nanometer-scale on the tungsten material under the suitable experimental condition. In this paper, binary-collision-approximation-based simulation is performed to reveal the mechanism and the conditions of fuzz formation of tungsten material under plasma irradiation. The irradiation of the plasma of hydrogen, deuterium, and tritium, and also the plasma of noble gas such as helium, neon, and argon atoms are investigated. The possibility of fuzz formation is discussed on the simulation result of penetration depth of the incident atoms

Formation of carbon containing layers on tungsten test limiters

International Nuclear Information System (INIS)

Rubel, M.; Philipps, V.; Huber, A.; Tanabe, T.

1999-01-01

Tungsten test limiters of mushroom shape and a plasma facing area of approximately 100 cm 2 were exposed at the TEXTOR-94 tokamak to a number of deuterium fuelled discharges performed under various operation conditions. Two types of limiters were tested: a sole tungsten limiter and a twin limiter consisting of two halves, one made of tungsten and another of graphite. The exposed surfaces were examined with ion beam analysis methods and laser profilometry. The formation of some deposition zones was observed near the edges of the limiters. The deuterium-to-carbon concentration ratio was in the range from 0.04 to 0.11 and around 0.2 for the sole tungsten and the twin limiter, respectively. Significant amounts of the co-deposited tungsten and silicon atoms were found on the graphite part of the twin limiter indicating the formation of mixed W-C-Si compounds. (orig.)

Plasma-wall interactions data compendium-1. ''Hydrogen retention property, diffusion and recombination coefficients database for selected plasma-facing materials''

Energy Technology Data Exchange (ETDEWEB)

Iwakiri, Hirotomo [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics; Matsuhiro, Kenjirou [Osaka Univ., Osaka (Japan); Hirooka, Yoshi [National Inst. for Fusion Science, Toki, Gifu (Japan); Yamamura, Yasunori [Okayama Univ. of Scinece, Okayama (Japan)

2002-05-01

A summary on the recent activities of the plasma-wall interactions database task group at the National Institute for Fusion Science is presented in this report. These activities are focused on the compilation of literature data on the key parameters related to wall recycling characteristics that affect dynamic particle balance during plasma discharges and also on-site tritium inventory. More specifically, in this task group a universal fitting formula has been proposed and successfully applied to help compile hydrogen implantation-induced retention data. Also, presented here are the data on hydrogen diffusion and surface recombination coefficients, both critical in modeling dynamic wall recycling behavior. Data compilation has been conducted on beryllium, carbon, tungsten and molybdenum, all currently used for plasma-facing components in magnetic fusion experiments. (author)

Peculiarity of deuterium ions interaction with tungsten surface in the condition imitating combination of normal operation with plasma disruption in ITER

Energy Technology Data Exchange (ETDEWEB)

Guseva, M.I. E-mail: martyn@nfi.kiae.ru; Vasiliev, V.I.; Gureev, V.M.; Danelyan, L.S.; Khirpunov, B.I.; Korshunov, S.N.; Kulikauskas, V.S.; Martynenko, Yu.V.; Petrov, V.B.; Strunnikov, V.N.; Stolyarova, V.G.; Zatekin, V.V.; Litnovsky, A.M

2001-03-01

Tungsten is a candidate material for the ITER divertor. For the simulation of ITER normal operation conditions in combination with plasma disruptions samples of various types of tungsten were exposed to both steady-state and high power pulsed deuterium plasmas. Tungsten samples were first exposed in a steady-state plasma with an ion current density {approx}10{sup 21} m{sup -2} s{sup -1} up to a dose of 10{sup 25} m{sup -2} at a temperature of 770 K. The energy of deuterium ions was 150 eV. The additional exposure of the samples to 10 pulses of deuterium plasma was performed in the electrodynamical plasma accelerator with an energy flux 0.45 MJ/m{sup 2} per pulse. Samples of four types of tungsten (W-1%La{sub 2}O{sub 3}, W-13I, monocrystalline W(1 1 1) and W-10%Re) were investigated. The least destruction of the surface was observed for W(1 1 1). The concentration of retained deuterium in tungsten decreased from 2.5x10{sup 19} m{sup -2} to 1.07x10{sup 19} m{sup -2} (for W(1 1 1)) as a result of the additional pulsed plasma irradiation. Investigation of the tungsten erosion products after the high power pulsed plasma shots was also carried out.

Hot tungsten plate based ionizer for cesium plasma in a multi-cusp field experiment

International Nuclear Information System (INIS)

Patel, Amitkumar D.; Sharma, Meenakshee; Ramasubramanian, Narayanan; Chattopadhyay, Prabal K.

2015-01-01

In a newly proposed basic experiment, contact-ionized cesium ions will be confined by a multi cups magnetic field configuration. The cesium ion will be produced by impinging collimated neutral atoms on an ionizer consisting of the hot tungsten plate. The temperature of the tungsten plate will also be made high enough (∼2700 K) such that it will contribute electrons also to the plasma. It is expected that at this configuration the cesium plasma would be really quiescent and would be free from even the normal drift waves observed in the classical Q-machines. For the ionizer a design based on F. F. Chen's design was made. This ionizer is very fine machining and exotic material like Tungsten plate, Molybdenum screws, rings, and Boron Nitride ceramics etc. The fine and careful machining of these materials was very hard. In this paper, the experience about to join the tungsten wire to molybdenum plate and alloy of tantalum and molybdenum ring is described. In addition experimental investigations have been made to measure 2D temperature distribution profile of the Tungsten hot plate using infrared camera and the uniformity of temperature distribution over the hot plate surface is discussed. (author)

Initiation of arcing on tungsten surface exposed to steady state He plasmas

Science.gov (United States)

Kajita, Shin; Noiri, Yasuyuki; Ohno, Noriyasu

2015-09-01

Arcing was initiated in steady state helium plasmas by negatively biasing a tungsten electrode to around -500 V. On the tungsten electrode, nanostructures were grown by the plasma irradiation. In this study, we characterized the property of the initiated arcing by measuring the temporal evolutions of the electrode potential and the arc current. The ignition frequency and duration of arcing were presented from the potential measurements; the arc duration was in the range of changing the biasing voltage. The behavior of arc spots was observed with a fast framing camera. It was shown that the spots split frequently, and sometimes, they run on the surface independently. From the fluctuation of the arc current, the fractal feature of arcing was revealed.

The manufacture of carbon armoured plasma-facing components for fusion devices

International Nuclear Information System (INIS)

Schedler, B.; Huber, T.; Zabernig, A.; Rainer, F.; Scheiber, K.H.; Schedle, D.

2001-01-01

Within the last decade Plansee has been active in the development and manufacture of different plasma-facing-components for nuclear fusion experiments consisting in a tungsten or CFC-armor joined onto metallic substrates like TZM, stainless steel or copper-alloys. The manufacture of these components requires unique joining technologies in order to obtain reliable thermo mechanical stable joints able to withstand highest heat fluxes without any deterioration of the joint. In an overview the different techniques will be presented by some examples of components already manufactured and successfully tested under high heat flux conditions. Furthermore an overview will be given on the manufacture of different high heat flux components for TORE SUPRA, Wendelstein 7-X and ITER. (author)

OEDGE modeling for the planned tungsten ring experiment on DIII-D

Directory of Open Access Journals (Sweden)

J.D. Elder

2017-08-01

Full Text Available The OEDGE code is used to model tungsten erosion and transport for experiments with toroidal rings of high-Z metal tiles in the DIII-D tokamak. Such modeling is needed for both experimental and diagnostic design to have estimates of the expected core and edge tungsten density and to understand the various factors contributing to the uncertainties in these calculations. OEDGE simulations are performed using the planned experimental magnetic geometries and plasma conditions typical of both L-mode and inter-ELM H-mode discharges in DIII-D. OEDGE plasma reconstruction based on specific representative discharges for similar geometries is used to determine the plasma conditions applied to tungsten plasma impurity simulations. A new model for tungsten erosion in OEDGE was developed which imports charge-state resolved carbon impurity fluxes and impact energies from a separate OEDGE run which models the carbon production, transport and deposition for the same plasma conditions as the tungsten simulations. These values are then used to calculate the gross tungsten physical sputtering due to carbon plasma impurities which is then added to any sputtering by deuterium ions; tungsten self-sputtering is also included. The code results are found to be dependent on the following factors: divertor geometry and closure, the choice of cross-field anomalous transport coefficients, divertor plasma conditions (affecting both tungsten source strength and transport, the choice of tungsten atomic physics data used in the model (in particular ionization rate for W-atoms, and the model of the carbon flux and energy used for calculating the tungsten source due to sputtering. Core tungsten density is found to be of order 1015m−3 (excluding effects of any core transport barrier and with significant variability depending on the other factors mentioned with density decaying into the scrape off layer. For the typical core density in the plasma conditions examined of 2 to 4�

Tungsten erosion and redeposition in the all-tungsten divertor of ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Mayer, M; Krieger, K; Matern, G; Neu, R; Rasinski, M; Rohde, V; Sugiyama, K; Wiltner, A [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, 85748 Garching (Germany); Andrzejczuk, M; Fortuna-Zalesna, E; Kurzydlowski, K J; Zielinski, W [Faculty of Materials Science and Engineering, Warsaw University of Technology, Association EURATOM-IPPLM, 02-507 Warsaw (Poland); Hakola, A; Koivuranta, S; Likonen, J [VTT Materials for Power Engineering, EURATOM Association, PO Box 1000, FI-02044 VTT (Finland); Ramos, G [CICATA-Qro, Instituto Politecnico Nacional, Queretaro (Mexico); Dux, R, E-mail: matej.mayer@ipp.mpg.de

2009-12-15

Net erosion and deposition of tungsten (W) in the ASDEX Upgrade divertor were determined after the 2007 campaign by using thin W marker stripes. ASDEX Upgrade had full-W plasma-facing components during this campaign. The inner divertor and the roof baffle were net W deposition areas with a maximum deposition of about 1x10{sup 18} W-atoms cm{sup -2} in the private flux region below the inner strike point. Net erosion of W was observed in the whole outer divertor, with the largest erosion close to the outer strike point. Only a small fraction of the W eroded in the main chamber and in the outer divertor was found in redeposits in the inner divertor, while a large fraction was either redeposited at unidentified places in the main chamber or has formed dust.

Quantum-Accurate Molecular Dynamics Potential for Tungsten

Energy Technology Data Exchange (ETDEWEB)

Wood, Mitchell; Thompson, Aidan P.

2017-03-01

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-He binary system.

Comprehensive simulation of vertical plasma instability events and their serious damage to ITER plasma facing components

International Nuclear Information System (INIS)

Hassanein, A.; Sizyuk, T.

2008-01-01

Safe and reliable operation is still one of the major challenges in the development of the new generation of ITER-like fusion reactors. The deposited plasma energy during major disruptions, edge-localized modes (ELMs) and vertical displacement events (VDEs) causes significant surface erosion, possible structural failure and frequent plasma contamination. While plasma disruptions and ELM will have no significant thermal effects on the structural materials or coolant channels because of their short deposition time, VDEs having longer-duration time could have a destructive impact on these components. Therefore, modelling the response of structural materials to VDE has to integrate detailed energy deposition processes, surface vaporization, phase change and melting, heat conduction to coolant channels and critical heat flux criteria at the coolant channels. The HEIGHTS 3D upgraded computer package considers all the above processes to specifically study VDE in detail. Results of benchmarking with several known laboratory experiments prove the validity of HEIGHTS implemented models. Beryllium and tungsten are both considered surface coating materials along with copper structure and coolant channels using both smooth tubes with swirl tape insert. The design requirements and implications of plasma facing components are discussed along with recommendations to mitigate and reduce the effects of plasma instabilities on reactor components.

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

Science.gov (United States)

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 e

Technology R&D Activities for the ITER Full-tungsten Divertor

Energy Technology Data Exchange (ETDEWEB)

Lorenzetto, P.; Bednarek, M.; Gavila, P.; Riccardi, B.; Saibene, G., E-mail: patrick.lorenzetto@f4e.europa.eu [Fusion for Energy, Barcelona (Spain); Escourbiac, F.; Hirai, T.; Merola, M.; Pitts, R. [ITER Organization, St Paul-lez-Durance (France); Suzuki, S. [JAEA, Ibaraki (Japan); Mazul, I. [Efremov Institute, St.Petersburg (Russian Federation)

2012-09-15

Full text: The current ITER Baseline foresees the use of carbon fibre composite (CFC) as armour material in the high heat flux strike point regions and tungsten (W) elsewhere in the divertor for the initial non-active phase of operation with hydrogen and helium plasmas. This divertor would then be replaced with a full-W divertor for the nuclear phase with deuterium and deuterium- tritium plasmas. To reduce costs the ITER Organization (IO) has proposed to install a full-W divertor from start of operations and to implement a work programme to develop a full-W divertor design, qualify the corresponding fabrication technology and investigate critical physics and operational issues with support from the R&D fusion community. An extensive R&D programme has been implemented over more than 15 years to develop fabrication technologies for the procurement of ITER divertor components. Significant effort has been devoted to the development of reliable armour/heat sink joining techniques such as Hot Isostatic Pressing (Europe), Hot Radial Pressing (Europe) or brazing (Japan, Russia). In this development programme, established for the CFC/W divertor variant, the design solution for W-armoured components was optimized for the divertor baffle and dome regions, namely for steady state operation conditions at heat flux values of typically 5 MW/m{sup 2} and for slow transient events at heat flux values up to 10 MW/m{sup 2}. A very positive outcome of this R&D work has been that some fabrication technologies mentioned above can achieve much higher performances, close to the expected slow transient conditions for the strike point region (20 MW/m{sup 2} for 10 s). To prepare for the procurement of a full-W divertor, a development work programme has been launched including in particular the manufacturing and high heat flux testing of small-scale mock-ups with improved monoblock geometries and full-W pre-qualification prototypes, and the manufacturing and testing of qualification full

Investigation of arcing on fiber-formed nanostructured tungsten by pulsed plasma during steady state plasma irradiation

Energy Technology Data Exchange (ETDEWEB)

Yajima, M., E-mail: yajima.miyuki@LHD.nifs.ac.jp [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, 509-5292 Japan (Japan); Ohno, N. [Graduate School of Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan); Kajita, S. [EcoTopia Science Institute, Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan); De Temmerman, G. [ITER Organization, Route de Vinon sur Verdon, CS 90 046-13067 St Paul Lez Durance Cedex (France); Bystrov, K.; Bardin, S.; Morgan, T.W. [FOM Institute DIFFER, Dutch Institute for Fundamental Energy Research, Partner in the Trilateral Euregio Cluster, 5612 AJ Eindhoven (Netherlands); Masuzaki, S. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, 509-5292 Japan (Japan)

2016-11-15

Arcing on fiber-formed nanostructured tungsten samples during ELM-like pulses was investigated using the superimposition of high power pulsed plasma on a steady state plasma with hydrogen gas in the linear plasma device Pilot-PSI. The ignition of arcing was observed when the floating potential of the samples was less than −75 V with sufficient heat flux. The surface observation showed that the arc spots were not in the center, but in the peripheral area of the plasma column. Considering the plasma potential profile in the Pilot-PSI, the arcing occurred at the position where the heat flux and the sheath potential drop are sufficiently large.

Tungsten erosion under plasma heat loads typical for ITER type I Elms and disruptions

Energy Technology Data Exchange (ETDEWEB)

Garkusha, I.E. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine)]. E-mail: garkusha@ipp.kharkov.ua; Bandura, A.N. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Byrka, O.V. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Chebotarev, V.V. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Landman, I.S. [Forschungszentrum Karlsruhe, IHM, 76021 Karlsruhe (Germany); Makhlaj, V.A. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Marchenko, A.K. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Solyakov, D.G. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Tereshin, V.I. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Trubchaninov, S.A. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine); Tsarenko, A.V. [Institute of Plasma Physics of the NSC KIPT, 61108 Kharkov (Ukraine)

2005-03-01

The behavior of pure sintered tungsten under repetitive plasma heat loads of {approx}1 MJ/m{sup 2} (which is relevant to ITER ELMs) and 25 MJ/m{sup 2} (ITER disruptions) is studied with the quasi-steady-state plasma accelerator QSPA Kh-50. The ELM relevant heat loads have resulted in formation of two kinds of crack networks, with typical sizes of 10-20 {mu}m and {approx}1 mm, at the surface. Tungsten preheating to 600 deg. C indicates that fine intergranular cracks are probably caused by thermal stresses during fast resolidification of the melt, whereas large cracks are the result of ductile-to-brittle transition. For several hundreds of ELM-like exposures, causing surface melting, the melt motion does not dominate the profile of the melt spot. The disruption relevant experiments demonstrated that melt motion became the main factor of tungsten damage.

Deuterium trapping in tungsten

Science.gov (United States)

Poon, Michael

Tungsten is one of the primary material candidates being investigated for use in the first-wall of a magnetic confinement fusion reactor. An ion accelerator was used to simulate the type of ion interaction that may occur at a plasma-facing material. Thermal desorption spectroscopy (TDS) was the primary tool used to analyze the effects of the irradiation. Secondary ion mass spectroscopy (SIMS) was used to determine the distribution of trapped D in the tungsten specimen. The tritium migration analysis program (TMAP) was used to simulate thermal desorption profiles from the D depth distributions. Fitting of the simulated thermal desorption profiles with the measured TDS results provided values of the D trap energies. Deuterium trapping in single crystal tungsten was studied as a function of the incident ion fluence, ion flux, irradiation temperature, irradiation history, and surface impurity levels during irradiation. The results show that deuterium was trapped at vacancies and voids. Two deuterium atoms could be trapped at a tungsten vacancy, with trapping energies of 1.4 eV and 1.2 eV for the first and second D atoms, respectively. In a tungsten void, D is trapped as atoms adsorbed on the inner walls of the void with a trap energy of 2.1 eV, or as D2 molecules inside the void with a trap energy of 1.2 eV. Deuterium trapping in polycrystalline tungsten was also studied as a function of the incident fluence, irradiation temperature, and irradiation history. Deuterium trapping in polycrystalline tungsten also occurs primarily at vacancies and voids with the same trap energies as in single crystal tungsten; however, the presence of grain boundaries promotes the formation of large surface blisters with high fluence irradiations at 500 K. In general, D trapping is greater in polycrystalline tungsten than in single crystal tungsten. To simulate mixed materials comprising of carbon (C) and tungsten, tungsten specimens were pre-irradiated with carbon ions prior to D

Deuterium trapping in tungsten

International Nuclear Information System (INIS)

Poon, M.

2004-01-01

Tungsten is one of the primary material candidates being investigated for use in the first-wall of a magnetic confinement fusion reactor. An ion accelerator was used to simulate the type of ion interaction that may occur at a plasma-facing material. Thermal desorption spectroscopy (TDS) was the primary tool used to analyze the effects of the irradiation Secondary ion mass spectroscopy (SIMS) was used to determine the distribution of trapped D in the tungsten specimen. The tritium migration analysis program (TMAP) was used to simulate thermal desorption profiles from the D depth distributions. Fitting of the simulated thermal desorption profiles with the measured TDS results provided values of the D trap energies. . Deuterium trapping in single crystal tungsten was studied as a function of the incident ion fluence, ion flux, irradiation temperature, irradiation history, and surface impurity levels during irradiation The results show that deuterium was trapped at vacancies and voids. Two deuterium atoms could be trapped at a tungsten vacancy, with trapping energies of 1.4 eV and 1.2 eV for the first and second D atoms, respectively. In a tungsten void, D is trapped as atoms adsorbed on the inner walls of the void with a trap energy of 2.1 eV, or as D 2 molecules inside the void with a trap energy of 1.2 eV. . Deuterium trapping in polycrystalline tungsten was also studied as a function of the incident fluence, irradiation temperature, and irradiation history. Deuterium trapping in polycrystalline tungsten also occurs primarily at vacancies and voids with the same trap energies as in single crystal tungsten; however, the presence of grain boundaries promotes the formation of large surface blisters with high fluence irradiations at 500 K. In general, D trapping is greater in polycrystalline tungsten than in single crystal tungsten. To simulate mixed materials comprising of carbon (C) and tungsten, tungsten specimens were pre-irradiated with carbon ions prior to D

Deuterium trapping in tungsten

Energy Technology Data Exchange (ETDEWEB)

Poon, M

2004-07-01

Tungsten is one of the primary material candidates being investigated for use in the first-wall of a magnetic confinement fusion reactor. An ion accelerator was used to simulate the type of ion interaction that may occur at a plasma-facing material. Thermal desorption spectroscopy (TDS) was the primary tool used to analyze the effects of the irradiation Secondary ion mass spectroscopy (SIMS) was used to determine the distribution of trapped D in the tungsten specimen. The tritium migration analysis program (TMAP) was used to simulate thermal desorption profiles from the D depth distributions. Fitting of the simulated thermal desorption profiles with the measured TDS results provided values of the D trap energies. . Deuterium trapping in single crystal tungsten was studied as a function of the incident ion fluence, ion flux, irradiation temperature, irradiation history, and surface impurity levels during irradiation The results show that deuterium was trapped at vacancies and voids. Two deuterium atoms could be trapped at a tungsten vacancy, with trapping energies of 1.4 eV and 1.2 eV for the first and second D atoms, respectively. In a tungsten void, D is trapped as atoms adsorbed on the inner walls of the void with a trap energy of 2.1 eV, or as D{sub 2} molecules inside the void with a trap energy of 1.2 eV. . Deuterium trapping in polycrystalline tungsten was also studied as a function of the incident fluence, irradiation temperature, and irradiation history. Deuterium trapping in polycrystalline tungsten also occurs primarily at vacancies and voids with the same trap energies as in single crystal tungsten; however, the presence of grain boundaries promotes the formation of large surface blisters with high fluence irradiations at 500 K. In general, D trapping is greater in polycrystalline tungsten than in single crystal tungsten. To simulate mixed materials comprising of carbon (C) and tungsten, tungsten specimens were pre-irradiated with carbon ions prior to D

Hybrid simulation research on formation mechanism of tungsten nanostructure induced by helium plasma irradiation

Energy Technology Data Exchange (ETDEWEB)

Ito, Atsushi M., E-mail: ito.atsushi@nifs.ac.jp [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Takayama, Arimichi; Oda, Yasuhiro [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Tamura, Tomoyuki; Kobayashi, Ryo; Hattori, Tatsunori; Ogata, Shuji [Nagoya Institute of Technology, Gokiso-cho, Showa-ku, Nagoya 466-8555 (Japan); Ohno, Noriyasu; Kajita, Shin [Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan); Yajima, Miyuki [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Noiri, Yasuyuki [Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan); Yoshimoto, Yoshihide [University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Saito, Seiki [Kushiro National College of Technology, Kushiro, Hokkaido 084-0916 (Japan); Takamura, Shuichi [Aichi Institute of Technology, 1247 Yachigusa, Yakusa-cho, Toyota 470-0392 (Japan); Murashima, Takahiro [Tohoku University, 6-3, Aramaki-Aza-Aoba, Aoba-Ward, Sendai 980-8578 (Japan); Miyamoto, Mitsutaka [Shimane University, Matsue, Shimane 690-8504 (Japan); Nakamura, Hiroaki [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan)

2015-08-15

The generation of tungsten fuzzy nanostructure by exposure to helium plasma is one of the important problems for the use of tungsten material as divertor plates in nuclear fusion reactors. In the present paper, the formation mechanisms of the helium bubble and the tungsten fuzzy nanostructure were investigated by using several simulation methods. We proposed the four-step process which is composed of penetration step, diffusion and agglomeration step, helium bubble growth step, and fuzzy nanostructure formation step. As the fourth step, the formation of the tungsten fuzzy nanostructure was successfully reproduced by newly developed hybrid simulation combining between molecular dynamics and Monte-Carlo method. The formation mechanism of tungsten fuzzy nanostructure observed by the hybrid simulation is that concavity and convexity of the surface are enhanced by the bursting of helium bubbles in the region around the concavity.

Plasma facing components integration studies for the WEST divertor

Energy Technology Data Exchange (ETDEWEB)

Ferlay, Fabien, E-mail: fabien.ferlay@cea.fr; Missirlian, Marc; Guilhem, Dominique; Firdaouss, Mehdi; Richou, Marianne; Doceul, Louis; Faisse, Frédéric; Languille, Pascal; Larroque, Sébastien; Martinez, André; Proust, Maxime; Louison, Céphise; Jeanne, Florian; Saille, Alain; Samaille, Frank; Verger, Jean-Marc; Bucalossi, Jérôme

2015-10-15

Highlights: • The divertor PFU integration has been studied regarding existing environment. • Magnetic, electric, thermal, hydraulic, mechanical loads and assembly are considered. - Abstract: In the context of the Tokamak Tore-Supra evolution, the CEA aims at transforming it into a test bench for ITER actively cooled tungsten (ACW) plasma facing components (PFC). This project named WEST (Tungsten Environment in Steady state Tokamak) is especially focused on the divertor target. The modification of the machine, by adding two axisymmetric divertors will make feasible an H-mode with an X-point close to the lower divertor. This environment will allow exposing the divertor ACW components up to 20 MW/m{sup 2} heat flux during long pulse. These specifications are well suited to test the ITER-like ACW target elements, respecting the ITER design. One challenge in such machine evolution is to integrate components in an existing vacuum vessel in order to obtain the best achievable performance. This paper deals with the design integration of ITER ACW target elements into the WEST environment considering magnetic, electric, thermal and mechanical loads. The feasibility of installation and maintenance has to be strongly considered as these PFC could be replaced several times. The ports size allows entering a 30° sector of pre-installed tungsten targets which will be plugged as quickly and easily as possible. The main feature of steady state operation is the active cooling, which leads to have many embedded cooling channels and bulky pipes on the PFC module including many connections and sealings between vacuum and water channels. The 30° sector design is now finalized regarding the ITER ACW elements specifications. No major modifications are expected.

Combined adsorption of lithium and oxygen on (111) face of tungsten

International Nuclear Information System (INIS)

Lozovoj, Ya.B.; Smereka, T.P.; Babkin, G.V.; Payukh, B.M.

1986-01-01

A contact potential difference technique has been employed to study the electron-adsorption properties of lithium films on a (111) face of tungsten, preliminary coated with different doses of oxygen. At all the lithium coverages studied the presence of oxygen on the surface leads to a significant decrease of the work function φ min and an increase of the thermal stability of lithium films. For optimal coverage φ=1.8 eV, q=2.2 eV

High Purity Tungsten Spherical Particle Preparation From WC-Co Spent Hard Scrap

Directory of Open Access Journals (Sweden)

Han Chulwoong

2015-06-01

Full Text Available Tungsten carbide-cobalt hard metal scrap was recycled to obtain high purity spherical tungsten powder by a combined hydrometallurgy and physical metallurgy pathway. Selective leaching of tungsten element from hard metal scrap occurs at solid / liquid interface and therefore enlargement of effective surface area is advantageous. Linear oxidation behavior of Tungsten carbide-cobalt and the oxidized scrap is friable to be pulverized by milling process. In this regard, isothermally oxidized Tungsten carbide-cobalt hard metal scrap was mechanically broken into particles and then tungsten trioxide particle was recovered by hydrometallurgical method. Recovered tungsten trioxide was reduced to tungsten particle in a hydrogen environment. After that, tungsten particle was melted and solidified to make a spherical one by RF (Ratio Frequency thermal plasma process. Well spherical tungsten micro-particle was successfully obtained from spent scrap. In addition to the morphological change, thermal plasma process showed an advantage for the purification of feedstock particle.

Thermal shock behaviour of different tungsten grades under varying conditions

Energy Technology Data Exchange (ETDEWEB)

Wirtz, Oliver Marius

2012-07-19

Thermonuclear fusion power plants are a promising option to ensure the energy supply for future generations, but in many fields of research enormous challenges have to be faced. A major step on the way to the prototype fusion reactor DEMO will be ITER which is build in Cadarache, southern France. One of the most critical issues is the field of in-vessel materials and components, in particular the plasma facing materials (PFM). PFMs that will be used in a device like ITER have to withstand severe environmental conditions in terms of steady state and transient thermal loads as well as high particle fluxes such as hydrogen, helium and neutrons. Candidate wall materials are beryllium, tungsten and carbon based materials like CFC (carbon fibre composite). Tungsten is the most promising material for an application in the divertor region with very severe loading conditions and it will most probably also be used as PFM for DEMO. Hence, this work focuses on the investigation of the thermal shock response of different tungsten grades in order to understand the damage mechanisms and to identify material parameters which influence this behaviour under ITER and DEMO relevant operation conditions. Therefore the microstructure and the mechanical and thermal properties of five industrially manufactured tungsten grades were characterised. All five tungsten grades were exposed to transient thermal events with very high power densities of up to 1.27 GWm{sup -2} at varying base temperatures between RT and 600 C in the electron beam device JUDITH 1. The pulse numbers were limited to a maximum of 1000 in order to avoid immoderate workload on the test facility and to have enough time to cover a wide range of loading conditions. The results of this damage mapping enable to define different damage and cracking thresholds for the investigated tungsten grades and to identify certain material parameters which influence the location of these thresholds and the distinction of the induced

Thermal shock behaviour of different tungsten grades under varying conditions

International Nuclear Information System (INIS)

Wirtz, Oliver Marius

2012-01-01

Thermonuclear fusion power plants are a promising option to ensure the energy supply for future generations, but in many fields of research enormous challenges have to be faced. A major step on the way to the prototype fusion reactor DEMO will be ITER which is build in Cadarache, southern France. One of the most critical issues is the field of in-vessel materials and components, in particular the plasma facing materials (PFM). PFMs that will be used in a device like ITER have to withstand severe environmental conditions in terms of steady state and transient thermal loads as well as high particle fluxes such as hydrogen, helium and neutrons. Candidate wall materials are beryllium, tungsten and carbon based materials like CFC (carbon fibre composite). Tungsten is the most promising material for an application in the divertor region with very severe loading conditions and it will most probably also be used as PFM for DEMO. Hence, this work focuses on the investigation of the thermal shock response of different tungsten grades in order to understand the damage mechanisms and to identify material parameters which influence this behaviour under ITER and DEMO relevant operation conditions. Therefore the microstructure and the mechanical and thermal properties of five industrially manufactured tungsten grades were characterised. All five tungsten grades were exposed to transient thermal events with very high power densities of up to 1.27 GWm -2 at varying base temperatures between RT and 600 C in the electron beam device JUDITH 1. The pulse numbers were limited to a maximum of 1000 in order to avoid immoderate workload on the test facility and to have enough time to cover a wide range of loading conditions. The results of this damage mapping enable to define different damage and cracking thresholds for the investigated tungsten grades and to identify certain material parameters which influence the location of these thresholds and the distinction of the induced damages

Optimization of armour geometry and bonding techniques for tungsten-armoured high heat flux components

International Nuclear Information System (INIS)

Giniyatulin, R.N.; Komarov, V.L.; Kuzmin, E.G.; Makhankov, A.N.; Mazul, I.V.; Yablokov, N.A.; Zhuk, A.N.

2002-01-01

Joining of tungsten with copper-based cooling structure and armour geometry optimization are the major aspects in development of the tungsten-armoured plasma facing components (PFC). Fabrication techniques and high heat flux (HHF) tests of tungsten-armoured components have to reflect different PFC designs and acceptable manufacturing cost. The authors present the recent results of tungsten-armoured mock-ups development based on manufacturing and HHF tests. Two aspects were investigated--selection of armour geometry and examination of tungsten-copper bonding techniques. Brazing and casting tungsten-copper bonding techniques were used in small mock-ups. The mock-ups with armour tiles (20x5x10, 10x10x10, 20x20x10, 27x27x10) mm 3 in dimensions were tested by cyclic heat fluxes in the range of (5-20) MW/m 2 , the number of thermal cycles varied from hundreds to several thousands for each mock-up. The results of the tests show the applicability of different geometry and different bonding technique to corresponding heat loading. A medium-scale mock-up 0.6-m in length was manufactured and tested. HHF tests of the medium-scale mock-up have demonstrated the applicability of the applied bonding techniques and armour geometry for full-scale PFC's manufacturing

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

Science.gov (United States)

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.

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

International Nuclear Information System (INIS)

Neu, R.

2006-01-01

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

Study of tungsten surface interaction with plasma streams at DPF-1000U

Directory of Open Access Journals (Sweden)

Ladygina Marina S.

2015-06-01

Full Text Available In this note experimental studies of tungsten (W samples irradiated by intense plasma-ion streams are reported. Measurements were performed using the modified plasma focus device DPF-1000U equipped with an axial gas-puffing system. The main diagnostic tool was a Mechelle®900 optical spectrometer. The electron density of a freely propagating plasma stream (i.e., the plasma stream observed without any target inside the vacuum chamber was estimated on the basis of the half-width of the Dβ spectral line, taking into account the linear Stark effect. For a freely propagating plasma stream the maximum electron density amounted to about 1.3 × 1017 cm−3 and was reached during the maximum plasma compression. The plasma electron density depends on the initial conditions of the experiments. It was thus important to determine first the plasma flow characteristics before attempting any target irradiation. These data were needed for comparison with plasma characteristics after an irradiation of the investigated target. In fact, spectroscopic measurements performed during interactions of plasma streams with the investigated W samples showed many WI and WII spectral lines. The surface erosion was determined from mass losses of the irradiated samples. Changes on the surfaces of the irradiated samples were also investigated with an optical microscope and some sputtering and melting zones were observed.

Induced tungsten melting events in the divertor of ASDEX Upgrade and their influence on plasma performance

International Nuclear Information System (INIS)

Krieger, K.; Lunt, T.; Dux, R.; Janzer, A.; Kallenbach, A.; Mueller, H.W.; Neu, R.; Puetterich, T.; Rohde, V.

2011-01-01

Tungsten rods of 1 x 1 x 3 mm were exposed at the outer divertor plate of ASDEX Upgrade using a manipulator system. Controlled melting of the W-rod in H-mode discharges was induced by moving the outer strike point towards the W-rod position. Visible light emission of ejected W droplets was recorded by fast camera systems. The resulting increase of tungsten concentration in the confined plasma was compared to that induced by W laser ablation into the outer main chamber boundary plasma. The resulting divertor retention expressed as ratio of tungsten core penetration probability from a divertor source to that of a main chamber source is ∼100. Ejected droplets are found to move always in general direction of the plasma flow. The measured magnitude of droplet acceleration shows that droplets are mainly subject to rocket forces and friction forces. Typical initial droplet size can be inferred from the time evolution of the droplet light emission to be ≥100μm.

Disruption simulation experiments in a pulsed plasma accelerator - energy absorption and damage evolution on plasma facing materials

International Nuclear Information System (INIS)

Bolt, H.; Barabash, V.; Gervash, A.; Linke, J.; Lu, L.P.; Ovchinnikov, I.; Roedig, M.

1995-01-01

Plasma accelerators are used as test beds for disruption simulation experiments on plasma facing materials, because the incident energy fluxes and the discharge duration are of similar order as those expected during disruptions in ITER. The VIKA facility was used for the testing of materials under incident energies up to 5 kJ/cm 2 . Different carbon materials, SiC, stainless steel, TZM and tungsten have been tested. From the experimental results a scaling of the ablation with incident energy density was derived. The resulting ablation depth on carbon materials is roughly 2 μm per kJcm -2 of incident energy density. For metals this ablation is much higher due to the partial loss of the melt layer from splashing. For stainless steel an ablation depth of 9.5 μm per kJcm -2 was determined. The result of a linear scaling of the ablation depth with incident energy density is consistent with a previous calorimetric study. (orig.)

On the origin, properties, and implications of asymmetries in the tungsten impurity density in tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Odstrcil, Tomas

2017-07-03

In this thesis, the transport of tungsten ions is studied in the plasma of ASDEX Upgrade tokamak. The plasma facing components of the fusion reactors are expected to be built from high-Z materials such as W, Mo or Fe. These materials provide advantages like a high melting point, small erosion rates, and low tritium retention. However, due to the interaction of the plasma with the wall, ions of this material will be inevitably present also in the main plasma. These ions are not entirely stripped even at fusion plasma temperatures, and therefore emit strong line radiation, which can significantly degrade the performance of the fusion plasma. Thus the understanding and control of impurity transport are of critical importance to the success of fusion. The high mass and charge of the heavy impurities make them susceptible to some of the forces acting upon the plasma, resulting in a poloidal variation of their density. The most prominent are the centrifugal force arising from the plasma rotation and the electric force caused by magnetically trapped non-thermal ions. Furthermore, the poloidal asymmetries should have a significant impact on the radial transport of heavy ions, which was widely ignored up to date. In the present work, the poloidal asymmetries in the heavy impurity density were inferred from the soft X-ray radiation using a newly developed tomographic method. The high accuracy of the tomography and of the model for the centrifugal force allowed to identify for the first time in an experiment the effect of the fast ion distribution produced by neutral beam injection on the poloidal asymmetry of the tungsten density. The measured asymmetry was compared to several fast ion models, and the best match was found with the Monte Carlo code in the TRANSP code suite that includes finite orbits effects of the fast ions. Similarly, fast ions accelerated by ion cyclotron heating and localized mainly in the outboard side of the plasma due to a magnetic trapping and produce

Surface morphology changes to tungsten under exposure to He ions from an electron cyclotron resonance plasma source

Science.gov (United States)

Donovan, David; Maan, Anurag; Duran, Jonah; Buchenauer, Dean; Whaley, Josh

2015-11-01

Exposure of tungsten to low energy (ALMT ITER grade tungsten samples. A similar He plasma exposure stage has now been developed at the University of Tennessee-Knoxville with an improved compact ECR plasma source. Status of the new UTK exposure stage will be discussed as well as planned experiments and new material characterization techniques (EBSD, GIXRD). Work supported by US DOE Contract DE-AC04-94AL85000 and the PSI Science Center.

Development and optimisation of tungsten armour geometry for ITER divertor

International Nuclear Information System (INIS)

Makhankov, A.; Mazul, I.; Safronov, V.; Yablokov, N.

1998-01-01

The plasma facing components (PFC) of the future thermonuclear reactor in great extend determine the time of non-stop operation of the reactor. In current ITER project the most of the divertor PFC surfaces are covered by tungsten armour. Therefore selection of tungsten grade and attachment scheme for joining the tungsten armour to heat sink is a matter of great importance. Two attachment schemes for highly loaded components (up to 20 MW/m 2 ) are described in this paper. The small size mock-ups were manufactured and successfully tested at heat fluxes up to 30 MW/m 2 in screening test and up to 20 MW/m 2 at thermal fatigue test. One mock-up with four different tungsten grades was tested by consequent thermal shock (15 MJ/m 2 at 50 μs) and thermal cycling loading (15 MW/m 2 ). The damages that could lead to mock-up failure were not found but the behaviour of tungsten grades was quite different. (author)

Performance of plasma facing materials under intense thermal loads in tokamaks and stellarators

Energy Technology Data Exchange (ETDEWEB)

Linke, J.; Hirai, T.; Roedig, M.; Singheiser, L. [Forschungszentrum Juelich GmbH, EURATOM Association, Juelich (Germany)

2003-07-01

Beside quasi-stationary plasma operation, short transient thermal pulses with deposited energy densities in the order of several ten MJm{sup -2} are a serious concern for next step devices, in particular for tokamak devices such as ITER. The most serious of these transient events are plasma disruptions. Here a considerable fraction of the plasma energy is deposited on a localized surface area in the divertor strike zone region; the time scale of these events is typically in the order of 1 ms. In spite of the fact that a dense cloud of ablation vapour will form above the strike zone, only partial shielding of the divertor armour from incident plasma particles will occur. As a consequence, thermal shock induced crack formation, vaporization, surface melting, melt layer ejection, and particle emission induced by brittle destruction processes will limit the lifetime of the components. In addition, dust particles (neutron activated metals or tritium enriched carbon) are a serious concern form a safety point of view. Other transient heat loads which occasionally occur in magnetic confinement experiments such as instabilities in the plasma positioning (vertical displacement events) also may cause irreversible damage to plasma facing components (PFC), particularly to metals such as beryllium and tungsten. Another serious damage to PFCs is due to intense fluxes of 14 MeV neutrons in D-T-burning plasma devices. Integrated neutron fluence of several ten dpa in future thermonuclear fusion reactors will degrade essential physical properties of the components (e.g. thermal conductivity); another serious concern is the embrittlement of the heat sink and the plasma facing materials (PFM). (orig.)

High Heat Load Properties of Ultra Fine Grain Tungsten

International Nuclear Information System (INIS)

Zhou, Z.; Du, J.; Ge, C.; Linke, J.; Pintsuk, G.; Song, S.X.

2007-01-01

Full text of publication follows: Tungsten is increasingly considered as a promising candidate armour materials facing the plasma in tokamaks for medium to high heat flux components (EAST, ASDEX, ITER). Fabrication tungsten with ultra fine grain size is considered as an effective way to ameliorate some disadvantages of tungsten, such as its brittleness at room temperature. But the research data on the performance of ultra fine grain tungsten is still very limit. In this work, high heat load properties of pure ultra-fine grain tungsten have been studied. The ultra fine grain tungsten samples with average grain size of 0.2 μm, 1 μm and 3 μm were fabricated by resistance sintering under ultra high pressure. The annealing experiments for the investigation of the material resistance against grain growth have been done by annealing samples in a vacuum furnace at different temperature holding for 2 hours respectively. It is found that recrystallization and grain growth occur at heating temperature of 1250 deg. c. The finer the initial grain sizes of tungsten, the smaller its grain growth grain. The effects of transient high thermal loads (off normal events like disruptions) on tungsten surface morphology have been performed in electron beam test facility JUDITH. The thermal loads tests have been carried out with 4 ms pulses at different power density of 0.22, 0.33, 0.44, 0.55 and 0.88 GW/m 2 respectively. Horizontal cracks formed for all tungsten samples at 0.44 GW/m 2 . Particle erosions occurred for tungsten with 3 μm size at 0.33 GW/m 2 and for tungsten with 0.2 and 1 μm size at 0.55 GW/m 2 . The weight loss of tungsten with 0.2, 1 and 3 μm size are 2,0.1,0.6 mg respectively at 0.88 GW/m 2 . The effects of a large number of very short transient repetitive thermal loads (ELM-like) on tungsten surface morphology also have been performed by using a fundamental wave of a YAG laser. It is found that tungsten with 0.2 μm size has the best performance. (authors)

High Heat Load Properties of Ultra Fine Grain Tungsten

Energy Technology Data Exchange (ETDEWEB)

Zhou, Z.; Du, J.; Ge, C. [Lab. of Special Ceramic and P/M, University of Science and Technology, 100083 Beijing (China); Linke, J.; Pintsuk, G. [FZJ-Forschungszentrum Juelich GmbH, Association Euratom-FZJ, Institut fur Plasmaphysik, Postfach 1913, D-52425 Juelich (Germany); Song, S.X. [Research Center on Fusion Materials (RCFM), University of Science and Technology Beijing (USTB), 100083 Beijing (China)

2007-07-01

Full text of publication follows: Tungsten is increasingly considered as a promising candidate armour materials facing the plasma in tokamaks for medium to high heat flux components (EAST, ASDEX, ITER). Fabrication tungsten with ultra fine grain size is considered as an effective way to ameliorate some disadvantages of tungsten, such as its brittleness at room temperature. But the research data on the performance of ultra fine grain tungsten is still very limit. In this work, high heat load properties of pure ultra-fine grain tungsten have been studied. The ultra fine grain tungsten samples with average grain size of 0.2 {mu}m, 1 {mu}m and 3 {mu}m were fabricated by resistance sintering under ultra high pressure. The annealing experiments for the investigation of the material resistance against grain growth have been done by annealing samples in a vacuum furnace at different temperature holding for 2 hours respectively. It is found that recrystallization and grain growth occur at heating temperature of 1250 deg. c. The finer the initial grain sizes of tungsten, the smaller its grain growth grain. The effects of transient high thermal loads (off normal events like disruptions) on tungsten surface morphology have been performed in electron beam test facility JUDITH. The thermal loads tests have been carried out with 4 ms pulses at different power density of 0.22, 0.33, 0.44, 0.55 and 0.88 GW/m{sup 2} respectively. Horizontal cracks formed for all tungsten samples at 0.44 GW/m{sup 2}. Particle erosions occurred for tungsten with 3 {mu}m size at 0.33 GW/m{sup 2} and for tungsten with 0.2 and 1 {mu}m size at 0.55 GW/m{sup 2}. The weight loss of tungsten with 0.2, 1 and 3 {mu}m size are 2,0.1,0.6 mg respectively at 0.88 GW/m{sup 2}. The effects of a large number of very short transient repetitive thermal loads (ELM-like) on tungsten surface morphology also have been performed by using a fundamental wave of a YAG laser. It is found that tungsten with 0.2 {mu}m size has

Recovery of tungsten surface with fiber-form nanostructure by the argon plasma irradiation at a high surface temperature

International Nuclear Information System (INIS)

Takamura, Shuichi; Miyamoto, Takanori

2011-01-01

One of the serious concerns for tungsten materials in fusion devices is the radiation defects caused by helium plasma irradiation, while the helium is one of fusion products. Fiber-formed nanostructure is worried to have a possible weakness against the plasma heat flux and may destroy the reflectivity as an optical mirror. In this communication an interesting method for a recovery of such a tungsten surface is shown. (author)

Elaboration of functionally graded materials for plasma facing components of the thermonuclear machines

International Nuclear Information System (INIS)

Autissier, Emmanuel

2014-01-01

The objective of this study was to develop a Functionally Graded Material (FGM) W/Cu to replace the compliance layer (Cu-OFHC) in the plasma facing components of thermonuclear fusion reactor like ITER. The peculiarity of this work is to elaborate these materials without exceeding the melting temperature of copper in order to control its microstructure. The co-sintering is the most attractive solution to achieve this goal. The first phase of this study has been to decrease the sintering temperature of the tungsten to achieve this co-sintering. The elaboration of a Functionally Graded Materials being delicate, thermomechanical calculations were performed in order to determine the number and chemical composition in order to increase the lifespan of Plasma Facing Components. Spark Plasma Sintering conditions were optimized in order to achieve maximum density of W x Cu 1-x composites. The effect of copper content and density of the W x Cu 1-x composites on thermal and mechanical properties was investigated. The SPS conditions were applied for W/CuCrZr assemblies with a compliance layer composed of several interlayers. The importance of time for the integrity of assemblies thereof has been highlighted. The study of the dwell time during W/CuCrZr assembly leads to identify a parameter to characterize the integrity of the interface regardless of the composition and the nature of the layer of compliance. Moreover, the phenomena associated with the formation of the interface assembly have been identified. The interface W/W x Cu 1-x is formed by the extrusion of the copper layer of the W x Cu 1-x inside the tungsten porosities. The W y Cu 1-y /CuCrZr interface is formed by copper migration of CuCrZr layer inside the W y Cu 1-y layer. Finally optimization assembly conditions showed that the mechanical stresses due to the densification of the Functionally Graded Materials can be limited by sintering the FGM before the assembly. (author)

The influence of substrate temperature and spraying distance on the properties of plasma sprayed tungsten and steel coatings deposited in a shrouding chamber

Czech Academy of Sciences Publication Activity Database

Matějíček, Jiří; Vilémová, Monika; Nevrlá, Barbara; Kocmanová, Lenka; Veverka, Jakub; Halasová, Martina; Hadraba, Hynek

2017-01-01

Roč. 318, May (2017), s. 217-223 ISSN 0257-8972. [International Meeting on Thermal Spraying (RIPT)/7./. Limoges, 09.12.2015-11.12.2015] R&D Projects: GA ČR GB14-36566G EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 ; RVO:68081723 Keywords : Tungsten * Steel * Atmospheric plasma spraying * Shrouding * Substrate temperature * Fusion reactor materials * Plasma facing components Subject RIV: JK - Corrosion ; Surface Treatment of Materials; JK - Corrosion ; Surface Treatment of Materials (UFM-A) OBOR OECD: Coating and films; Coating and films (UFM-A) Impact factor: 2.589, year: 2016 http://www.sciencedirect.com/science/ article /pii/S0257897216310520

Hydrogen retention studies on lithiated tungsten exposed to glow discharge plasmas under varying lithiation environments using Thermal Desorption Spectroscopy and mass spectrometry

Energy Technology Data Exchange (ETDEWEB)

Castro, A. de, E-mail: alfonso.decastro@ciemat.es [Fusion National Laboratory-CIEMAT, Av Complutense 40, 28040 Madrid (Spain); Valson, P. [Max-Planck-Institut für Plasmaphysik, Wendelsteinstraße 1, 17491 Greifswald (Germany); Tabarés, F.L. [Fusion National Laboratory-CIEMAT, Av Complutense 40, 28040 Madrid (Spain)

2017-04-15

For the design of a Fusion Reactor based on a liquid lithium divertor target and a tungsten first wall at high temperature, the interaction of the wall material with plasmas of significant lithium content must be assessed, as issues like fuel retention, tungsten embrittlement and enhanced sputtering may represent a showstopper for the selection of the first wall material compatible with the presence of liquid metal divertor. In this work we address this topic for the first time at the laboratory level, hot W samples (100 °C) have been exposed to Glow Discharges of H{sub 2} or Li-seeded H{sub 2} followed by in situ thermal desorption studies (TDS) of the uptake of H{sub 2} on the samples. Pure and pre-lithiated tungsten was investigated in order to evaluate the differential effect of Li ion implantation on H retention. Global particle balance was also used for the determination of trapped H into the full W wall of the plasma chamber. A factor of 3-4 lower retention was deduced for samples and main W wall exposed to H/Li plasma than that measured on pre-lithiated W.

Results of high heat flux tests of tungsten divertor targets under plasma heat loads expected in ITER and tokamaks (review)

Energy Technology Data Exchange (ETDEWEB)

Budaev, V. P., E-mail: budaev@mail.ru [National Research Centre Kurchatov Institute (Russian Federation)

2016-12-15

Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach ~10MW m{sup −2} in the steady state of DT discharges, increasing to ~0.6–3.5 GW m{sup −2} under disruptions and ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production of submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma–wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.

Effect of high-flux H/He plasma exposure on tungsten damage due to transient heat loads

Energy Technology Data Exchange (ETDEWEB)

De Temmerman, G., E-mail: gregory.detemmerman@iter.org [FOM Institute DIFFER, Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregion Cluster, Postbus 1207, 3430BE Nieuwegein (Netherlands); ITER Organization, Route de Vinon sur Verdon, CS 90 096, 13067 Saint Paul-lez-Durance (France); Morgan, T.W.; Eden, G.G. van; Kruif, T. de [FOM Institute DIFFER, Dutch Institute for Fundamental Energy Research, Association EURATOM-FOM, Trilateral Euregion Cluster, Postbus 1207, 3430BE Nieuwegein (Netherlands); Wirtz, M. [Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research – Microstructure and Properties of Materials (IEK-2), EURATOM Association, 52425 Jülich (Germany); Matejicek, J.; Chraska, T. [Institute of Plasma Physics, Association EURATOM-IPP, CR Prague (Czech Republic); Pitts, R.A. [ITER Organization, Route de Vinon sur Verdon, CS 90 096, 13067 Saint Paul-lez-Durance (France); Wright, G.M. [MIT Plasma Science and Fusion Center, 77 Massachusetts Ave., Cambridge, MA 02139 (United States)

2015-08-15

The thermal shock behaviour of tungsten exposed to high-flux plasma is studied using a high-power laser. The cases of laser-only, sequential laser and hydrogen (H) plasma and simultaneous laser plus H plasma exposure are studied. H plasma exposure leads to an embrittlement of the material and the appearance of a crack network originating from the centre of the laser spot. Under simultaneous loading, significant surface melting is observed. In general, H plasma exposure lowers the heat flux parameter (F{sub HF}) for the onset of surface melting by ∼25%. In the case of He-modified (fuzzy) surfaces, strong surface deformations are observed already after 1000 laser pulses at moderate F{sub HF} = 19 MJ m{sup −2} s{sup −1/2}, and a dense network of fine cracks is observed. These results indicate that high-fluence ITER-like plasma exposure influences the thermal shock properties of tungsten, lowering the permissible transient energy density beyond which macroscopic surface modifications begin to occur.

Two component tungsten powder injection molding – An effective mass production process

International Nuclear Information System (INIS)

Antusch, Steffen; Commin, Lorelei; Mueller, Marcus; Piotter, Volker; Weingaertner, Tobias

2014-01-01

Tungsten and tungsten-alloys are presently considered to be the most promising materials for plasma facing components for future fusion power plants. The Karlsruhe Institute of Technology (KIT) divertor design concept for the future DEMO power plant is based on modular He-cooled finger units and the development of suitable mass production methods for such parts was needed. A time and cost effective near-net-shape forming process with the advantage of shape complexity, material utilization and high final density is Powder Injection Molding (PIM). This process allows also the joining of two different materials e.g. tungsten with a doped tungsten alloy, without brazing. The complete technological process of 2-Component powder injection molding for tungsten materials and its application on producing real DEMO divertor parts, characterization results of the finished parts e.g. microstructure, hardness, density and joining zone quality are discussed in this contribution

Fuel hydrogen retention of tungsten and the reduction by inert gas glow discharges

Energy Technology Data Exchange (ETDEWEB)

Hino, T., E-mail: tomhino@qe.eng.hokudai.ac.jp [Laboratory of Plasma Physics and Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Yamauchi, Y.; Kimura, Y. [Laboratory of Plasma Physics and Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Nishimura, K. [National Institute for Fusion Science, Toki-shi, Gifu-ken 509-5292 (Japan); Ueda, Y. [Graduate School of Engineering, Osaka University, Suita-shi 565-0872 (Japan)

2012-08-15

Highlights: Black-Right-Pointing-Pointer The performances of inert gas glow discharges for reduction of fuel hydrogen retention in tungsten were systematically investigated. Black-Right-Pointing-Pointer For the tungsten with rough surface structure, the reduction of fuel hydrogen retention by inert gas discharges is quite small. Black-Right-Pointing-Pointer The deuterium glow discharge is quite useful to reduce the tritium retention in plasma facing walls in fusion reactor. Black-Right-Pointing-Pointer The wall baking with temperature higher than 700-800 K is also useful to reduce the tritium retention in plasma facing walls. - Abstract: Polycrystalline tungsten was exposed to deuterium glow discharge followed by He, Ne or Ar glow discharge. The amount of retained deuterium in the tungsten was measured using residual gas analysis. The amount of desorbed deuterium during the inert gas glow discharge was also measured. The amount of retained deuterium was 2-3 times larger compared with a case of stainless steel. The ratios of desorbed amount of deuterium by He, Ne and Ar glow discharges were 4.6, 3.1 and 2.9%, respectively. These values were one order of magnitude smaller compared with the case of stainless steel. The inert gas glow discharge is not suitable to reduce the fuel hydrogen retention for tungsten walls. However, the wall baking with a temperature higher than 700 K is suitable to reduce the fuel hydrogen retention. It is also shown that the use of deuterium glow discharge is effective to reduce the in-vessel tritium inventory in fusion reactors through the hydrogen isotope exchange.

Effect of rhenium addition on tungsten fuzz formation in helium plasmas

NARCIS (Netherlands)

Khan, A.; De Temmerman, G.; Morgan, T. W.; M. B. Ward,

2016-01-01

The effect of the addition of rhenium to tungsten on the formation of a nanostructure referred to as ‘fuzz’ when exposed to helium plasmas at fusion relevant ion fluxes was investigated in the Magnum and Pilot PSI devices at the FOM Institute DIFFER. The effect rhenium had on fuzz growth was seen to

Simulating tritium retention in tungsten with a multiple trap model in the TMAP code

International Nuclear Information System (INIS)

Merrill, Brad J.; Shimada, Masashi; Humrickhouse, Paul W.

2013-01-01

Accurately predicting the quantity of tritium retained in plasma facing components is a key safety issue for licensing future fusion power reactors. Retention of tritium in the lattice damage caused when high energy neutrons collide with atoms in the structural material of the reactor's plasma facing components (PFCs) is an area of ongoing experimental research at the Idaho National Laboratory (INL) under the US/Japan TITAN collaboration. Recent experiments with the Tritium Plasma Experiment (TPE), located in the INL's Safety and Tritium Applied Research (STAR) facility, demonstrate that this damage can only be simulated by computer codes like the Tritium Migration Analysis Program (TMAP) if one assumes that the lattice damage produced by these neutrons results in multiple types of hydrogen traps (energy wells) within the material, each possessing a different trap energy and density. Previous attempts to simulate the quantity of deuterium released from neutron irradiated TPE tungsten targets indicated that at least six different traps are required by TMAP to model this release. In this paper we describe a recent extension of the TMAP trap site model to include as many traps as required by the user to simulate retention of tritium in neutron damaged tungsten. This model has been applied to data obtained for tungsten irradiated to a damage level of 0.025 dpa in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) after exposure to a plasma in TPE. (author)

Results of high heat flux tests and structural analysis of the new solid tungsten divertor tile for ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Jaksic, Nikola, E-mail: nikola.jaksic@ipp.mpg.de; Greuner, Henri; Herrmann, Albrecht; Böswirth, Bernd; Vorbrugg, Stefan

2015-10-15

Highlights: • The main motivation for the HHF investigation of tungsten tiles was an untypical deformation of some specimens under thermal loading, observed during the previous tests in GLADIS test facility. • A nonlinear finite element (FE) model for simulations of the GLADIS tests has been built. • The unexpected plastic deformations are mainly caused by internal stresses due to the manufacturing process. The small discrepancies among the FEA investigated and measured plastic deformations are most likely caused, beside of the practical difficulties by measuring of low items, also by tile internal stresses. • The influences of the residual stresses caused by special production processes have to be taken into account by design of the structural part made of solid tungsten. - Abstract: Tungsten as plasma-facing material for fusion devices is currently the most favorable candidate. In general solid tungsten is used for shielding the plasma chamber interior against the high heat generated from the plasma. For the purposes of implementation at ASDEX Upgrade and as a contribution to ITER the thermal performance of tungsten tiles has been extensively tested in the high heat flux test facility GLADIS during the development phase and beyond. These tests have been performed on full scale tungsten tile prototypes including their clamping and cooling structure. Simulating the adiabatically thermal loading due to plasma operation in ASDEX Upgrade, the tungsten tiles have been subjected to a thermal load with central heat flux of 10–24 MW/m{sup 2} and absorbed energy between 370 and 680 kJ. This loading results in maximum surface temperatures between 1300 °C and 2800 °C. The tests in GLADIS have been accompanied by intensive numerical investigations using FEA methods. For this purpose a multiple nonlinear finite element model has been set up. This paper discusses the main results of the high heat flux final tests and their numerical simulation. Moreover, first

Results of high heat flux tests and structural analysis of the new solid tungsten divertor tile for ASDEX Upgrade

International Nuclear Information System (INIS)

Jaksic, Nikola; Greuner, Henri; Herrmann, Albrecht; Böswirth, Bernd; Vorbrugg, Stefan

2015-01-01

Highlights: • The main motivation for the HHF investigation of tungsten tiles was an untypical deformation of some specimens under thermal loading, observed during the previous tests in GLADIS test facility. • A nonlinear finite element (FE) model for simulations of the GLADIS tests has been built. • The unexpected plastic deformations are mainly caused by internal stresses due to the manufacturing process. The small discrepancies among the FEA investigated and measured plastic deformations are most likely caused, beside of the practical difficulties by measuring of low items, also by tile internal stresses. • The influences of the residual stresses caused by special production processes have to be taken into account by design of the structural part made of solid tungsten. - Abstract: Tungsten as plasma-facing material for fusion devices is currently the most favorable candidate. In general solid tungsten is used for shielding the plasma chamber interior against the high heat generated from the plasma. For the purposes of implementation at ASDEX Upgrade and as a contribution to ITER the thermal performance of tungsten tiles has been extensively tested in the high heat flux test facility GLADIS during the development phase and beyond. These tests have been performed on full scale tungsten tile prototypes including their clamping and cooling structure. Simulating the adiabatically thermal loading due to plasma operation in ASDEX Upgrade, the tungsten tiles have been subjected to a thermal load with central heat flux of 10–24 MW/m"2 and absorbed energy between 370 and 680 kJ. This loading results in maximum surface temperatures between 1300 °C and 2800 °C. The tests in GLADIS have been accompanied by intensive numerical investigations using FEA methods. For this purpose a multiple nonlinear finite element model has been set up. This paper discusses the main results of the high heat flux final tests and their numerical simulation. Moreover, first results

Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

International Nuclear Information System (INIS)

Escourbiac, F; Richou, M; Guigon, R; Durocher, A; Schlosser, J; Grosman, A; Constans, S; Merola, M; Riccardi, B

2009-01-01

Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

Science.gov (United States)

Escourbiac, F.; Richou, M.; Guigon, R.; Constans, S.; Durocher, A.; Merola, M.; Schlosser, J.; Riccardi, B.; Grosman, A.

2009-12-01

Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

Energy Technology Data Exchange (ETDEWEB)

Escourbiac, F; Richou, M; Guigon, R; Durocher, A; Schlosser, J; Grosman, A [CEA/IRFM, F-13108, Saint-Paul-lez-Durance (France); Constans, S [AREVA-NP, Le Creusot (France); Merola, M [ITER Organization, Cadarache (France); Riccardi, B [Fusion For Energy, Barcelona (Spain)], E-mail: frederic.escourbiac@cea.fr

2009-12-15

Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

Thermal Spray Coating of Tungsten for Tokamak Device

International Nuclear Information System (INIS)

Jiang Xianliang; Gitzhofer, F; Boulos, M I

2006-01-01

Thermal spray, such as direct current (d.c.) plasma spray or radio frequency induced plasma spray, was used to deposit tungsten coatings on the copper electrodes of a tokamak device. The tungsten coating on the outer surface of one copper electrode was formed directly through d.c. plasma spraying of fine tungsten powder. The tungsten coating/lining on the inner surface of another copper electrode could be formed indirectly through induced plasma spraying of coarse tungsten powder. Scanning electron microscopy (SEM) was used to examine the cross section and the interface of the tungsten coating. Energy Dispersive Analysis of X-ray (EDAX) was used to analyze the metallic elements attached to a separated interface. The influence of the particle size of the tungsten powder on the density, cracking behavior and adhesion of the coating is discussed. It is found that the coarse tungsten powder with the particle size of 45 ∼ 75 μm can be melted and the coating can be formed only by using induced plasma. The coating deposited from the coarse powder has much higher cohesive strength, adhesive strength and crack resistance than the coating made from the fine powder with a particle size of 5 μm

Plasma-wall interaction and plasma facing materials

International Nuclear Information System (INIS)

Tanabe, Tetsuo; Miyahara, Akira.

1990-01-01

The recognition that plasma-wall interaction plays the essential role from both standpoints of energy balance and particle balance for realizing nuclear fusion reactors has become to prevail. However, on how each elementary process acts and what competitive effect the synthetic action brings about, the stage of doing the qualitative discussion has just come, and the quantitative investigation is the problem for the future. In this paper, the plasma-wall interaction as seen from the research field of plasma-facing materials is discussed centering around graphite materials which have been mostly used at present, and the present status of the research and development on the problems of impurities, hydrogen recycling and heat resistance and radiation resistance is mentioned. Moreover, the problems are pointed out, and the course for the future is looked for. The recent experiment with large tokamaks adopted graphite or carbon as the plasma-facing materials, and the reduction of metallic impurities in plasma showed the clear improvement of plasma confinement characteristics. However, for the next device which requires forced cooling, the usability of graphite is doubtful. (K.I.) 51 refs

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

Science.gov (United States)

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

Carbon and tungsten effect on characteristics of sputtered and re-deposited beryllium target layers under deuteron bombardment

International Nuclear Information System (INIS)

Danelyan, L.S.; Gureev, V.M.; Elistratov, N.G.

2004-01-01

The behavior of the plasma facing Be-elements in the International Thermonuclear Experimental Reactor ITER will be affected by the re-deposition of other eroded plasma facing materials. The effect of carbon- and tungsten-additions on the microstructure, chemical composition and hydrogen isotope accumulation in the sputtered and re-deposited layers of beryllium TGP-56 at its interaction with 200 - 300 eV hydrogen isotope ions was studied in the MAGRAS facility equipped with a magnetron sputtering system. (author)

Properties of deposited layer formed by interaction with Be seeded D–He mixture plasma and tungsten

Energy Technology Data Exchange (ETDEWEB)

Tokunaga, K., E-mail: tokunaga@riam.kyushu-u.ac.jp [Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan); Baldwin, M.J.; Nishijima, D.; Doerner, R.P. [Center for Energy Research, University of California at San Diego, 9500 Gilman Drive, La Jolla, CA 92093-0417 (United States); Nagata, S. [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai 980-8577 (Japan); Tsuchiya, B. [Department of General Education, Faculty of Science and Technology, Meiji University, 1-501 Shiogamaguchi, Tempaku-ku, Nagoya, 468-8502 (Japan); Kurishita, H. [International Research Center for Nuclear Materials Science, IMR, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Fujiwara, T.; Araki, K.; Miyamoto, Y. [Research Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan); Ohno, N. [School of Engineering, Nagoya University, Nagoya 464-8603 (Japan); Ueda, Y. [Graduate School of Engineering, Osaka University, Suita, Osaka 565-0871 (Japan)

2013-11-15

Be-seeded, high-flux, deuterium/helium mixture plasma exposure experiments on tungsten target materials have been performed to simulate ITER all tungsten divertor erosion/modification and deposition phenomena. The exposure conditions are kept fixed at a typical low-ion-energy of 60 eV and a flux of 3–6 × 10{sup 22}/m{sup 2}/s. Sample temperature is 1123 K and plasma exposure times spanning 1050–10,100 s are explored. The typical ratio of He/D ions is 0.2 and Be content is 0.2%. A He-induced nanostructure layer is formed on the exposure surfaces of tungsten materials and the surface of the nanostructure is covered by a thin layer of Be and O. A fraction of the re-eroded Be from the target is deposited on a glassy carbon plate with line of sight to the tungsten target. Rutherford backscattering spectrometry analyses show that the Be redeposit layer is in the form of laminae. Small amounts of Mo, W and C are also found in the redeposited Be layer. Elastic recoil detection analyses show that D, He and H are also included in the redeposited Be layer.

Tungsten oxide thin film exposed to low energy D and He plasma: evidence for a thermal enhancement of the erosion yield

Science.gov (United States)

Hijazi, Hussein; Martin, C.; Roubin, P.; Addab, Y.; Cabie, C.; Pardanaud, C.; Bannister, M.; Meyer, F.

2017-10-01

Nanocrystalline tungsten oxide thin films (25 nm - 250 nm thickness) produced by thermal oxidation of a tungsten substrate were exposed to low energy D and He plasma. Low energy D plasma exposure (11 eV/D+) of these films have resulted in the formation of a tungsten bronze (DxWO3) clearly observed by Raman microscopy. D plasma bombardment (4 1021 m-2) has also induced a color change of the oxide layer which is similar to the well-known electro-chromic effect and has been named ``plasma-chromic effect''. To unravel physical and chemical origins of the modifications observed under exposure, similar tungsten oxide films were also exposed to low energy helium plasma (20 eV/He+) . Due to the low fluence (4 1021 m-2) and low ion energy (20 eV), at room temperature, He exposure has induced only very few morphological and structural modifications. On the contrary, at 673 K, significant erosion is observed, which gives evidence for an unexpected thermal enhancement of the erosion yield. We present here new results concerning He beam exposures at low fluence (4 1021 m-2) varying the He+ energy from 20 eV to 320 eV to measure the tungsten oxide sputtering threshold energy. Detailed analyses before/after exposure to describe the D and He interaction with the oxide layer, its erosion and structural modification at the atomic and micrometer scale will be presented.

Exposures of tungsten nanostructures to divertor plasmas in DIII-D

International Nuclear Information System (INIS)

Rudakov, D L; Doerner, R P; Baldwin, M J; Boedo, J A; Hollmann, E M; Moyer, R A; Wong, C P C; Chrobak, C P; Guo, H Y; Leonard, A W; Pace, D C; Thomas, D M; Wright, G M; Abrams, T; Briesemeister, A R; McLean, A G; Fenstermacher, M E; Lasnier, C J; Watkins, J G

2016-01-01

Tungsten nanostructures (W-fuzz) prepared in the PISCES-A linear device have been found to survive direct exposure to divertor plasmas in DIII-D. W-fuzz was exposed in the lower divertor of DIII-D using the divertor material evaluation system. Two samples were exposed in lower single null (LSN) deuterium H-mode plasmas. The first sample was exposed in three discharges terminated by vertical displacement event disruptions, and the second in two discharges near the lowered X-point. More recently, three samples were exposed near the lower outer strike point in predominantly helium H-mode LSN plasmas. In all cases, the W-fuzz survived plasma exposure with little obvious damage except in the areas where unipolar arcing occurred. Arcing is effective in W-fuzz removal, and it appears that surfaces covered with W-fuzz can be more prone to arcing than smooth W surfaces. (paper)

Behavior and microstructural changes in different tungsten-based materials under pulsed plasma loading

Directory of Open Access Journals (Sweden)

M. Vilémová

2016-12-01

Full Text Available In this study, morphological, microstructural and phase changes of four types of tungsten materials after exposure to dense deuterium plasma were examined. The microstructures of the prepared materials mutually differ by the porosity, grain size and phase content. It was found that inherent porosity of sintered materials leads to a specific mechanism of erosion and might be a significant source of dust in the case of materials with higher porosity. Further, a preferential erosion of the dispersed particles by melting and evaporation and subsequent formation of thin film on the surface of W-Y2O3 was described as well.

Tungsten migration studies by controlled injection of volatile compounds

Energy Technology Data Exchange (ETDEWEB)

Rubel, M., E-mail: rubel@kth.se [Royal Institute of Technology (KTH), Association EURATOM-VR, Stockholm (Sweden); Coenen, J. [IEK-4, Plasma Physics, Forschungszentrum Jülich, Association EURATOM-FZJ, Jülich (Germany); Ivanova, D. [Royal Institute of Technology (KTH), Association EURATOM-VR, Stockholm (Sweden); Möller, S. [IEK-4, Plasma Physics, Forschungszentrum Jülich, Association EURATOM-FZJ, Jülich (Germany); Petersson, P. [Royal Institute of Technology (KTH), Association EURATOM-VR, Stockholm (Sweden); Brezinsek, S.; Kreter, A.; Philipps, V.; Pospieszczyk, A.; Schweer, B. [IEK-4, Plasma Physics, Forschungszentrum Jülich, Association EURATOM-FZJ, Jülich (Germany)

2013-07-15

Volatile tungsten hexa-fluoride was locally injected into the TEXTOR tokamak as a marker for material migration studies. The injection was accompanied by puffing N-15 rare isotope as a nitrogen tracer in discharges with edge cooling by impurity seeding. The objective was to assess material balance by qualitative and quantitative determination of a global and local deposition pattern, material mixing effects and fluorine residence in plasma-facing components. Spectroscopy and ex situ ion beam analysis techniques were used. Tungsten was detected on all types of limiter tiles and short-term probes retrieved from the vessel. Over 80% of the injected W was identified. The largest tungsten concentration, 1 × 10{sup 18} cm{sup −2}, was in the vicinity of the gas inlet. Co-deposits contained tungsten and a mix of light isotopes: H, D, He-4, B-10, B-11, C-12, C-13, N-14, N-15, O-16 and small quantities of F-19 thus showing that both He and nitrogen are trapped following wall conditioning (He glow) and edge cooling.

Integrated modelling of the edge plasma and plasma facing components

International Nuclear Information System (INIS)

Coster, D.P.; Bonnin, X.; Mutzke, A.; Schneider, R.; Warrier, M.

2007-01-01

Modelling of the interaction between the edge plasma and plasma facing components (PFCs) has tended to place more emphasis on either the plasma or the PFCs. Either the PFCs do not change with time and the plasma evolution is studied, or the plasma is assumed to remain static and the detailed interaction of the plasma and the PFCs are examined, with no back-reaction on the plasma taken into consideration. Recent changes to the edge simulation code, SOLPS, now allow for changes in both the plasma and the PFCs to be considered. This has been done by augmenting the code to track the time-development of the properties of plasma facing components (PFCs). Results of standard mixed-materials scenarios (base and redeposited C; Be) are presented

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

Science.gov (United States)

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.

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Riccardi, B. [Associazione Euratom-ENEA sulla Fusione, CR Frascati, P.B. 65 00044 Frascati, Roma (Italy)]. E-mail: riccardi@frascati.enea.it; Montanari, R. [Dipartimento di Ingegneria Meccanica, Universita di Roma, Tor Vergata, 00133 Roma (Italy); Casadei, M. [Centro Sviluppo Materiali, 00100 Roma (Italy); Costanza, G. [Dipartimento di Ingegneria Meccanica, Universita di Roma, Tor Vergata, 00133 Roma (Italy); Filacchioni, G. [ENEA CR Casaccia, I-00060 S. M. di Galeria, Roma (Italy); Moriani, A. [Associazione Euratom-ENEA sulla Fusione, CR Frascati, P.B. 65 00044 Frascati, Roma (Italy)

2006-06-30

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.

Observation of reduction of secondary electron emission from helium ion impact due to plasma-generated nanostructured tungsten fuzz

International Nuclear Information System (INIS)

Hollmann, E M; Doerner, R P; Nishijima, D; Pigarov, A Yu

2017-01-01

Growth of nanostructured fuzz on a tungsten target in a helium plasma is found to cause a significant (∼3×) reduction in ion impact secondary electron emission in a linear plasma device. The ion impact secondary electron emission is separated from the electron impact secondary electron emission by varying the target bias voltage and fitting to expected contributions from electron impact, both thermal and non-thermal; with the non-thermal electron contribution being modeled using Monte-Carlo simulations. The observed (∼3×) reduction is similar in magnitude to the (∼2×) reduction observed in previous work for the effect of tungsten fuzz formation on secondary electron emission due to electron impact. It is hypothesized that the observed reduction results from re-absorption of secondary electrons in the tungsten fuzz. (paper)

Determination of elastic modulus and residual stress of plasma-sprayed tungsten coating on steel substrate

Science.gov (United States)

You, J. H.; Höschen, T.; Lindig, S.

2006-01-01

Plasma-sprayed tungsten, which is a candidate material for the first wall armour, shows a porous, heterogeneous microstructure. Due to its characteristic morphology, the properties are significantly different from those of its dense bulk material. Measurements of the elastic modulus of this coating have not been reported in the literature. In this work Young's modulus of highly porous plasma-sprayed tungsten coatings deposited on steel (F82H) substrates was measured. For the fabrication of the coating system the vacuum plasma-spray process was applied. Measurements were performed by means of three-point and four-point bending tests. The obtained modulus values ranged from 53 to 57 GPa. These values could be confirmed by the test result of a detached coating strip, which was 54 GPa. The applied methods produced consistent results regardless of testing configurations and specimen sizes. The errors were less than 1%. Residual stress of the coating was also estimated.

Determination of elastic modulus and residual stress of plasma-sprayed tungsten coating on steel substrate

International Nuclear Information System (INIS)

You, J.H.; Hoeschen, T.; Lindig, S.

2006-01-01

Plasma-sprayed tungsten, which is a candidate material for the first wall armour, shows a porous, heterogeneous microstructure. Due to its characteristic morphology, the properties are significantly different from those of its dense bulk material. Measurements of the elastic modulus of this coating have not been reported in the literature. In this work Young's modulus of highly porous plasma-sprayed tungsten coatings deposited on steel (F82H) substrates was measured. For the fabrication of the coating system the vacuum plasma-spray process was applied. Measurements were performed by means of three-point and four-point bending tests. The obtained modulus values ranged from 53 to 57 GPa. These values could be confirmed by the test result of a detached coating strip, which was 54 GPa. The applied methods produced consistent results regardless of testing configurations and specimen sizes. The errors were less than 1%. Residual stress of the coating was also estimated

Effects of fibre-form nanostructures on particle emissions from a tungsten surface in plasmas

International Nuclear Information System (INIS)

Takamura, S.; Miyamoto, T.; Ohno, N.

2012-01-01

The effects of fibre-form nanostructure of a tungsten surface on both electron emission and sputtering in helium/argon plasmas are represented. Generally, a nano-fibre forest, the so-called ‘fuzz’, made of tungsten with helium gas inside is found to have the tendency of suppressing the particle emission substantially. The electron emission comes from the impact of high-energy primary electrons. In addition, a deeply biased tungsten target, which inhibits the influx of even energetic primary electrons, seems to produce an electron emission, and it may be suppressed on the way to nanostructure formation on the surface of the W target. Such an emission process is discussed here. The sputtering yield of the He-damaged tungsten surface with the fibre-form nanostructure depends on the surface morphology while the sputtering itself changes the surface morphology, so that the time evolutions of sputtering yield from the W surface with an originally well-developed nanostructure are found to show a minimum in sputtering yield, which is about a half for the fresh nanostructured tungsten and roughly one-fifth of the yield for the original flat normal tungsten surface. The surface morphology at that time is, for the first time, made clear with field emission scanning electron microscopy observation. The physical mechanism for the appearance of such a minimum in sputtering yield is discussed. (paper)

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

Science.gov (United States)

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.

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

International Nuclear Information System (INIS)

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

1996-01-01

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. (orig.)

A review of the US joining technologies for plasma facing components in the ITER fusion reactor

International Nuclear Information System (INIS)

Odegard, B.C. Jr.; Cadden, C.H.; Watson, R.D.; Slattery, K.T.

1998-02-01

This paper is a review of the current joining technologies for plasma facing components in the US for the International Thermonuclear Experimental Reactor (ITER) project. Many facilities are involved in this project. Many unique and innovative joining techniques are being considered in the quest to join two candidate armor plate materials (beryllium and tungsten) to a copper base alloy heat sink (CuNiBe, OD copper, CuCrZr). These techniques include brazing and diffusion bonding, compliant layers at the bond interface, and the use of diffusion barrier coatings and diffusion enhancing coatings at the bond interfaces. The development and status of these joining techniques will be detailed in this report

Chemically produced nanostructured ODS-lanthanum oxide-tungsten composites sintered by spark plasma

International Nuclear Information System (INIS)

Yar, Mazher Ahmed; Wahlberg, Sverker; Bergqvist, Hans; Salem, Hanadi G.; Johnsson, Mats; Muhammed, Mamoun

2011-01-01

High purity W and W-0.9La 2 O 3 (wt.%) nanopowders were produced by a wet chemical route. The precursor was prepared by the reaction of ammonium paratungstate (APT) with lanthanum salt in aqueous solutions. High resolution electron microscopy investigations revealed that the tungstate particles were coated with oxide precipitates. The precursor powder was reduced to tungsten metal with dispersed lanthanum oxide. Powders were consolidated by spark plasma sintering (SPS) at 1300 and 1400 o C to suppress grain growth during sintering. The final grain size relates to the SPS conditions, i.e. temperature and heating rate, regardless of the starting powder particle size. Scanning electron microscopy revealed that oxide phases were mainly accumulated at grain boundaries while the tungsten matrix constituted of nanosized sub-grains. The transmission electron microscopy revealed that the tungsten grains consist of micron-scale grains and finer sub-grains. EDX analysis confirmed the presence of W in dispersed oxide phases with varying chemical composition, which evidenced the presence of complex oxide phases (W-O-La) in the sintered metals.

Preparation of tungsten coatings on graphite by electro-deposition via Na2WO4–WO3 molten salt system

International Nuclear Information System (INIS)

Sun, Ning-bo; Zhang, Ying-chun; Jiang, Fan; Lang, Shao-ting; Xia, Min

2014-01-01

Highlights: • Tungsten coatings on graphite were firstly obtained by electro-deposition method via Na 2 WO 4 –WO 3 molten salt system. • Uniform and dense tungsten coatings could be easily prepared in each face of the sample, especially the complex components. • The obtained tungsten coatings are with high purity, ultra-low oxygen content (about 0.022 wt%). • Modulate pulse parameters can get tungsten coatings with different thickness and hardness. - Abstract: Tungsten coating on graphite substrate is one of the most promising candidate materials as the ITER plasma facing components. In this paper, tungsten coatings on graphite substrates were fabricated by electro-deposition from Na 2 WO 4 –WO 3 molten salt system at 1173 K in atmosphere. Tungsten coatings with no impurities were successfully deposited on graphite substrates under various pulsed current densities in an hour. By increasing the current density from 60 mA cm −2 to 120 mA cm −2 an increase of the average size of tungsten grains, the thickness and the hardness of tungsten coatings occurs. The average size of tungsten grains can reach 7.13 μm, the thickness of tungsten coating was in the range of 28.8–51 μm, and the hardness of coating was higher than 400 HV. No cracks or voids were observed between tungsten coating and graphite substrate. The oxygen content of tungsten coating is about 0.022 wt%

Preliminary assessment of the tritium inventory and permeation in the plasma facing components of ITER

International Nuclear Information System (INIS)

Federici, G.; Holland, D.; Brooks, J.; Causey, R.; Dolan, T.J.; Longhurst, G.

1995-01-01

This paper discusses preliminary quantitative predictions for the tritium inventory in- and permeation through the first-wall and divertor PFC's of ITER. The primary plasma facing material under consideration is beryllium, with possible use of tungsten or carbon fiber composites (CFC's) on high-heat-flux surfaces. They use state-of-the-art tritium transport models, in conjunction with design parameters, and loading conditions anticipated for the first-wall, baffle, limiter and divertor. The analysis includes the synergistic effects of erosion on tritium implantation and trapping, which are expected to play a key role, particularly in the divertor regions where the interaction of the plasma with the surfaces will be most severe. The influence of several key parameters that strongly affect tritium build-up and release is assessed. Finally, they discuss the uncertainties in materials properties under ITER operating conditions and the R and D needed to resolve these uncertainties

Sequential and simultaneous thermal and particle exposure of tungsten

International Nuclear Information System (INIS)

Steudel, I; Huber, A; Kreter, A; Linke, J; Sergienko, G; Unterberg, B; Wirtz, M

2016-01-01

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. (paper)

Overview of processing technologies for tungsten-steel composites and FGMs for fusion applications

Czech Academy of Sciences Publication Activity Database

Matějíček, Jiří; Nevrlá, Barbara; Vilémová, Monika; Boldyryeva, Hanna

2015-01-01

Roč. 60, č. 2 (2015), s. 267-273 ISSN 0029-5922. [Kudowa Summer School „Towards Fusion Energy“. Kudowa Zdrój, 09.06.2014-13.06.2014] R&D Projects: GA ČR(CZ) GAP108/12/1872 Institutional support: RVO:61389021 Keywords : plasma-facing components * functionally graded materials (FGMs), * tungsten * steel * plasma spraying * powder metallurgy Subject RIV: JK - Corrosion ; Surface Treatment of Materials Impact factor: 0.546, year: 2015 http://www.nukleonika.pl/#/?p=1222

Dynamic behavior of plasma-facing materials during plasma instabilities in tokamak reactors

International Nuclear Information System (INIS)

Hassanein, A.; Konkashbaev, I.

1997-01-01

Damage to plasma-facing and nearby components due to plasma instabilities remains a major obstacle to a successful tokamak concept. The high energy deposited on facing materials during plasma instabilities can cause severe erosion, plasma contamination, and structural failure of these components. Erosion damage can take various forms such as surface vaporization, spallation, and liquid ejection of metallic materials. Comprehensive thermodynamic and radiation hydrodynamic codes have been developed, integrated, and used to evaluate the extent of various damage to plasma-facing and nearby components. The eroded and splashed materials will be transported and then redeposited elsewhere on other plasma-facing components. Detailed physics of plasma/solid-liquid/vapor interaction in a strong magnetic field have been developed, optimized, and implemented in a self-consistent model. The plasma energy deposited in the evolving divertor debris is quickly and intensely reradiated, which may cause severe erosion and melting of other nearby components. Factors that influence and reduce vapor-shielding efficiency such as vapor diffusion and turbulence are also discussed and evaluated

Laser re-melting of tungsten damaged by transient heat loads

Czech Academy of Sciences Publication Activity Database

Loewenhoff, Th.; Linke, J.; Matějíček, Jiří; Rasinski, M.; Vostřák, M.; Wirtz, M.

2016-01-01

Roč. 9, December (2016), s. 165-170 ISSN 2352-1791. [International Conference of Fusion Reactor Material (ICFRM-17) /17./. Aachen, 11.10.2015-16.10.2015] R&D Projects: GA ČR(CZ) GA14-12837S Institutional support: RVO:61389021 Keywords : Plasma facing material * Laser surface remelting * Transient heat load * Tungsten Subject RIV: JG - Metallurgy http://dx.doi.org/10.1016/j.nme.2016.04.004

Divertor tungsten tiles erosion in the region of the castellated gaps

Energy Technology Data Exchange (ETDEWEB)

Hu, Wanpeng, E-mail: wangdez@dlut.edu.cn; Sang, Chaofeng; Sun, Zhenyue; Wang, Dezhen

2016-11-01

Highlights: • Simulation of the tungsten tiles erosion by different impurities in the divertor gap region is done by using a 2d3v Particle-In-Cell code. • High-Z impurity causes the largest erosion rate on W tile. • The peak physical sputtering erosion rate locates at the plasma-facing corners. - Abstract: Erosion of tungsten (W) is a very important issue for the future fusion device. The castellated divertor makes it more complicated due to complex geometry of the gap between the tiles. In this work, the plasma behaviors and resulting W tile erosion in the divertor tile gap region are studied by using a two dimension-in-space and three dimension-in-velocity (2d3 v) Particle-In-Cell (PIC) code. Deuterium ions (D{sup +}) and electrons are traced self-consistently in the simulation to provide the plasma background. Since there are lots of impurities, which may make a great impact on the tile erosion, in the divertor region to radiate the power, the erosion of W tile by different species are thus considered. The contributions of deuterium and impurities: Li, C, Ne, and Ar, to the W erosion, are studied under EAST conditions to show a straightforward insight. It is observed that the physical sputtering of W tile by impurities is much higher than that by the D ions, and the peak erosion region locates at the plasma-facing corners.

Lifetime evaluation of plasma-facing materials during a tokamak disruption

International Nuclear Information System (INIS)

Hassanein, A.; Konkashbaev, I.

1995-09-01

Erosion losses of plasma-facing materials in a tokamak reactor during major disruptions, giant ELMS, and large power excursions are serious concerns that influence component survivability and overall lifetime. Two different mechanisms lead to material erosion during these events: surface vaporization and loss of the melt layer. Hydrodynamics and radiation transport in the rapidly developed vapor-cloud region above the exposed area are found to control and determine the net erosion thickness from surface vaporization. A comprehensive self-consistent kinetic model has been developed in which the time-dependent optical properties and the radiation field of the vapor cloud are calculated in order to correctly estimate the radiation flux at the divertor surface. The developed melt layer of metallic divertor materials will, however, be free to move and can be eroded away due to various forces. , Physical mechanisms that affect surface vaporization and cause melt layer erosion are integrated in a comprehensive model. It is found that for metallic components such as beryllium and tungsten, lifetime due to these abnormal events will be controlled and dominated by the evolution and hydrodynamics of the melt layer during the disruption. The dependence of divertor plate lifetime on various aspects of plasma/material interaction physics is discussed

Plasma interaction with tungsten samples in the COMPASS tokamak in ohmic ELMy H-modes

International Nuclear Information System (INIS)

Dimitrova, M; Weinzettl, V; Matejicek, J; Dejarnac, R; Stöckel, J; Havlicek, J; Panek, R; Popov, Tsv; Marinov, S; Costea, S

2016-01-01

This paper reports experimental results on plasma interaction with tungsten samples with or without pre-grown He fuzz. Under the experimental conditions, arcing was observed on the fuzzy tungsten samples, resulting in localized melting of the fuzz structure that did not extend into the bulk. The parallel power flux densities were obtained from the data measured by Langmuir probes embedded in the divertor tiles on the COMPASS tokamak. Measurements of the current-voltage probe characteristics were performed during ohmic ELMy H-modes reached in deuterium plasmas at a toroidal magnetic field B T = 1.15 T, plasma current I p = 300 kA and line-averaged electron density n e = 5×10 19 m -3 . The data obtained between the ELMs were processed by the recently published first-derivative probe technique for precise determination of the plasma potential and the electron energy distribution function (EEDF). The spatial profile of the EEDF shows that at the high-field side it is Maxwellian with a temperature of 5 -- 10 eV. The electron temperatures and the ion-saturation current density obtained were used to evaluate the radial distribution of the parallel power flux density as being in the order of 0.05 -- 7 MW/m 2 . (paper)

Effects of hydrogen and helium irradiation on optical property of tungsten

International Nuclear Information System (INIS)

Kazutoshi Tokunaga; Tadashi Fujiwara; Naoaki Yoahida; Koichiro Ezato; Satoshi Suzuki; Masato Akiba

2006-01-01

Plasma-wall interactions cause surface modification, compositional and structural change on material surface due to sputtering, impurity deposition and radiation damage, etc. As a result, optical property (response of electron and lattice on material for electromagnetic wave) on surface of the plasma facing components would be changed. In particular, diagnostic components, such as metallic mirrors, mounted close to the plasma will be subjected by plasma particles such as hydrogen isotope and helium in the fusion devices. It is well recognized that decrease of optical reflectivity of the metallic mirrors due to the plasma-material interaction will be critical issues for the plasma diagnosis. In the present work, tungsten has been irradiated by hydrogen and helium beam. After that, optical reflectivity and surface modification have been measured to investigate fundamental process of optical property change due to hydrogen and helium beam irradiation. Samples used in the present experiment are powder metallurgy tungsten. Hydrogen and helium irradiations are performed in an ion beam facility at JAEA, the Particle Beam Engineering Facility (PBEF). The energy of hydrogen and helium is 19.0 and 18.7 keV, respectively. Beam duration is 1.3 - 3.5 s. The samples are irradiated up to a fluence of the orders between 10 22 and 10 24 He/m 2 by the repeated pulse irradiations of 14-450 cycles. The surface temperature is measured with an optical pyrometer. After the repeated irradiation experiments, surface modification and composition are examined with a scanning electron microscope (SEM) and a scanning probe microscope (SPM), etc. In addition, the optical reflectivity is measured in the wavelength range of 190 - 2400 nm using an ultraviolet-visible and near-infrared spectrophotometer. The reflectivity after the irradiation decreases depending on fluence and a peak temperature of the samples during the irradiation. In addition, their reflectivity spectra also change. This means

Radiative capture of slow electrons by tungsten surface

International Nuclear Information System (INIS)

Artamonov, O.M.; Belkina, G.M.; Samarin, S.N.; Yakovlev, I.I.

1987-01-01

Isochromatic spectra of radiation capture of slow electrons by the surface of mono- and polycrystal tungsten recorded on 322 and 405 nm wave lengths are presented. The effect of oxygen adsorption on isochromates of the (110) face of tungsten monocrystal is investigated. The obtained isochromatic spectra are compared with energy band structure of tungsten. Based on the analysis of the obtained experimental results it is assumed that optical transition to the final state at the energy of 7.3 eV relatively to Fermi level is conditioned by surface states of the tungsten face (110)

ITER plasma facing components

International Nuclear Information System (INIS)

Kuroda, T.; Vieider, G.; Akiba, M.

1991-01-01

This document summarizes results of the Conceptual Design Activities (1988-1990) for the International Thermonuclear Experimental Reactor (ITER) project, namely those that pertain to the plasma facing components of the reactor vessel, of which the main components are the first wall and the divertor plates. After an introduction and an executive summary, the principal functions of the plasma-facing components are delineated, i.e., (i) define the low-impurity region within which the plasma is produced, (ii) absorb the electromagnetic radiation and charged-particle flux from the plasma, and (iii) protect the blanket/shield components from the plasma. A list of critical design issues for the divertor plates and the first wall is given, followed by discussions of the divertor plate design (including the issues of material selection, erosion lifetime, design concepts, thermal and mechanical analysis, operating limits and overall lifetime, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, and advanced divertor concepts) and the first wall design (armor material and design, erosion lifetime, overall design concepts, thermal and mechanical analysis, lifetime and operating limits, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, an alternative first wall design, and the limiters used instead of the divertor plates during start-up). Refs, figs and tabs

Mixed plasma-facing materials research at INEEL

International Nuclear Information System (INIS)

Anderl, R.A.; Longhurst, G.R.; Pawelko, R.J.

2001-01-01

Mixed-materials research at the Idaho National Engineering and Environmental Laboratory (INEEL) has focused on Be-C and W-C systems. The purpose of this work was to investigate hydrogen isotope retention in these systems. Plasma-mixed material layers using carbon coated Be and W specimens that were heat-treated and tungsten carbide specimens prepared by chemical vapor deposition (CVD) were simulated. Hydrogen isotope retention was investigated by means of thermal desorption spectroscopy (TDS) measurements on deuterium implanted samples

Numerical analysis of tungsten erosion and deposition processes under a DEMO divertor plasma

Directory of Open Access Journals (Sweden)

Yuki Homma

2017-08-01

Full Text Available Erosion reduction of tungsten (W divertor target is one of the most important research subjects for the DEMO fusion reactor design, because the divertor target has to sustain large fluence of incident particles, composed mainly of fuel ions and seeded impurities, during year-long operation period. Rate of net erosion and deposition on outer divertor target has been studied by using the integrated SOL/divertor plasma code SONIC and the kinetic full-orbit impurity transport code IMPGYRO. Two background plasmas have been used: one is lower density ni and higher temperature case and the other is higher ni and lower temperature case. Net erosion has been seen in the lower ni case. But in the higher ni case, the net erosion has been almost suppressed due to increased return rate and reduced self-sputtering yield. Following two factors are important to understand the net erosion formation: (i ratio of the 1st ionization length of sputtered W atom to the Larmor gyro radius of W+ ion, (ii balance between the friction force and the thermal force exerted on W ions. DEMO divertor design should take into account these factors to prevent target erosion.

Mobility of hydrogen-helium clusters in tungsten studied by molecular dynamics

Energy Technology Data Exchange (ETDEWEB)

Grigorev, Petr, E-mail: grigorievpit@gmail.com [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, Mol, 2400 (Belgium); Ghent University, Applied Physics EA17 FUSION-DC, St.Pietersnieuwstraat, 41 B4, B-9000, Gent (Belgium); Department of Experimental Nuclear Physics K-89, Institute of Physics, Nanotechnologies, and Telecommunications, Peter the Great St.Petersburg Polytechnic University, St. Petersburg (Russian Federation); Terentyev, Dmitry; Bonny, Giovanni [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, Mol, 2400 (Belgium); Zhurkin, Evgeny E. [Department of Experimental Nuclear Physics K-89, Institute of Physics, Nanotechnologies, and Telecommunications, Peter the Great St.Petersburg Polytechnic University, St. Petersburg (Russian Federation); Oost, Guido van [Ghent University, Applied Physics EA17 FUSION-DC, St.Pietersnieuwstraat, 41 B4, B-9000, Gent (Belgium); Noterdaeme, Jean-Marie [Ghent University, Applied Physics EA17 FUSION-DC, St.Pietersnieuwstraat, 41 B4, B-9000, Gent (Belgium); Max-Planck-Institut für Plasmaphysik, Garching (Germany)

2016-06-15

Tungsten is a primary candidate material for plasma facing components in fusion reactors. Interaction of plasma components with the material is unavoidable and will lead to degradation of the performance and the lifetime of the in-vessel components. In order to gain better understanding the mechanisms driving the material degradation at atomic level, atomistic simulations are employed. In this work we study migration, stability and self-trapping properties of pure helium and mixed helium-hydrogen clusters in tungsten by means of molecular dynamics simulations. We test two versions of an embedded atom model interatomic potential by comparing it with ab initio data regarding the binding properties of He clusters. By analysing the trajectories of the clusters during molecular dynamics simulations at finite temperatures we obtain the diffusion parameters. The results show that the diffusivity of mixed clusters is significantly lower, than that of pure helium clusters. The latter suggest that the formation of mixed clusters during mixed hydrogen helium plasma exposure will affect the helium diffusivity in the material.

The detection of He in tungsten following ion implantation by laser-induced breakdown spectroscopy

Science.gov (United States)

Shaw, G.; Bannister, M.; Biewer, T. M.; Martin, M. Z.; Meyer, F.; Wirth, B. D.

2018-01-01

Laser-induced breakdown spectroscopy (LIBS) results are presented that provide depth-resolved identification of He implanted in polycrystalline tungsten (PC-W) targets by a 200 keV He+ ion beam, with a surface temperature of approximately 900 °C and a peak fluence of 1023 m-2. He retention, and the influence of He on deuterium and tritium recycling, permeation, and retention in PC-W plasma facing components are important questions for the divertor and plasma facing components in a fusion reactor, yet are difficult to quantify. The purpose of this work is to demonstrate the ability of LIBS to identify helium in tungsten; to investigate the sensitivity of laser parameters including, laser energy and gate delay, that directly influence the sensitivity and depth resolution of LIBS; and to perform a proof-of-principle experiment using LIBS to measure relative He intensities as a function of depth. The results presented demonstrate the potential not only to identify helium but also to develop a methodology to quantify gaseous impurity concentration in PC-W as a function of depth.

Material Challenges For Plasma Facing Components in Future Fusion Reactors

International Nuclear Information System (INIS)

Linke, J; Pintsuk, G.; Rödig, M.

2013-01-01

Increasing attention is directed towards thermonuclear fusion as a possible future energy source. Major advantages of this energy conversion technology are the almost inexhaustible resources and the option to produce energy without CO2-emissions. However, in the most advanced field of magnetic plasma confinement a number of technological challenges have to be met. In particular high-temperature resistant and plasma compatible materials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PFMs) and components (PFCs) in such fusion devices, i.e. the first wall (FW), the limiters and the divertor, are strongly affected by the plasma wall interaction processes and the applied intense thermal loads during plasma operation. On the one hand, these mechanisms have a strong influence on the plasma performance; on the other hand, they have major impact on the lifetime of the plasma facing armour. In present-day and next step devices the resulting thermal steady state heat loads to the first wall remain below 1 MWm-2; the limiters and the divertor are expected to be exposed to power densities being at least one order of magnitude above the FW-level, i.e. up to 20 MWm-2 for next step tokamaks such as ITER or DEMO. These requirements are responsible for high demands on the selection of qualified PFMs and heat sink materials as well as reliable fabrication processes for actively cooled plasma facing components. The technical solutions which are considered today are mainly based on the PFMs beryllium, carbon or tungsten joined to copper alloys or stainless steel heat sinks. In addition to the above mentioned quasi-stationary heat loads, short transient thermal pulses with deposited energy densities up to several tens of MJm-2 are a serious concern for next step tokamak devices. The most frequent events are so-called Edge Localized Modes (type I ELMs) and plasma disruptions

Deuterium trapping in tungsten deposition layers formed by deuterium plasma sputtering

International Nuclear Information System (INIS)

Alimov, V.Kh.; Roth, J.; Shu, W.M.; Komarov, D.A.; Isobe, K.; Yamanishi, T.

2010-01-01

A study of the influence of the deposition conditions on the surface morphology and deuterium (D) concentration in tungsten (W) deposition layers formed by magnetron sputtering and in the linear plasma generator has been carried out. Thick W layers (≥0.4 μm) deposited onto copper substrates demonstrate areas of pilling and, after post-deposition heating to 1300 K, flaking-off and fracturing. For thin W layers (≤80 nm) deposited onto stainless steel (SS) and W substrates, no areas of flaking-off and fracturing exist both after deposition and after post-deposition heating to 673 K for the SS substrate and to 1300 K for the W substrate. The concentration of deuterium in the W layers was found to decrease with increasing substrate temperature and with increasing tungsten deposition rate. For layers with relatively high concentration of oxygen (0.20-0.60 O/W), a decrease of the D concentration with increasing substrate temperature is more pronounced than that for layers deposited in good vacuum conditions. To describe the evolution of the D/W ratio with the substrate temperature and the tungsten deposition rate, an empirical equation proposed by De Temmerman and Doerner [J. Nucl. Mater. 389 (2009) 479] but with alternative parameters has been used.

Effect of rhenium addition on tungsten fuzz formation in helium plasmas

Energy Technology Data Exchange (ETDEWEB)

Khan, Aneeqa, E-mail: aneeqa.khan-3@postgrad.manchester.ac.uk [School of Mechanical, Aerospace and Civil Engineering, The University of Manchester, M13 9PL (United Kingdom); De Temmerman, Gregory [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046 - 13067 St Paul Lez Durance Cedex (France); Morgan, Thomas W. [FOM Institute DIFFER – Dutch Institute for Fundamental Energy Research, Partner in the Trilateral Euregio Cluster, Eindhoven (Netherlands); Ward, Michael B. [Institute for Materials Research, School of Chemical Process Engineering, University of Leeds, Leeds, LS2 9JT (United Kingdom)

2016-06-15

The effect of the addition of rhenium to tungsten on the formation of a nanostructure referred to as ‘fuzz’ when exposed to helium plasmas at fusion relevant ion fluxes was investigated in the Magnum and Pilot PSI devices at the FOM Institute DIFFER. The effect rhenium had on fuzz growth was seen to be dependent on time, temperature and flux. Initial fuzz growth was seen to be highly dependent on grain orientation, with rhenium having little effect. Once the fuzz was fully developed, the effect of grain orientation disappeared and the rhenium had an inhibiting effect on growth. This could be beneficial for inhibiting fuzz growth in a future fusion reactor, where transmutation of tungsten to rhenium is expected. It also appears that erosion or annealing of the fuzz is limiting growth of fuzz at higher temperatures in the range of ∼1340 °C.

Effect of rhenium addition on tungsten fuzz formation in helium plasmas

International Nuclear Information System (INIS)

Khan, Aneeqa; De Temmerman, Gregory; Morgan, Thomas W.; Ward, Michael B.

2016-01-01

The effect of the addition of rhenium to tungsten on the formation of a nanostructure referred to as ‘fuzz’ when exposed to helium plasmas at fusion relevant ion fluxes was investigated in the Magnum and Pilot PSI devices at the FOM Institute DIFFER. The effect rhenium had on fuzz growth was seen to be dependent on time, temperature and flux. Initial fuzz growth was seen to be highly dependent on grain orientation, with rhenium having little effect. Once the fuzz was fully developed, the effect of grain orientation disappeared and the rhenium had an inhibiting effect on growth. This could be beneficial for inhibiting fuzz growth in a future fusion reactor, where transmutation of tungsten to rhenium is expected. It also appears that erosion or annealing of the fuzz is limiting growth of fuzz at higher temperatures in the range of ∼1340 °C.

Electron work function of stepped tungsten surfaces

International Nuclear Information System (INIS)

Krahl-Urban, B.

1976-03-01

The electron work function of tungsten (110) vicinal faces was measured with the aid of thermionic emission, and its dependence on the crystallographic orientation and the surface structure was investigated. The thermionic measurements were evaluated with the aid of the Richardson plot. The real temperature of the emitting tungsten faces was determined with an accuracy of +- 0.5% in the range between 2,200 and 2,800 K. The vicinal faces under investigation have been prepared with an orientation exactness of +- 15'. In the tungsten (110) vicinal faces under investigation, a strong dependence of the temperature coefficient d PHI/dT of the work function on the crystallographic orientation was found. A strong influence of the edge structure as well as of the step density on the temperature coefficient was observed. (orig./HPOE) [de

Influence of the base temperature on the performance of tungsten under thermal and particle exposure

Directory of Open Access Journals (Sweden)

I. Steudel

2017-08-01

Full Text Available Tungsten, the plasma facing material (PFM for the divertor in ITER, must sustain severe, distinct loading conditions. This broad array of exposure conditions necessitates comprehensive experiments that cover most of the expected loading parameters to predict qualitative statements about the performance and as a consequence thereof the intended operation time. However, comprehensive experiments are inherently difficult to realize due to the fact that there is no device that is capable of simulating all loading conditions simultaneously. Nevertheless, the linear plasma device PSI-2 enables experiments combining thermal and particle exposure at the same time. In this work, sequential and simultaneous loads on pure tungsten at different base temperatures were investigated to study not only the performance of the material, but also the influence of the experimental parameters. The detailed analysis and comparison of the obtained results showed different kinds of damage depending on the loading sequence, power density, microstructure of the samples, and base temperature. Finally, samples with transversal grain orientation (T showed the weakest damage resistance and the increase of the base temperature could not compensate the detrimental impact of deuterium.

Dynamic control of low-Z material deposition and tungsten erosion by strike point sweeping on DIII-D

Directory of Open Access Journals (Sweden)

J. Guterl

2017-08-01

Full Text Available Carbon deposition on tungsten between ELMs was investigated in DIII-D in semi-attached/detached H-mode plasma conditions using fixed outer strike point (OSP positions. Carbon deposition during plasma exposure of tungsten was monitored in-situ by measuring the reflectivity of the tungsten sample surface. No significant carbon deposition, i.e., without strong variations of the reflectivity, was observed during these experiments including discharges at high densities. In contrast, ERO modeling predicts a significant carbon deposition on the tungsten surface for those high density plasma conditions. The surface reflectivity decreases with methane injection, consistent with increased carbon coverage, as expected. The sweeping of OSP leads to a pronounced increase of the surface reflectivity, suggesting that the strike point sweeping may provide an effective means to remove carbon coating from tungsten surface. The ERO modeling however predicts again a regime of carbon deposition for these experiments. The discrepancies between carbon deposition regime predicted by the ERO model and the experimental observations suggest that carbon erosion during ELMs may significantly affect carbon deposition on tungsten.

Cracks and nanodroplets produced on tungsten surface samples by dense plasma jets

Science.gov (United States)

Ticoş, C. M.; Galaţanu, M.; Galaţanu, A.; Luculescu, C.; Scurtu, A.; Udrea, N.; Ticoş, D.; Dumitru, M.

2018-03-01

Small samples of 12.5 mm in diameter made from pure tungsten were exposed to a dense plasma jet produced by a coaxial plasma gun operated at 2 kJ. The surface of the samples was analyzed using a scanning electron microscope (SEM) before and after applying consecutive plasma shots. Cracks and craters were produced in the surface due to surface tensions during plasma heating. Nanodroplets and micron size droplets could be observed on the samples surface. An energy-dispersive spectroscopy (EDS) analysis revealed that the composition of these droplets coincided with that of the gun electrode material. Four types of samples were prepared by spark plasma sintering from powders with the average particle size ranging from 70 nanometers up to 80 μm. The plasma power load to the sample surface was estimated to be ≈4.7 MJ m-2 s-1/2 per shot. The electron temperature and density in the plasma jet had peak values 17 eV and 1.6 × 1022 m-3, respectively.

Challenges and opportunities of modeling plasma–surface interactions in tungsten using high-performance computing

Energy Technology Data Exchange (ETDEWEB)

Wirth, Brian D., E-mail: bdwirth@utk.edu [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Nuclear Science and Engineering Directorate, Oak Ridge National Laboratory, Oak Ridge, TN (United States); Hammond, K.D. [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Krasheninnikov, S.I. [University of California, San Diego, La Jolla, CA (United States); Maroudas, D. [University of Massachusetts, Amherst, Amherst, MA 01003 (United States)

2015-08-15

The performance of plasma facing components (PFCs) is critical for ITER and future magnetic fusion reactors. The ITER divertor will be tungsten, which is the primary candidate material for future reactors. Recent experiments involving tungsten exposure to low-energy helium plasmas reveal significant surface modification, including the growth of nanometer-scale tendrils of “fuzz” and formation of nanometer-sized bubbles in the near-surface region. The large span of spatial and temporal scales governing plasma surface interactions are among the challenges to modeling divertor performance. Fortunately, recent innovations in computational modeling, increasingly powerful high-performance computers, and improved experimental characterization tools provide a path toward self-consistent, experimentally validated models of PFC and divertor performance. Recent advances in understanding tungsten–helium interactions are reviewed, including such processes as helium clustering, which serve as nuclei for gas bubbles; and trap mutation, dislocation loop punching and bubble bursting; which together initiate surface morphological modification.

Challenges and opportunities of modeling plasma–surface interactions in tungsten using high-performance computing

International Nuclear Information System (INIS)

Wirth, Brian D.; Hammond, K.D.; Krasheninnikov, S.I.; Maroudas, D.

2015-01-01

The performance of plasma facing components (PFCs) is critical for ITER and future magnetic fusion reactors. The ITER divertor will be tungsten, which is the primary candidate material for future reactors. Recent experiments involving tungsten exposure to low-energy helium plasmas reveal significant surface modification, including the growth of nanometer-scale tendrils of “fuzz” and formation of nanometer-sized bubbles in the near-surface region. The large span of spatial and temporal scales governing plasma surface interactions are among the challenges to modeling divertor performance. Fortunately, recent innovations in computational modeling, increasingly powerful high-performance computers, and improved experimental characterization tools provide a path toward self-consistent, experimentally validated models of PFC and divertor performance. Recent advances in understanding tungsten–helium interactions are reviewed, including such processes as helium clustering, which serve as nuclei for gas bubbles; and trap mutation, dislocation loop punching and bubble bursting; which together initiate surface morphological modification

Prediction for disruption erosion of ITER plasma facing components; a comparison of experimental and numerical results

International Nuclear Information System (INIS)

Laan, J.G. van der; Akiba, M.; Seki, M.; Hassanein, A.; Tanchuk, V.

1991-01-01

An evaluation is given for the prediction for disruption erosion in the International Thermonuclear Engineering Reactor (ITER). At first, a description is given of the relation between plasma operating paramters and system dimensions to the predictions of loading parameters of Plasma Facing Components (PFC) in off-normal events. Numerical results from ITER parties on the prediction of disruption erosion are compared for a few typical cases and discussed. Apart from some differences in the codes, the observed discrepancies can be ascribed to different input data of material properties and boundary conditions. Some physical models for vapour shielding and their effects on numerical results are mentioned. Experimental results from ITER parties, obtained with electron and laser beams, are also compared. Erosion rates for the candidate ITER PFC materials are shown to depend very strongly on the energy deposition parameters, which are based on plasma physics considerations, and on the assumed material loss mechanisms. Lifetimes estimates for divertor plate and first wall armour are given for carbon, tungsten and beryllium, based on the erosion in the thermal quench phase. (orig.)

Simulations with current constraints of ELM- induced tungsten melt motion in ASDEX Upgrade.

Czech Academy of Sciences Publication Activity Database

Thorén, E.; Bazylev, B.; Ratynskaia, S.; Tolias, P.; Krieger, K.; Pitts, R.A.; Pestchanyi, S.; Komm, Michael; Sieglin, B.

T170, December (2017), č. článku 014006. ISSN 0031-8949. [PFMC 2017: 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications. Düsseldorf, 16.05.2017-19.05.2017] EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : thermionic * MEMOS * AUG * tungsten * melting Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 1.3 Physical sciences Impact factor: 1.280, year: 2016 http://iopscience.iop.org/article/10.1088/1402-4896/aa8855/meta

The electronic work function of the different faces of tungsten

International Nuclear Information System (INIS)

Modinos, A.

1978-01-01

A semi-empirical theory of the electronic work function of the different faces of tungsten is presented. All the parameters entering the theory, except one, are estimated independently. The one adjustable parameter relates to the isotropic contribution to the work function, and, can, in principle, be determined from a self-consistent calculation of the band-structure of the energy levels in the bulk of the metal. The calculated values for the work function are in reasonably good agreement with available experimental data for practically all of the crystallographic planes with the exception of the (100) plane. For the latter, the calculated value is 0.3 eV above the experimental value. It is suggested that a negative contribution to the surface dipole potential from surface states, that exist on this plane, may be the reason of this discrepancy. (Auth.)

Electrodeposition of metallic tungsten coating from binary oxide molten salt on low activation steel substrate

International Nuclear Information System (INIS)

Liu, Y.H.; Zhang, Y.C.; Jiang, F.; Fu, B.J.; Sun, N.B.

2013-01-01

Tungsten is considered a promising plasma facing armor material for future fusion devices. An electrodeposited metallic tungsten coating from Na 2 WO 4 –WO 3 binary oxide molten salt on low activation steel (LAS) substrate was investigated in this paper. Tungsten coatings were deposited under various pulsed currents conditions at 1173 K in atmosphere. Cathodic current density and pulsed duty cycle were investigated for pulsed current electrolysis. The crystal structure and microstructure of tungsten coatings were characterized by X-ray diffractometry, scanning electron microscopy, and energy X-ray dispersive analysis techniques. The results indicated that pulsed current density and duty cycle significantly influence tungsten nucleation and electro-crystallization phenomena. The average grain size of the coating becomes much larger with increasing cathodic current density, which demonstrates that appropriate high cathodic current density can accelerate the growth of grains on the surface of the substrate. The micro-hardness of tungsten coatings increases with the increasing thickness of coatings; the maximum micro-hardness is 482 HV. The prepared tungsten coatings have a smooth surface, a porosity of less than 1%, and an oxygen content of 0.024 wt%

Electrodeposition of metallic tungsten coating from binary oxide molten salt on low activation steel substrate

Energy Technology Data Exchange (ETDEWEB)

Liu, Y. H. [School of Materials Science and Engineering, University of Science and Technology BeiJing, Beijing (China); State Nuclear Power Research Institute, Xicheng District, Beijing (China); Zhang, Y.C., E-mail: zycustb@163.com [School of Materials Science and Engineering, University of Science and Technology BeiJing, Beijing (China); Jiang, F.; Fu, B. J.; Sun, N. B. [School of Materials Science and Engineering, University of Science and Technology BeiJing, Beijing (China)

2013-11-15

Tungsten is considered a promising plasma facing armor material for future fusion devices. An electrodeposited metallic tungsten coating from Na{sub 2}WO{sub 4}–WO{sub 3} binary oxide molten salt on low activation steel (LAS) substrate was investigated in this paper. Tungsten coatings were deposited under various pulsed currents conditions at 1173 K in atmosphere. Cathodic current density and pulsed duty cycle were investigated for pulsed current electrolysis. The crystal structure and microstructure of tungsten coatings were characterized by X-ray diffractometry, scanning electron microscopy, and energy X-ray dispersive analysis techniques. The results indicated that pulsed current density and duty cycle significantly influence tungsten nucleation and electro-crystallization phenomena. The average grain size of the coating becomes much larger with increasing cathodic current density, which demonstrates that appropriate high cathodic current density can accelerate the growth of grains on the surface of the substrate. The micro-hardness of tungsten coatings increases with the increasing thickness of coatings; the maximum micro-hardness is 482 HV. The prepared tungsten coatings have a smooth surface, a porosity of less than 1%, and an oxygen content of 0.024 wt%.

Re-emission and thermal desorption of deuterium from plasma sprayed tungsten coatings for application in ASDEX-upgrade

International Nuclear Information System (INIS)

Garcia-Rosales, C.; Franzen, P.; Plank, H.; Roth, J.; Gauthier, E.

1996-01-01

The trapping and release of deuterium implanted with an energy of 100 eV in wrought and in plasma sprayed tungsten of different manufacture and structure has been investigated by means of re-emission as well as thermal and isothermal desorption spectroscopy. The experimental data for wrought tungsten are compared with model calculations with the PIDAT code in order to estimate the parameters governing diffusion, surface recombination and trapping in tungsten. The amount of retained deuterium in tungsten is of the same order of magnitude as in graphite for the implantation parameters used in this work. The mobile hydrogen concentration in tungsten during the implantation is of the same order of magnitude than the trapped one, being released after the termination of the implantation. The fraction of deuterium trapped to defects increases strongly with the porosity of the samples. The temperature needed for the release of the trapped deuterium (∝600 K) are considerably lower than for graphite, due to the smaller trapping energy (≤1.5 eV). (orig.)

A new vision of plasma facing components

Energy Technology Data Exchange (ETDEWEB)

Nygren, Richard E., E-mail: renygre@sandia.gov [Sandia National Laboratories, Albuquerque, NM (United States); Youchison, Dennis L. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Wirth, Brian D. [University of Tennessee, Knoxville, TN (United States); Snead, Lance L.

2016-11-01

Highlights: • New approach recommended to develop refractory fusion plasma facing components. • Need to develop engineered materials architecture with nano-features. • Need to develop PFCs with gas jet cooling with very fine scale for jet arrays. • Emphasis on role of additive manufacturing as needed method for fabrication. - Abstract: This paper advances a vision for plasma facing components (PFCs) that includes the following points. The solution for plasma facing materials likely consists of engineered structures in which the layer of plasma facing material (PFM) is integrated with an engineered structure that cools the PFM and may also transition with graded composition. The key to achieving this PFC architecture will likely lie in advanced manufacturing methods, e.g., additive manufacturing, that can produce layers with controlled porosity and features such as micro-fibers and/or nano-particles that can collect He and transmutation products, limit tritium retention, and do all this in a way that maintains adequate robustness for a satisfactory lifetime. This vision has significant implications for how we structure a development program.

Response of fuzzy tungsten surfaces to pulsed plasma bombardment

International Nuclear Information System (INIS)

Nishijima, D.; Doerner, R.P.; Iwamoto, D.; Kikuchi, Y.; Miyamoto, M.; Nagata, M.; Sakuma, I.; Shoda, K.; Ueda, Y.

2013-01-01

Damage of fuzzy tungsten surfaces due to a transient plasma load is characterized in terms of mass loss, surface morphology, and optical properties. A single D pulsed (∼0.1–0.2 ms) plasma shot with surface absorbed energy density of ∼1.1 MJ m −2 leads to a mass loss of ∼80 μg, which cannot be explained by physical sputtering. Thus, macroscopic erosion processes such as droplets and dust release as well as arcing are thought to be responsible for the mass loss. In fact, scanning electron microscopy observations reveal the melting of the tips of fuzz and arc tracks. The optical reflectivity of the damaged (melted) surface is measured to be higher than that of an undamaged fuzzy surface (below ∼0.01%). Spectroscopic ellipsometry shows that the refractive index, n, and extinction coefficient, k, increase from n ≈ 1 and k ≈ 0 for an undamaged fuzzy surface with an increase in the degree of damage of fuzz

Laser Remelting of Plasma-Sprayed Tungsten Coatings

Czech Academy of Sciences Publication Activity Database

Matějíček, Jiří; Holub, P.

2014-01-01

Roč. 23, č. 4 (2014), s. 750-754 ISSN 1059-9630 R&D Projects: GA ČR(CZ) GAP108/12/1872 Grant - others:European Project ExtreMat(XE) NMP-CT-2004-500253 Institutional support: RVO:61389021 Keywords : functionally graded coatings * laser remelting * plasma facing materials * thermal conductivity * water stabilized plasma Subject RIV: JK - Corrosion ; Surface Treatment of Materials Impact factor: 1.344, year: 2014 http://link.springer.com/article/10.1007%2Fs11666-014-0067-4

Measuring deuterium permeation through tungsten near room temperature under plasma loading using a getter layer and ion-beam based detection

Directory of Open Access Journals (Sweden)

Stefan Kapser

2017-08-01

Full Text Available A method to measure deuterium permeation through tungsten near room temperature under plasma loading is presented. The permeating deuterium is accumulated in a getter layer of zirconium, titanium or erbium, respectively, on the unexposed side of the sample. Subsequently, the amount of deuterium in the getter is measured ex-situ using nuclear reaction analysis. A cover layer system on the getter prevents direct loading of the getter with deuterium from the gas phase during plasma loading. In addition, it enables the distinction of deuterium in the getter and at the cover surface. The method appears promising to add additional permeation measurement capabilities to deuterium retention experiments, also in other plasma devices, without the need for a complex in-situ permeation measurement setup.

The Design and Use of Tungsten Coated TZM Molybdenum Tile Inserts in the DIII-D Tokamak Divertor

Energy Technology Data Exchange (ETDEWEB)

Murphy, Christopher [General Atomics, San Diego; Nygren, R. E. [Sandia National Laboratories (SNL); Chrobak, C P. [General Atomics, San Diego; Buchenauer, Dean [Sandia National Laboratories (SNL); Holtrop, Kurt [General Atomics, San Diego; Unterberg, Ezekial A. [ORNL; Zach, Mike P. [ORNL

2017-08-01

Future tokamak devices are envisioned to utilize a high-Z metal divertor with tungsten as theleading candidate. However, tokamak experiments with tungsten divertors have seen significantdetrimental effects on plasma performance. The DIII-D tokamak presently has carbon as theplasma facing surface but to study the effect of tungsten on the plasma and its migration aroundthe vessel, two toroidal rows of carbon tiles in the divertor region were modified with high-Zmetal inserts, composed of a molybdenum alloy (TZM) coated with tungsten. A dedicated twoweek experimental campaign was run with the high-Z metal inserts. One row was coated withtungsten containing naturally occurring levels of isotopes. The second row was coated withtungsten where the isotope 182W was enhanced from the natural level of 26% up to greater than90%. The different isotopic concentrations enabled the experiment to differentiate between thetwo different sources of metal migration from the divertor. Various coating methods wereexplored for the deposition of the tungsten coating, including chemical vapor deposition,electroplating, vacuum plasma spray, and electron beam physical vapor deposition. The coatingswere tested to see if they were robust enough to act as a divertor target for the experiment. Testsincluded cyclic thermal heating using a high power laser and high-fluence deuterium plasmabombardment. The issues associate with the design of the inserts (tile installation, thermal stress,arcing, leading edges, surface preparation, etc.), are reviewed. The results of the tests used toselect the coating method and preliminary experimental observations are presented.

Deuterium-induced nanostructure formation on tungsten exposed to high-flux plasma

Energy Technology Data Exchange (ETDEWEB)

Xu, H.Y., E-mail: donaxu@163.com [Key Laboratory of Advanced Materials, Department of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); Science and Technology on Surface Physics and Chemistry Laboratory, Mianyang, Sichuan 621907 (China); De Temmerman, G. [FOM Institute DIFFER, Dutch Institute For Fundamental Energy Research, Ass. EURATOM-FOM, Trilateral Euregio Cluster, Postbus 1207, 3430BE Nieuwegein (Netherlands); ITER Organization, Route de Vinon-sur-Verdon CS 90046-13067, St Paul Lez Durance Cedex (France); Luo, G.-N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Jia, Y.Z.; Yuan, Y.; Fu, B.Q.; Godfrey, A. [Key Laboratory of Advanced Materials, Department of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China); Liu, W., E-mail: liuw@mail.tsinghua.edu.cn [Key Laboratory of Advanced Materials, Department of Materials Science and Engineering, Tsinghua University, Beijing 100084 (China)

2015-08-15

Surface topography of polycrystalline tungsten (W) have been examined after exposure to a low-energy (38 eV/D), high-flux (∼1.1–1.5 × 10{sup 24} m{sup −2} s{sup −1}) deuterium plasma in the Pilot-PSI linear plasma device. The methods used were scanning electron microscopy (SEM), transmission electron microscopy (TEM), positron annihilation Doppler broadening (PADB) and grazing incident X-ray diffraction (GI-XRD). After exposure to high flux D plasma, blisters and nanostructures are formed on the W surface. Generation of defects was evidenced by PADB, while high stress and mixture of phases were detected in depth of 50 nm by GI-XRD. TEM observation revealed fluctuations and disordered microstructure on the outmost surface layer. Based on these results, surface reconstruction is considered as a possible mechanism for the formation of defects and nanostructures.

Effect of neutron irradiation on the microstructure of tungsten

Directory of Open Access Journals (Sweden)

M. Klimenkov

2016-12-01

Full Text Available Two grades of pure tungsten, single and polycrystalline, were irradiated for 282 days in the HFR reactor, Petten, at 900 °C to an average damage level of 1.6dpa. Each grade of tungsten was investigated using the transmission electron microscope (TEM to assess the effect of neutron irradiation on tungsten microstructure. Investigations revealed the formation of faceted cavities, whose diameter varies from 4 to 14nm in both materials. The cavities are homogeneously distributed only inside single crystalline tungsten. The local distribution of cavities in polycrystalline tungsten is strongly influenced by grain boundaries. The number densities of cavities were measured to be 4×1021 m−3 for polycrystalline and 2.5×1021 m−3 for single crystalline tungsten. This corresponds to volumetric densities of 0.45% and 0.33% respectively. High-resolution transmission electron microscopy (HRTEM revealed that faces of cavities are oriented in (110 plane. Analytical investigations showed precipitation of rhenium and osmium produced by a transmutation reaction around cavities and at grain boundaries.

Technologies for ITER divertor vertical target plasma facing components

International Nuclear Information System (INIS)

Schlosser, J.; Escourbiac, F.; Merola, M.; Fouquet, S.; Bayetti, P.; Cordier, J.J.; Grosman, A.; Missirlian, M.; Tivey, R.; Roedig, M.

2005-01-01

The ITER divertor vertical target has to sustain heat fluxes up to 20 MW m -2 . The concept developed for this plasma facing component working at steady state is based on carbon fibre composite armour for the lower straight part and tungsten for the curved upper part. The main challenges involved in the use of such components include the removal of the high heat fluxes deposited and mechanically and thermally joining the armour to the metallic heat sink, despite the mismatch in the thermal expansions. Two solutions based on the use of a CuCrZr hardened copper alloy and an active metal casting (AMC (registered) ) process were investigated during the ITER EDA phase: the first one called 'flat tile geometry' was mainly developed for the Tore Supra pumped limiter, the second one called 'monoblock geometry' was developed by the EU Participating Team for the ITER project. This paper presents a review of these two solutions and analyses their assets and drawbacks: pressure drop, critical heat flux, surface temperature and expected behaviour during operation, risks during the manufacture, control of the armour defects during the manufacture and at the reception, and the possibility of repairing defective tiles

Structure and property evaluation of a vacuum plasma sprayed nanostructured tungsten-hafnium carbide bulk composite

NARCIS (Netherlands)

Rea, K. E.; Viswanathan, V.; Kruize, A.; De Hosson, J. Th. M.; O'Dell, S.; McKechnie, T.; Rajagopalan, S.; Vaidyanathan, R.; Seal, S.; O’Dell, S.

2008-01-01

Vacuum plasma spray (VPS) forming of tungsten-based metal matrix nanocomposites (MMCs) has shown to be a cost effective and time saving method for the formation of bulk monolithic nanostructured then no-mechanical components. Spray drying of powder feedstock appears to have a significant effect on

Increase in the energy absorption of pulsed plasma by the formation of tungsten nanostructure

Science.gov (United States)

Sato, D.; Ohno, N.; Domon, F.; Kajita, S.; Kikuchi, Y.; Sakuma, I.

2017-06-01

The synergistic effects of steady-state and pulsed plasma irradiation to material have been investigated in the device NAGDIS-PG (NAGoya DIvertor Simulator with Plasma Gun). The duration of the pulsed plasma was ~0.25 ms. To investigate the pulsed plasma heat load on the materials, we developed a temperature measurement system using radiation from the sample in a high time resolution. The heat deposited in response to the transient plasma on a tungsten surface was revealed by using this system. When the nanostructures were formed by helium plasma irradiation, the temperature increase on the bulk sample was enhanced. The result suggested that the amount of absorbed energy on the surface was increased by the formation of nanostructures. The possible mechanisms causing the phenomena are discussed with the calculation of a sample temperature in response to the transient heat load.

Deuterium trapping at vacancy clusters in electron/neutron-irradiated tungsten studied by positron annihilation spectroscopy

Science.gov (United States)

Toyama, T.; Ami, K.; Inoue, K.; Nagai, Y.; Sato, K.; Xu, Q.; Hatano, Y.

2018-02-01

Deuterium trapping at irradiation-induced defects in tungsten, a candidate material for plasma facing components in fusion reactors, was revealed by positron annihilation spectroscopy. Pure tungsten was electron-irradiated (8.5 MeV at ∼373 K and to a dose of ∼1 × 10-3 dpa) or neutron-irradiated (at 573 K to a dose of ∼0.3 dpa), followed by post-irradiation annealing at 573 K for 100 h in deuterium gas of ∼0.1 MPa. In both cases of electron- or neutron-irradiation, vacancy clusters were found by positron lifetime measurements. In addition, positron annihilation with deuterium electrons was demonstrated by coincidence Doppler broadening measurements, directly indicating deuterium trapping at vacancy-type defects. This is expected to cause significant increase in deuterium retention in irradiated-tungsten.

Overview of the US-Japan collaborative investigation on hydrogen isotope retention in neutron-irradiated and ion-damaged tungsten

Energy Technology Data Exchange (ETDEWEB)

Shimada, Masashi, E-mail: Masashi.Shimada@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Hatano, Y. [Hydrogen Isotope Research Center, University of Toyama, Toyama (Japan); Oya, Y. [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka (Japan); Oda, T. [Department of Nuclear Engineering and Management, The University of Tokyo, Tokyo (Japan); Hara, M. [Hydrogen Isotope Research Center, University of Toyama, Toyama (Japan); Cao, G. [Department of Engineering Physics, University of Wisconsin-Madison, Madison, WI (United States); Kobayashi, M. [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka (Japan); Sokolov, M. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Watanabe, H. [Research Institute for Applied Mechanics, Kyushu University, Fukuoka (Japan); Tyburska-Pueschel, B. [Department of Engineering Physics, University of Wisconsin-Madison, Madison, WI (United States); Institute fuer Plasmaphysik, EURATOM Association, Garching (Germany); Ueda, Y. [Graduate School of Engineering, Osaka University, Osaka (Japan); Calderoni, P. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Okuno, K. [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka (Japan)

2012-08-15

The effect of neutron-irradiation damage has been mainly simulated using high-energy ion bombardment. A recent MIT report (PSFC/RR-10-4, An assessment of the current data affecting tritium retention and its use to project towards T retention in ITER, Lipschultz et al., 2010) summarizes the observations from high-energy ion bombardment studies and illustrates the saturation trend in deuterium concentration due to damage from ion irradiation in tungsten and molybdenum above 1 displacement per atom (dpa). While this prior database of results is quite valuable for understanding the behavior of hydrogen isotopes in plasma facing components (PFCs), it does not encompass the full range of effects that must be considered in a practical fusion environment due to short penetration depth, damage gradient, high damage rate, and high primary knock-on atom (PKA) energy spectrum of the ion bombardment. In addition, neutrons change the elemental composition via transmutations, and create a high radiation environment inside PFCs, which influences the behavior of hydrogen isotope in PFCs, suggesting the utilization of fission reactors is necessary for neutron-irradiation. Under the framework of the US-Japan TITAN program, tungsten samples (99.99 at.% purity from A.L.M.T. Co.) were irradiated by fission neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory (ORNL), at 50 and 300 Degree-Sign C to 0.025, 0.3, and 2.4 dpa, and the investigation of deuterium retention in neutron-irradiated tungsten was performed in the Tritium Plasma Experiment (TPE), the unique high-flux linear plasma facility that can handle tritium, beryllium and activated materials. This paper reports the recent results from the comparison of ion-damaged tungsten via various ion species (2.8 MeV Fe{sup 2+}, 20 MeV W{sup 2+}, and 700 keV H{sup -}) with that from neutron-irradiated tungsten to identify the similarities and differences among them.

On helium cluster dynamics in tungsten plasma facing components of fusion devices

International Nuclear Information System (INIS)

Krasheninnikov, S.I.; Faney, T.; Wirth, B.D.

2014-01-01

This paper describes the dynamics of helium clustering behaviour within either a nanometer-sized tendril of fuzz, or a half-space domain, as predicted by a reaction–diffusion model. This analysis has identified a dimensionless parameter, P Δ , which is a balance of the reaction and diffusion actions of insoluble He in a metal matrix and which governs the self-trapping effects of He into growing bubbles within a tendril. The impact of He self-trapping, as well as trapping caused by pre-existing traps in the form of lattice defects or clusters of impurities, within a half-space domain results in the formation of a densely packed layer of nanometer-sized bubbles with high number density. This prediction is consistent with available experimental observations in which a dense zone of helium bubbles is observed in tungsten, which are compared to estimates of the layer characteristics. Direct numerical simulation of the reaction–diffusion cluster dynamics supports the analysis presented here. (paper)

Evaluation of surface, microstructure and phase modifications on various tungsten grades induced by pulsed plasma loading

Czech Academy of Sciences Publication Activity Database

Vilémová, Monika; Pala, Zdeněk; Jäger, Aleš; Matějíček, Jiří; Chernyshova, M.; Kowalska-Strzęciwilk, E.; Tonarová, Dana; Gribkov, V. A.

2016-01-01

Roč. 91, č. 3 (2016), č. článku 034003. ISSN 0031-8949. [PLASMA 2015 : International Conference on Research and Applications of Plasmas. Warsaw, 07.09.2015-11.09.2015] R&D Projects: GA ČR(CZ) GA14-12837S Institutional support: RVO:61389021 ; RVO:68378271 Keywords : tungsten * titanium carbide * yttrium oxide * plasma focus * damage Subject RIV: BL - Plasma and Gas Discharge Physics; BL - Plasma and Gas Discharge Physics (FZU-D) OBOR OECD: 1.3 Physical sciences; 1.3 Physical sciences (FZU-D) Impact factor: 1.280, year: 2016 http://iopscience.iop.org/article/10.1088/0031-8949/91/3/034003/meta

Plasma exposure behavior of re-deposited tungsten on structural materials of fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Xu, Yu-Ping; Wang, Jing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Science Island Branch of Graduate School, University of Science & Technology of China, Hefei 230031 (China); Zhou, Hai-Shan, E-mail: haishanzhou@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Liu, Feng [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Li, Zeng-De [General Research Institute for Nonferrous Metals, Beijing 100088 (China); Li, Xiao-Chun; Lu, Tao [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Liu, Hao-Dong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Science Island Branch of Graduate School, University of Science & Technology of China, Hefei 230031 (China); Ding, Fang; Mao, Hong-Min; Zhao, Ming-Zhong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Lin, Chen-Guang [General Research Institute for Nonferrous Metals, Beijing 100088 (China); Luo, Guang-Nan [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Science Island Branch of Graduate School, University of Science & Technology of China, Hefei 230031 (China); Hefei Center for Physical Science and Technology, Hefei 230031 (China); Hefei Science Center of Chinese Academy of Science, Hefei 230027 (China)

2017-05-15

To evaluate the effects of re-deposited tungsten (W) on the surface modification and hydrogen isotope retention behavior of fusion structural materials, the plasma exposure behavior of re-deposited W samples prepared by magnetron sputtering on the F82H steel, the V-5Cr-5Ti alloy as well as bare substrate samples was investigated. All the samples were exposed to 367 shots of deuterium plasmas in the 2015 spring EAST campaign. After the plasma exposure, large area of W layer was exfoliated, while big blisters were found at the interface between the remaining W layer and the substrate materials. The deuterium retention behavior of the samples with re-deposited W layer was characterized by thermal desorption spectroscopy and compared with the bare substrate samples.

Tungsten/copper composite deposits produced by a cold spray

International Nuclear Information System (INIS)

Kang, Hyun-Ki; Kang, Suk Bong

2003-01-01

An agglomerated tungsten/copper composite powder was both cold sprayed and plasma sprayed onto a mild steel substrate for electronic package applications. Most pores resulting from the spraying were found in the vicinity of the tungsten-rich regions of the final product. The levels of porosity varied with the amount of tungsten present. No copper oxidation was found at the cold-sprayed deposit, but relatively high copper oxidation was observed at the plasma-sprayed deposit

Preparation of tungsten coatings on graphite by electro-deposition via Na{sub 2}WO{sub 4}–WO{sub 3} molten salt system

Energy Technology Data Exchange (ETDEWEB)

Sun, Ning-bo [School of Materials Science and Engineering, University of Science and Technology Beijing, 30 Xueyuan Road, Haidian District, Beijing 100083 (China); Zhang, Ying-chun, E-mail: zycustb@163.com [School of Materials Science and Engineering, University of Science and Technology Beijing, 30 Xueyuan Road, Haidian District, Beijing 100083 (China); Jiang, Fan; Lang, Shao-ting [School of Materials Science and Engineering, University of Science and Technology Beijing, 30 Xueyuan Road, Haidian District, Beijing 100083 (China); Xia, Min [School of Materials Science and Engineering, University of Science and Technology Beijing, 30 Xueyuan Road, Haidian District, Beijing 100083 (China); Institute of Powder Metallurgy and Advanced Ceramics, Southwest Jiaotong University, 111, 1st Section, Northern 2nd Ring Road, Chengdu (China)

2014-11-15

Highlights: • Tungsten coatings on graphite were firstly obtained by electro-deposition method via Na{sub 2}WO{sub 4}–WO{sub 3} molten salt system. • Uniform and dense tungsten coatings could be easily prepared in each face of the sample, especially the complex components. • The obtained tungsten coatings are with high purity, ultra-low oxygen content (about 0.022 wt%). • Modulate pulse parameters can get tungsten coatings with different thickness and hardness. - Abstract: Tungsten coating on graphite substrate is one of the most promising candidate materials as the ITER plasma facing components. In this paper, tungsten coatings on graphite substrates were fabricated by electro-deposition from Na{sub 2}WO{sub 4}–WO{sub 3} molten salt system at 1173 K in atmosphere. Tungsten coatings with no impurities were successfully deposited on graphite substrates under various pulsed current densities in an hour. By increasing the current density from 60 mA cm{sup −2} to 120 mA cm{sup −2} an increase of the average size of tungsten grains, the thickness and the hardness of tungsten coatings occurs. The average size of tungsten grains can reach 7.13 μm, the thickness of tungsten coating was in the range of 28.8–51 μm, and the hardness of coating was higher than 400 HV. No cracks or voids were observed between tungsten coating and graphite substrate. The oxygen content of tungsten coating is about 0.022 wt%.

An assessment of the tritium inventory in, permeation through and releases from the NET plasma facing materials

International Nuclear Information System (INIS)

Wu, C.H.

1986-01-01

The tritium retention, permeation and release characteristics of D-T tokamaks are extremely important from both an environmental and a plasma physics point of view. Tokamak measurements have demonstrated that release of retained hydrogen isotopes by plasma-wall interactions play a dominant role in fuel recycling during a discharge. In addition, retained tritium in the plasma facing materials may contribute substantially to the on-site tritium inventory of D-T devices. Austenitic and martensitic steels are being considered as first wall materials. Tungsten and molybdenum will be possibly used as divertor armour materials for NET. By using a computer code, the tritium inventory in, permeation through and release from these materials have been calculated as functions of material thickness, temperature and impinging fluxes. It is shown that the tritium inventory in the first wall will be strongly affected by the temperature gradient in the materials. It is evident, that the tritium permeation as well as the tritium inventory can be reduced appropriately by controlling the temperatures at the plasma and cooling sides of the first wall. The results are discussed and the possible consequences are analysed. (author)

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

Science.gov (United States)

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.

Damage to preheated tungsten targets after multiple plasma impacts simulating ITER ELMs

Energy Technology Data Exchange (ETDEWEB)

Garkusha, I.E. [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine)], E-mail: garkusha@ipp.kharkov.ua; Bandura, A.N.; Byrka, O.V.; Chebotarev, V.V. [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine); Landman, I. [Forschungszentrum Karlsruhe, IHM, 76021 Karlsruhe (Germany); Makhlaj, V.A. [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine); Pestchanyi, S. [Forschungszentrum Karlsruhe, IHM, 76021 Karlsruhe (Germany); Tereshin, V.I. [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine)

2009-04-30

The behavior of a preheated at 650 deg. C tungsten targets under repetitive ELM-like plasma pulses is studied in simulation experiments with the quasi-stationary plasma accelerator QSPA Kh-50. The targets have been exposed up to 350 pulses of the duration 0.25 ms and the surface heat loads either 0.45 MJ/m{sup 2} or 0.75 MJ/m{sup 2}, which is below and above the melting threshold, respectively. The development of surface morphology of the exposed targets as well as cracking and swelling at the surface is discussed. First comparisons of obtained experimental results with corresponding numerical simulations of the code PEGASUS-3D are presented.

Comparative Investigation of Tungsten Fibre Nets Reinforced Tungsten Composite Fabricated by Three Different Methods

Directory of Open Access Journals (Sweden)

Linhui Zhang

2017-07-01

Full Text Available Tungsten fibre nets reinforced tungsten composites (Wf/W containing four net layers were fabricated by spark plasma sintering (SPS, hot pressing (HP and cold rolling after HP (HPCR, with the weight fraction of fibres being 17.4%, 10.5% and 10.5%, respectively. The relative density of the HPCRed samples is the highest (99.8% while that of the HPed composites is the lowest (95.1%. Optical and scanning electron microscopy and electron back scattering diffraction were exploited to characterize the microstructure, while tensile and hardness tests were used to evaluate the mechanical properties of the samples. It was found that partial recrystallization of fibres occurred after the sintering at 1800 °C. The SPSed and HPed Wf/W composites begin to exhibit plastic deformation at 600 °C with tensile strength (TS of 536 and 425 MPa and total elongation at break (TE of 11.6% and 23.0%, respectively, while the HPCRed Wf/W composites exhibit plastic deformation at around 400 °C. The TS and TE of the HPCRed Wf/W composites at 400 °C are 784 MPa and 8.4%, respectively. The enhanced mechanical performance of the Wf/W composites over the pure tungsten can be attributed to the necking, cracking, and debonding of the tungsten fibres.

Optical spectroscopy of free-propagating plasma and its interaction with tungsten targets in PF-1000 facility

Energy Technology Data Exchange (ETDEWEB)

Skladnik-Sadowska, E.; Malinowski, K. [The Andrzej Soltan Institute for Nuclear Studies, IPJ, 05-400 Otwock-Swierk (Poland); Sadowski, M.J. [The Andrzej Soltan Institute for Nuclear Studies, IPJ, 05-400 Otwock-Swierk (Poland)] [Institute of Plasma Physics and Laser Microfusion, IPPLM, 01-497 Warsaw (Poland); Kubkowska, M.; Jakubowska, K.; Paduch, M.; Scholz, M. [Institute of Plasma Physics and Laser Microfusion, IPPLM, 01-497 Warsaw (Poland); Garkusha, I.E.; Ladygina, M.; Tereshin, V.I. [Institute of Plasma Physics, NSC KIPT, 61-108 Kharkov (Ukraine)

2011-07-01

The paper reports on optical spectroscopy of pulsed plasma streams during their free propagation within a vacuum chamber and their interaction with tungsten targets. Experiments were performed with the PF-1000 facility and particular attention was paid to improvements in spectroscopic diagnostics techniques. In contrary to preliminary studies, the recent spectroscopic measurements of the free plasma streams were carried out perpendicular to the z-axis and at a larger distance from the electrode outlet. The center of the observation quartz-window was located at z = 30 cm in order to observe first a pure deuterium-plasma stream, and later on some heavy impurities which might reach that distance with a delay induced by differences in their production and time-of-flight. The recorded spectral lines were identified by means of a Kurucz database. It was confirmed that at the pure D{sub 2}-filling the PF-1000 facility emits first the deuterium-plasma stream and one can observe intense deuterium Balmer lines, but at a distance z = 30 cm, after about 2 microseconds there appear many impurity lines originating mainly from the Cu-electrodes, i.e. Cu-lines. The second part of the experiment concerned the spectroscopic measurements of metal plasma 'pillow' produced by the plasma stream impinging upon a solid target made of pure tungsten. The described measurements enabled the most intense spectral lines to be identified. This document is composed of an abstract followed by the slides of the presentation

Modification of vacuum plasma sprayed tungsten coating on reduced activation ferritic/martensitic steels by friction stir processing

International Nuclear Information System (INIS)

Tanigawa, Hiroyasu; Ozawa, Kazumi; Morisada, Yoshiaki; Noh, Sanghoon; Fujii, Hidetoshi

2015-01-01

Highlights: • Friction stir processing (FSP) was applied on vacuum plasma spray (VPS) W to improve its low thermal conductivity and weakness due to high porosity. • FSP can achieve significant improvement both in mechanical and thermal properties of VPS-W coating. • It was indicated that the double pass FSP at 600 rpm/50 mm/min/2 ton on VPS-W show the most dense microstructure and hardest mechanical property. • Hardness test over FSPed VPS-W layer revealed that the hardness of W becomes higher than that of bulk W. • The thermal conductivity of double pass FSPed VPS-W was about 80% of bulk W at 200 °C, and it becomes equivalent to that of bulk W over 800 °C. - Abstract: Tungsten (W) is the primary candidate material as a plasma facing material in fusion devices, as for its high melting temperature, good thermal conductivity and low sputtering rate, and vacuum plasma spray (VPS) technique is preferred as it is applicable for large area without brittle interlayer, but the thermal conductivity of W layer is very poor, and easy to detach, mainly caused by its porous structure. W Friction stir processing (FSP) was applied on VPS-W to improve these poor properties, and it was suggested that FSP can contribute to significant improvement in both mechanical and thermal properties of the VPS-W coating.

Surface damage characteristics of CFC and tungsten with repetitive ELM-like pulsed plasma irradiation

Energy Technology Data Exchange (ETDEWEB)

Kikuchi, Y., E-mail: ykikuchi@eng.u-hyogo.ac.jp [Graduate School of Engineering, University of Hyogo, 2167 Shosha, Himeji, 671-2280 Hyogo (Japan); Nishijima, D. [Center for Energy Research, University of California at San Diego, 9500 Gilman Drive, La Jolla, CA 92093-0417 (United States); Nakatsuka, M.; Ando, K.; Higashi, T.; Ueno, Y.; Ishihara, M.; Shoda, K.; Nagata, M. [Graduate School of Engineering, University of Hyogo, 2167 Shosha, Himeji, 671-2280 Hyogo (Japan); Kawai, T.; Ueda, Y. [Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Fukumoto, N. [Graduate School of Engineering, University of Hyogo, 2167 Shosha, Himeji, 671-2280 Hyogo (Japan); Doerner, R.P. [Center for Energy Research, University of California at San Diego, 9500 Gilman Drive, La Jolla, CA 92093-0417 (United States)

2011-08-01

Surface damage of carbon fiber composite (CFC) and tungsten (W) due to repetitive ELM-like pulsed plasma irradiation has been investigated by using a magnetized coaxial plasma gun. CX2002U CFC and stress-relieved W samples were exposed to repetitive pulsed deuterium plasmas with duration of {approx}0.5 ms, incident ion energy of {approx}30 eV, and surface absorbed energy density of {approx}0.3-0.7 MJ/m{sup 2}. Bright spots on a CFC surface during pulsed plasma exposures were clearly observed with a high-speed camera, indicating a local surface heating. No melting of a W surface was observed under a single plasma pulse exposure at energy density of {approx}0.7 MJ/m{sup 2}, although cracks were formed. Cracking of the W surface grew with repetitive pulsed plasma exposures. Subsequently, the surface melted due to localized heat absorption.

Toughness enhancement of tungsten reinforced with short tungsten fibres

Energy Technology Data Exchange (ETDEWEB)

Jiang, Y. [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zhang, L.H. [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Fang, Q.F., E-mail: qffang@issp.ac.cn [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Zhang, T.; Wang, X.P.; Hao, T.; Liu, C.S. [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, Hefei 230031 (China)

2017-04-06

The feasibility and toughening efficiency of the short tungsten fibre reinforcement on tungsten were investigated in W{sub f}/W composites fabricated by powder metallurgy method of spark plasma sintering. Fibres in the composites presented a Z-free laminar structure. Partial recrystallization of fibre grains occurred but fibre crack or damage was not detected. Fracture energy of W{sub f}/W composites was estimated in tensile tests, and the results indicated great toughness improvement over pure tungsten in virtue of frictional pullout and plastic deformation of fibres, and matrix-fibres interfacial debonding since 873 K. The specimen with mass fraction of 10% and fibre diameter of 100 µm exhibits the largest elongation of 9±1.1% and the highest ultimate strength of 482±13 MPa at 873 K.

Influence of boronization on operation with high-Z plasma facing components in Alcator C-Mod

International Nuclear Information System (INIS)

Lipschultz, B.; Lin, Y.; Marmar, E.S.; Whyte, D.G.; Wukitch, S.; Hutchinson, I.H.; Irby, J.; LaBombard, B.; Reinke, M.L.; Terry, J.L.; Wright, G.

2007-01-01

We report the results of operation of Alcator C-Mod with all high-Z molybdenum plasma facing component (PFC) surfaces. Without boron-coated PFCs energy confinement was poor (H ITER,89 ∼ 1) due to high core molybdenum (n Mo /n e ≤ 0.1%) and radiation. After applying boron coatings, n Mo /n e was reduced by a factor of 10-20 with H ITER,89 approaching 2. Results of between-discharge boronization, localized at various major radii, point towards important molybdenum source regions being small, outside the divertor, and due to RF-sheath-rectification. Boronization also has a significant effect on the plasma startup phase lowering Z eff , radiation, and lowering the runaway electron damage. The requirement of low-Z coatings over at least a fraction of the Mo PFCs in C-Mod for best performance together with the larger than expected D retention in Mo, give impetus for further high-Z PFC investigations to better predict the performance of un-coated tungsten surfaces in ITER and beyond

Development of bonding techniques between W and Cu-alloys for plasma facing components by HIP method (3). Bonding tests with Au-foil insert

International Nuclear Information System (INIS)

Saito, Shigeru

2002-07-01

In recent years, it has been considered that W (tungsten) is one of candidate materials for armor tiles of plasma a facing components (PFC), like first wall or divertor, of fusion reactor. On the other hand, Cu-alloys, like OFHC-Cu or DS-Cu, are proposed as heat sink materials behind the plasma facing materials because of its high thermal conductivity. It is necessary to develop a reliable bonding techniques in order to fabricate PFC. JAERI has developed the hot isostatic press (HIP) bonding process to bond W with Cu-alloys. In this experiments, bonding tests with Au-foil insert were performed. We could get the best HIP bonding conditions for W and Cu-alloys with Au-foil as 1123K x 2hours x 147MPa. It was shown that the HIP temperature was 150K lower than that of without Au-foil. Furthermore, the tensile strength was similar to that of with without Au-foil. (author)

Plasma Spraying and Characterization of Tungsten Carbide-Cobalt Coatings by the Water-Stabilized System WSP

Czech Academy of Sciences Publication Activity Database

Ctibor, Pavel; Kašparová, M.; Bellin, J.; Le Guen, E.; Zahálka, F.

2009-01-01

Roč. 2009, - (2009), s. 1-11 ISSN 1687-8434 R&D Projects: GA AV ČR 1QS200430560 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tungsten karbide – cobalt, cermet * wear resistance * abrasion * plasma spraying Subject RIV: JG - Metallurgy http://www.hindawi.com/journals/amse/2009/254848.html

Exfoliation of the tungsten fibreform nanostructure by unipolar arcing in the LHD divertor plasma

Science.gov (United States)

Tokitani, M.; Kajita, S.; Masuzaki, S.; Hirahata, Y.; Ohno, N.; Tanabe, T.; LHD Experiment Group

2011-10-01

The tungsten nanostructure (W-fuzz) created in the linear divertor simulator (NAGDIS) was exposed to the Large Helical Device (LHD) divertor plasma for only 2 s (1 shot) to study exfoliation/erosion and microscopic modifications due to the high heat/particle loading under high magnetic field conditions. Very fine and randomly moved unipolar arc trails were clearly observed on about half of the W-fuzz area (6 × 10 mm2). The fuzzy surface was exfoliated by continuously moving arc spots even for the very short exposure time. This is the first observation of unipolar arcing and exfoliation of some areas of the W-fuzz structure itself in a large plasma confinement device with a high magnetic field. The typical width and depth of each arc trail were about 8 µm and 1 µm, respectively, and the arc spots moved randomly on the micrometre scale. The fractality of the arc trails was analysed using a box-counting method, and the fractal dimension (D) of the arc trails was estimated to be D ≈ 1.922. This value indicated that the arc spots moved in Brownian motion, and were scarcely influenced by the magnetic field. One should note that such a large scale exfoliation due to unipolar arcing may enhance the surface erosion of the tungsten armour and act as a serious impurity source for fusion plasmas.

Exfoliation of the tungsten fibreform nanostructure by unipolar arcing in the LHD divertor plasma

International Nuclear Information System (INIS)

Tokitani, M.; Masuzaki, S.; Kajita, S.; Hirahata, Y.; Ohno, N.; Tanabe, T.

2011-01-01

The tungsten nanostructure (W-fuzz) created in the linear divertor simulator (NAGDIS) was exposed to the Large Helical Device (LHD) divertor plasma for only 2 s (1 shot) to study exfoliation/erosion and microscopic modifications due to the high heat/particle loading under high magnetic field conditions. Very fine and randomly moved unipolar arc trails were clearly observed on about half of the W-fuzz area (6 x 10 mm 2 ). The fuzzy surface was exfoliated by continuously moving arc spots even for the very short exposure time. This is the first observation of unipolar arcing and exfoliation of some areas of the W-fuzz structure itself in a large plasma confinement device with a high magnetic field. The typical width and depth of each arc trail were about 8 μm and 1 μm, respectively, and the arc spots moved randomly on the micrometre scale. The fractality of the arc trails was analysed using a box-counting method, and the fractal dimension (D) of the arc trails was estimated to be D ∼ 1.922. This value indicated that the arc spots moved in Brownian motion, and were scarcely influenced by the magnetic field. One should note that such a large scale exfoliation due to unipolar arcing may enhance the surface erosion of the tungsten armour and act as a serious impurity source for fusion plasmas. (letter)

The Influence of Interface Characteristics on the Adhesion/Cohesion of Plasma Sprayed Tungsten Coatings

Czech Academy of Sciences Publication Activity Database

Matějíček, Jiří; Vilémová, Monika; Mušálek, Radek; Sachr, P.; Horník, J.

2013-01-01

Roč. 3, č. 2 (2013), s. 108-125 ISSN 2079-6412 R&D Projects: GA ČR(CZ) GAP108/12/1872; GA MPO FR-TI2/702 Grant - others:EFDA(XE) WP12-MAT-01-HHFM Institutional support: RVO:61389021 Keywords : Tungsten * plasma spraying * adhesion * cohesion * PVD * interlayers Subject RIV: JG - Metallurgy http://www.mdpi.com/2079-6412/3/2/108

Development and evaluation of plasma facing materials for future thermonuclear fusion reactors

International Nuclear Information System (INIS)

Linke, J.; Pintsuk, G.; Roedig, M.; Schmidt, A.; Thomser, C.

2010-01-01

sink materials as well as reliable fabrication processes for actively cooled plasma facing components. The technical solutions which are considered today are mainly based on the PFMs beryllium, carbon or tungsten joined to copper alloys or stainless steel heat sinks. To test and to demonstrate the acceptability of plasma facing materials and components special high heat flux test facilities based on intense ion or electron beams are being used routinely to demonstrate the heat removal efficiency and the lifetime under fusion specific loading conditions. In addition to the above mentioned quasi-stationary heat loads, short transient thermal pulses with deposited energy densities up to several tens of MJm -2 are a serious concern for next step tokamak devices. The most frequent events are so-called Edge Localized Modes (type I ELMs) and plasma disruptions. Here a considerable fraction of the plasma energy is deposited on a localized surface area in the divertor strike zone; the time scale of these events is typically in the order of 1 ms. As a consequence, thermal shock induced crack formation, vaporization, surface melting and droplet ejection as well as particle emission induced by brittle destruction processes will limit the lifetime of the components. This is also valid for instabilities in the plasma positioning (vertical displacement events) which cause irreversible damage to plasma facing components, particularly to the metallic wall armour. Moreover, dust particles (neutron activated or toxic metals or tritium enriched carbon) are a serious concern from a safety point of view. In order to investigate the thermally induced plasma wall interaction under fusion specific thermal loads, high heat flux simulation tests are performed in electron or ion beam test facilities as well as in quasi stationary plasma devices. These experiments cover thermal fatigue loads and/or thermal shock tests with relevant operational loading conditions. Furthermore, the wall bombardment

Development and evaluation of plasma facing materials for future thermonuclear fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Linke, J.; Pintsuk, G.; Roedig, M.; Schmidt, A.; Thomser, C. [Forschungszentrum Juelich GmbH, EURATOM Association, Juelich (Germany)

2010-07-01

and heat sink materials as well as reliable fabrication processes for actively cooled plasma facing components. The technical solutions which are considered today are mainly based on the PFMs beryllium, carbon or tungsten joined to copper alloys or stainless steel heat sinks. To test and to demonstrate the acceptability of plasma facing materials and components special high heat flux test facilities based on intense ion or electron beams are being used routinely to demonstrate the heat removal efficiency and the lifetime under fusion specific loading conditions. In addition to the above mentioned quasi-stationary heat loads, short transient thermal pulses with deposited energy densities up to several tens of MJm{sup -2} are a serious concern for next step tokamak devices. The most frequent events are so-called Edge Localized Modes (type I ELMs) and plasma disruptions. Here a considerable fraction of the plasma energy is deposited on a localized surface area in the divertor strike zone; the time scale of these events is typically in the order of 1 ms. As a consequence, thermal shock induced crack formation, vaporization, surface melting and droplet ejection as well as particle emission induced by brittle destruction processes will limit the lifetime of the components. This is also valid for instabilities in the plasma positioning (vertical displacement events) which cause irreversible damage to plasma facing components, particularly to the metallic wall armour. Moreover, dust particles (neutron activated or toxic metals or tritium enriched carbon) are a serious concern from a safety point of view. In order to investigate the thermally induced plasma wall interaction under fusion specific thermal loads, high heat flux simulation tests are performed in electron or ion beam test facilities as well as in quasi stationary plasma devices. These experiments cover thermal fatigue loads and/or thermal shock tests with relevant operational loading conditions. Furthermore, the

Analysis of magnetic-dipole transitions in tungsten plasmas using detailed and configuration-average descriptions

Science.gov (United States)

Na, Xieyu; Poirier, Michel

2017-06-01

This paper is devoted to the analysis of transition arrays of magnetic-dipole (M1) type in highly charged ions. Such transitions play a significant role in highly ionized plasmas, for instance in the tungsten plasma present in tokamak devices. Using formulas recently published and their implementation in the Flexible Atomic Code for M1-transition array shifts and widths, absorption and emission spectra arising from transitions inside the 3*n complex of highly-charged tungsten ions are analyzed. A comparison of magnetic-dipole transitions with electric-dipole (E1) transitions shows that, while the latter are better described by transition array formulas, M1 absorption and emission structures reveal some insufficiency of these formulas. It is demonstrated that the detailed spectra account for significantly richer structures than those predicted by the transition array formalism. This is due to the fact that M1 transitions may occur between levels inside the same relativistic configuration, while such inner configuration transitions are not accounted for by the currently available averaging expression. In addition, because of configuration interaction, transition processes involving more than one electron jump, such as 3p1/23d5/2 → 3p3/23d3/2, are possible but not accounted for in the transition array formulas. These missing transitions are collected in pseudo-arrays using a post-processing method described in this paper. The relative influence of inner- and inter-configuration transitions is carefully analyzed in cases of tungsten ions with net charge around 50. The need for an additional theoretical development is emphasized.

Power handling of a segmented bulk W tile for JET under realistic plasma scenarios

Science.gov (United States)

Jet-Efda Contributors Mertens, Ph.; Coenen, J. W.; Eich, T.; Huber, A.; Jachmich, S.; Nicolai, D.; Riccardo, V.; Senik, K.; Samm, U.

2011-08-01

A solid tungsten divertor row has been designed for JET in the frame of the ITER-like Wall project (ILW). The plasma-facing tiles are segmented in four stacks of tungsten lamellae oriented in the toroidal direction. Earlier estimations of the expected tile performance were carried out mostly for engineering purposes, to compare the permissible heat load with the power density of 7 MW/m2 originally specified for the ILW as a uniform load for 10 s.The global thermal model developed for the W modules delivers results for more realistic plasma footprints: the poloidal extension of the outer strike point was reduced from the full lamella width of 62 mm to ⩾15 mm. Model validation is given by the experimental exposure of a 1:1 prototype stack in the ion beam facility MARION (incidence ˜6°, load E ⩽ 66 MJ/m2 on the wetted surface). Spreading the deposited energy by appropriate sweeping over one or several stacks in the torus is beneficial for the tungsten lamellae and for the support structure.

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

Energy Technology Data Exchange (ETDEWEB)

Fan, Jinglian, E-mail: fjl@csu.edu.cn; Han, Yong; Li, Pengfei; Sun, Zhiyu; Zhou, Qiang

2014-12-15

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 (Y{sub 2}O{sub 3}/La{sub 2}O{sub 3}) 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/m{sup 2} (5 ms)

Experimental study of plasma energy transfer and material erosion under ELM-like heat loads

Energy Technology Data Exchange (ETDEWEB)

Garkusha, I.E., E-mail: garkusha@ipp.kharkov.u [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine); Makhlaj, V.A.; Chebotarev, V.V. [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine); Landman, I. [Forschungszentrum Karlsruhe, IHM, 76021 Karlsruhe (Germany); Tereshin, V.I.; Aksenov, N.N.; Bandura, A.N. [Institute of Plasma Physics of the NSC KIPT, Akademicheskaya 1, 61108 Kharkov (Ukraine)

2009-06-15

Main features of plasma-surface interaction and energy transfer to tokamak plasma facing components are studied at different heat loads in ELM simulation experiments with the plasma gun QSPA Kh-50. Repetitive plasma exposures of tungsten, graphite and different combined W-C targets were performed at the pulse duration of 0.25 ms and the heat loads varied in the range 0.2-2.5 MJ/m{sup 2}. The onset of vapor shield in front of the surface was investigated. The evaporation is immediately followed by a saturation of surface heat load if further increasing the impact energy. The presence of graphite essentially decreases the heat flux to the nearby tungsten surface, which is due to the carbon vapor shield. Droplet splashing at the tungsten surface and formation of hot spots on the graphite surface are discussed.

Experimental study of plasma energy transfer and material erosion under ELM-like heat loads

International Nuclear Information System (INIS)

Garkusha, I.E.; Makhlaj, V.A.; Chebotarev, V.V.; Landman, I.; Tereshin, V.I.; Aksenov, N.N.; Bandura, A.N.

2009-01-01

Main features of plasma-surface interaction and energy transfer to tokamak plasma facing components are studied at different heat loads in ELM simulation experiments with the plasma gun QSPA Kh-50. Repetitive plasma exposures of tungsten, graphite and different combined W-C targets were performed at the pulse duration of 0.25 ms and the heat loads varied in the range 0.2-2.5 MJ/m 2 . The onset of vapor shield in front of the surface was investigated. The evaporation is immediately followed by a saturation of surface heat load if further increasing the impact energy. The presence of graphite essentially decreases the heat flux to the nearby tungsten surface, which is due to the carbon vapor shield. Droplet splashing at the tungsten surface and formation of hot spots on the graphite surface are discussed.

Strain-dependent diffusion behavior of H within tungsten

International Nuclear Information System (INIS)

Ding, Wenyi; He, Haiyan; Liu, Changsong; Ding, Rui; Chen, Junling; Pan, Bicai

2014-01-01

The diffusion behaviors of H in tungsten, a promising material serving as the first wall facing the plasma in nuclear reactors, under either biaxial strain or isotropic strain are theoretically studied. We find that under the isotropic strain, an individual H atom may diffuse along all pathways, and under the biaxial strain, it preferably migrates along the direction perpendicular to the loaded strain. Moreover, in the case of either the isotropic or the biaxial strain, the loaded compressive strain weakens the diffusion of H, while the loaded tensile strain enhances the diffusion of H in bulk W.

Strain-dependent diffusion behavior of H within tungsten

Energy Technology Data Exchange (ETDEWEB)

Ding, Wenyi; He, Haiyan [Department of Physics, University of Science and Technology of China, Hefei, Anhui 230026 (China); Liu, Changsong [Key Laboratory of Materials Physics, Institute of Solid State Physics, Chinese Academy of Sciences, P.O. Box 1129, Hefei 230031 (China); Ding, Rui; Chen, Junling [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Pan, Bicai, E-mail: bcpan@ustc.edu.cn [Department of Physics, University of Science and Technology of China, Hefei, Anhui 230026 (China); Hefei National Laboratory for Physical Science at Microscale, University of Science and Technology of China, Hefei, Anhui 230026 (China)

2014-06-15

The diffusion behaviors of H in tungsten, a promising material serving as the first wall facing the plasma in nuclear reactors, under either biaxial strain or isotropic strain are theoretically studied. We find that under the isotropic strain, an individual H atom may diffuse along all pathways, and under the biaxial strain, it preferably migrates along the direction perpendicular to the loaded strain. Moreover, in the case of either the isotropic or the biaxial strain, the loaded compressive strain weakens the diffusion of H, while the loaded tensile strain enhances the diffusion of H in bulk W.

Structural impact of creep in tungsten monoblock divertor target at 20 MW/m2

Directory of Open Access Journals (Sweden)

Muyuan Li

2018-01-01

Full Text Available In order to increase erosion lifetime of the divertor target, in the 2nd design phase of R&D work package ‘Divertor’ for European DEMO, armor thickness of tungsten monoblock divertor target is increased from 5 mm to 8 mm. By increasing armor thickness, surface temperature increases nearly linearly, which makes effect of creep no longer negligible at slow transients of 20 MW/m2. In this work, structural impact of creep in tungsten monoblock divertor target is for the first time quantitatively analyzed with the aid of finite element method. The numerical simulations have revealed that creep results in an increase of inelastic strain accumulation. With increasing armor thickness, tensile surface stress along x-axis (the longer edge at the plasma-facing surface of tungsten monoblock reduces, while surface stress along z-axis (axial direction of the cooling tube changes from tensile to compressive. Creep will accelerate this change. With increasing grain size, creep strain accumulation at loading surface increases due to higher creep rates, while plastic strain accumulation decreases. Creep can mitigate the risk of deep cracking by reducing the driving force for crack opening, and has a positive impact for preventing the contact between the upper parts of neighboring monoblocks in high heat flux tests.

Joint absorption of lithium and oxygen on the tungsten (100) face

International Nuclear Information System (INIS)

Gupalo, M.S.; Smereka, T.P.; Babkin, G.V.; Palyukh, B.M.

1981-01-01

The paper deals with studying the effect of oxygen on emission-adsorption properties of metal-film Li-W system. Data on work of phi yield and adsorption heat q of lithium on tungsten face (100), preliminarily coated with different quantity of oxygen, are obtained. The method of contact potential difference was used. Strong decrease of yield work with the increase of oxygen coatings, as well as essential growth of initial adsorption heat are observed. Temperature dependence of phi is not practically observed. The data obtained point out, that oxygen presence on the surface causes increase of dipole momentum of lithium adatoms, which results in bond energy growth and reduction phi minimal value of lithium in the presence of oxygen phi=2 eV, q=2.2 eV is obtained at optimal compositions of oxygen-lithium layers on W (100) [ru

Microstructural stability of spark-plasma-sintered Wf/W composite with zirconia interface coating under high-heat-flux hydrogen beam irradiation

OpenAIRE

M. Avello de Lama; M. Balden; H. Greuner; T. Höschen; J. Matejicek; J.H. You

2017-01-01

Tungsten is considered as the most suitable material for the plasma-facing armour of future fusion reactors. However, in spite of many advantageous properties, pure tungsten has a major drawback, namely, brittleness at lower temperatures and embrittlement by neutron irradiation. Tungsten fibre-reinforced tungsten (Wf/W) composites are thought to be a promising candidate material for armour owing to the pseudo-toughness effect which is based on controlled cracking of coated interfaces. In this...

Modification of adhered dust on plasma-facing surfaces due to exposure to ELMy H-mode plasma in DIII-D

Directory of Open Access Journals (Sweden)

I. Bykov

2017-08-01

Full Text Available Transient heat load tests have been conducted in the lower divertor of DIII-D using DiMES manipulator in order to study the behavior of dust on tungsten Plasma Facing Components (PFCs during ELMy H-mode discharges. Samples with pre-adhered, pre-characterized dust have been exposed at the outer strike point (OSP in a series of discharges with varied intra-(inter- ELM heat fluxes. We used C dust because of its high sublimation temperature and non-metal properties. Al dust as a surrogate for Be and W dust were employed as relevant to that in the ITER divertor. The poor initial thermal contact between the substrate and the particles led to overheating, sublimation and shrinking of the carbon dust, and wetting induced coagulation of Al dust. Little modification of the W dust was observed. An enhanced surface adhesion and improvement of the thermal contact of C and Al dust were the result of exposure. A post mortem “adhesive tape” sampling showed that 70% of Al, <5% of W and C particles could not be removed from the surface owing to the improved adhesion. Al and C but not W particles that could be lifted had W inclusions indicating damage to the substrate. This suggests that non destructive methods may be inefficient for removal of dust in ITER.

Investigation on the effect of temperature excursion on the helium defects of tungsten surface by using compact plasma device

International Nuclear Information System (INIS)

Takamura, S.; Miyamoto, T.; Tomida, Y.; Minagawa, T.; Ohno, N.

2011-01-01

The effects of temperature excursion on the helium defects of tungsten surface have been investigated by using compact plasma device AIT-PID (Aichi Institute of Technology - Plasma Irradiation Device). An initial stage of bubble formation has been identified with an order of smaller (sub-micron) bubbles and holes than those in the past in which the micron size is the standard magnitude. The radiation cooling has been detected when a blacking of tungsten surface coming from nanostructure formation is proceeding due to an increase in the emissivity. The temperature increase to the domain (∼1600 K) in bubble/hole formation from that in nanostructure formation has been found to bring a constriction in diameter and a reduction in length of fiber-form nanostructure.

Extension of the ECRH operational space with O2 and X3 heating schemes to control tungsten accumulation in ASDEX Upgrade

Science.gov (United States)

Höhnle, H.; Stober, J.; Herrmann, A.; Kasparek, W.; Leuterer, F.; Monaco, F.; Neu, R.; Schmid-Lorch, D.; Schütz, H.; Schweinzer, J.; Stroth, U.; Wagner, D.; Vorbrugg, S.; Wolfrum, E.; ASDEX Upgrade Team

2011-08-01

ASDEX Upgrade has been operated with tungsten-coated plasma-facing components for several years. H-mode operation with good confinement has been demonstrated. Nevertheless, purely neutral beam injection-heated H-modes with reduced gas puff, moderate heating power or/and increased triangularity tend to accumulate tungsten, followed by a radiative collapse. Under these conditions, central electron heating with electron cyclotron resonance heating (ECRH), usually in X2 polarization, changes the impurity transport in the plasma centre, reducing the central tungsten concentration and, in many cases, stabilizing the plasma. In order to extend the applicability of central ECRH to a wider range of magnetic field and plasma current additional ECRH schemes with reduced single-pass absorption have been implemented: X3 heating allows us to reduce the magnetic field by 30%, such that the first H-modes with an ITER-like value of the safety factor of q95 = 3 could be run in the tungsten-coated device. O2 heating increases the cutoff density by a factor of 2 allowing higher currents and triangularities to be addressed. For both schemes, scenarios have been developed to cope with the associated reduced absorption. In the case of central X3 heating, the X2 resonance lies close to the pedestal top at the high-field side of the plasma, serving as a beam dump. For O2, holographic mirrors have been developed which guarantee a second pass through the plasma centre. The beam position on these reflectors is controlled by fast thermocouples. Stray-radiation protection has been implemented using sniffer probes.

Extension of the ECRH operational space with O2 and X3 heating schemes to control tungsten accumulation in ASDEX Upgrade

International Nuclear Information System (INIS)

Hoehnle, H.; Kasparek, W.; Stroth, U.; Stober, J.; Herrmann, A.; Leuterer, F.; Monaco, F.; Neu, R.; Schmid-Lorch, D.; Schuetz, H.; Schweinzer, J.; Wagner, D.; Vorbrugg, S.; Wolfrum, E.

2011-01-01

ASDEX Upgrade has been operated with tungsten-coated plasma-facing components for several years. H-mode operation with good confinement has been demonstrated. Nevertheless, purely neutral beam injection-heated H-modes with reduced gas puff, moderate heating power or/and increased triangularity tend to accumulate tungsten, followed by a radiative collapse. Under these conditions, central electron heating with electron cyclotron resonance heating (ECRH), usually in X2 polarization, changes the impurity transport in the plasma centre, reducing the central tungsten concentration and, in many cases, stabilizing the plasma. In order to extend the applicability of central ECRH to a wider range of magnetic field and plasma current additional ECRH schemes with reduced single-pass absorption have been implemented: X3 heating allows us to reduce the magnetic field by 30%, such that the first H-modes with an ITER-like value of the safety factor of q 95 = 3 could be run in the tungsten-coated device. O2 heating increases the cutoff density by a factor of 2 allowing higher currents and triangularities to be addressed. For both schemes, scenarios have been developed to cope with the associated reduced absorption. In the case of central X3 heating, the X2 resonance lies close to the pedestal top at the high-field side of the plasma, serving as a beam dump. For O2, holographic mirrors have been developed which guarantee a second pass through the plasma centre. The beam position on these reflectors is controlled by fast thermocouples. Stray-radiation protection has been implemented using sniffer probes.

Ion cyclotron resonance heating for tungsten control in various JET H-mode scenarios

Science.gov (United States)

Goniche, M.; Dumont, R. J.; Bobkov, V.; Buratti, P.; Brezinsek, S.; Challis, C.; Colas, L.; Czarnecka, A.; Drewelow, P.; Fedorczak, N.; Garcia, J.; Giroud, C.; Graham, M.; Graves, J. P.; Hobirk, J.; Jacquet, P.; Lerche, E.; Mantica, P.; Monakhov, I.; Monier-Garbet, P.; Nave, M. F. F.; Noble, C.; Nunes, I.; Pütterich, T.; Rimini, F.; Sertoli, M.; Valisa, M.; Van Eester, D.; Contributors, JET

2017-05-01

Ion cyclotron resonance heating (ICRH) in the hydrogen minority scheme provides central ion heating and acts favorably on the core tungsten transport. Full wave modeling shows that, at medium power level (4 MW), after collisional redistribution, the ratio of power transferred to the ions and the electrons vary little with the minority (hydrogen) concentration n H/n e but the high-Z impurity screening provided by the fast ions temperature increases with the concentration. The power radiated by tungsten in the core of the JET discharges has been analyzed on a large database covering the 2013-2014 campaign. In the baseline scenario with moderate plasma current (I p = 2.5 MA) ICRH modifies efficiently tungsten transport to avoid its accumulation in the plasma centre and, when the ICRH power is increased, the tungsten radiation peaking evolves as predicted by the neo-classical theory. At higher current (3-4 MA), tungsten accumulation can be only avoided with 5 MW of ICRH power with high gas injection rate. For discharges in the hybrid scenario, the strong initial peaking of the density leads to strong tungsten accumulation. When this initial density peaking is slightly reduced, with an ICRH power in excess of 4 MW,very low tungsten concentration in the core (˜10-5) is maintained for 3 s. MHD activity plays a key role in tungsten transport and modulation of the tungsten radiation during a sawtooth cycle is correlated to the fishbone activity triggered by the fast ion pressure gradient.

Nanoindentation study of the combined effects of crystallography, heat treatment and exposure to high-flux deuterium plasma in tungsten

Energy Technology Data Exchange (ETDEWEB)

Zayachuk, Y., E-mail: yevhen.zayachuk@materials.ox.ac.uk [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Armstrong, D.E.J. [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Bystrov, K. [FOM Institute DIFFER- Dutch Institute for Fundamental Energy Research, Trilateral Euregio Cluster, De Zaale 20, 3612 AJ Eindhoven (Netherlands); Van Boxel, S. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Morgan, T. [FOM Institute DIFFER- Dutch Institute for Fundamental Energy Research, Trilateral Euregio Cluster, De Zaale 20, 3612 AJ Eindhoven (Netherlands); Roberts, S.G. [Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

2017-04-01

tungsten samples were heat-treated to achieve partial recrystallization and exposed to high ion flux deuterium plasma at different temperatures and fluences. Continuous stiffness nanoindentation measurements of near-surface hardness were performed in the grains of specific annealing states and of specific crystallographic orientation, determined by electron backscatter diffraction (EBSD); indentation pile-up was investigated using surface profilometry. Bulk hardness of unexposed tungsten does not strongly depend on grain orientation, but depends on the annealing state of the grain, with values between ∼4.3 GPa for recrystallized grains and ∼5.5 for non-recrystallized ones. Grains with <111> surface normal orientation feature the least pile-up, while grains with <001> orientation the most; pile-up also depends on the annealing state, being generally lower in recrystallized grains. Plasma exposure leads to the increase of hardness, most significantly near the surface. The width of plasma-affected zone increases with the increase of exposure temperature and fluence, as well in recrystallized grains, correlating with the increase of diffusion depth. Plasma exposure does not lead to the emergence of orientation-dependence of hardness. Both indentation pile-up and near-surface indentation pop-ins are generally suppressed by plasma exposure.

Heat loads to divertor nearby components from secondary radiation evolved during plasma instabilities

Energy Technology Data Exchange (ETDEWEB)

Sizyuk, V., E-mail: vsizyuk@purdue.edu; Hassanein, A., E-mail: hassanein@purdue.edu [Center for Materials under Extreme Environment, School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States)

2015-01-15

A fundamental issue in tokamak operation related to power exhaust during plasma instabilities is the understanding of heat and particle transport from the core plasma into the scrape-off layer and to plasma-facing materials. During abnormal and disruptive operation in tokamaks, radiation transport processes play a critical role in divertor/edge-generated plasma dynamics and are very important in determining overall lifetimes of the divertor and nearby components. This is equivalent to or greater than the effect of the direct impact of escaped core plasma on the divertor plate. We have developed and implemented comprehensive enhanced physical and numerical models in the upgraded HEIGHTS package for simulating detailed photon and particle transport in the evolved edge plasma during various instabilities. The paper describes details of a newly developed 3D Monte Carlo radiation transport model, including optimization methods of generated plasma opacities in the full range of expected photon spectra. Response of the ITER divertor's nearby surfaces due to radiation from the divertor-developed plasma was simulated by using actual full 3D reactor design and magnetic configurations. We analyzed in detail the radiation emission spectra and compared the emission of both carbon and tungsten as divertor plate materials. The integrated 3D simulation predicted unexpectedly high damage risk to the open stainless steel legs of the dome structure in the current ITER design from the intense radiation during a disruption on the tungsten divertor plate.

Melt layer macroscopic erosion of tungsten and other metals under plasma heat loads simulating ITER off-normal events

International Nuclear Information System (INIS)

Garkusha, I.E.; Bandura, A.N.; Byrka, O.V.; Kulik, N.V.; Landman, I.; Wuerz, H.

2002-01-01

This paper is focused on experimental analysis of metal layer erosion and droplet splashing of tungsten and other metals under heat loads typical for ITER FEAT off-normal events,such as disruptions and VDE's. Plasma pressure gradient action on melt layer results in erosion crater formation with mountains of displaced material at the crater edge. It is shown that macroscopic motion of melt layer and surface cracking are the main factors responsible for tungsten damage. Weight loss measurements of all exposed materials demonstrate inessential contribution of evaporation process to metals erosion

Influence of tungsten microstructure and ion flux on deuterium plasma-induced surface modifications and deuterium retention

NARCIS (Netherlands)

Buzi, L.; De Temmerman, G.; Unterberg, B.; M. Reinhart,; Dittmar, T.; Matveev, D.; Linsmeier, C.; Breuer, U.; Kreter, A.; Van Oost, G.

2015-01-01

The influence of surface temperature, particle flux density and material microstructure on the surface morphology and deuterium retention was studied by exposing tungsten targets (20 μm and 40 μm grain size) to deuterium plasma at the same particle fluence (1026 m−2) and

Heat transfer for plasma facing components

International Nuclear Information System (INIS)

Boyd, R.D.; Meng, X.; Maughan, H.

1995-01-01

Although the high heat flux requirements for plasma-facing components have been reduced drastically from 40.0 MW/m 2 to near 10.0 MW/m 2 , there are still some refinements needed. This paper highlights: (1) recent accomplishments and pinpoints new thermal solutions and problem areas of immediate concern to the development of plasma-facing components, and (2) next generation thermal hydraulic problems which must be addressed to insure safety and reliability in component operation. More specifically the near-term thermal hydraulic problems entail: (1) generating an appropriate data base to insure the development of single-side heat flux correlations; and (2) adapting the existing vast uniform heat flux literature to the case of non-uniform heat flux distributions found in plasma facing components in fusion reactors. Results are presented for the latter task which includes: (a) an accurate subcooled flow boiling curve correlation for the partial nucleate boiling regime which can be adapted using previously proposed correlations relating single-side boundary heat flux to heat transfer, in uniformly heated channels, (b) the evaluation of the possibility of using the existing literature directly with redefined parameters, and (c) an estimation of circumferential variations in the heat transfer coefficient

Recrystallization and grain growth induced by ELMs-like transient heat loads in deformed tungsten samples

Science.gov (United States)

Suslova, A.; El-Atwani, O.; Sagapuram, D.; Harilal, S. S.; Hassanein, A.

2014-11-01

Tungsten has been chosen as the main candidate for plasma facing components (PFCs) due to its superior properties under extreme operating conditions in future nuclear fusion reactors such as ITER. One of the serious issues for PFCs is the high heat load during transient events such as ELMs and disruption in the reactor. Recrystallization and grain size growth in PFC materials caused by transients are undesirable changes in the material, since the isotropic microstructure developed after recrystallization exhibits a higher ductile-to-brittle transition temperature which increases with the grain size, a lower thermal shock fatigue resistance, a lower mechanical strength, and an increased surface roughening. The current work was focused on careful determination of the threshold parameters for surface recrystallization, grain growth rate, and thermal shock fatigue resistance under ELM-like transient heat events. Transient heat loads were simulated using long pulse laser beams for two different grades of ultrafine-grained tungsten. It was observed that cold rolled tungsten demonstrated better power handling capabilities and higher thermal stress fatigue resistance compared to severely deformed tungsten. Higher recrystallization threshold, slower grain growth, and lower degree of surface roughening were observed in the cold rolled tungsten.

Design, R&D and commissioning of EAST tungsten divertor

Science.gov (United States)

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.

Behavior and microstructural changes in different tungsten-based materials under pulsed plasma loading

Czech Academy of Sciences Publication Activity Database

Vilémová, Monika; Pala, Zdeněk; Jäger, Aleš; Matějíček, Jiří; Chernyshova, M.; Kowalska-Strzeciwilk, E.; Gribkov, V. A.; Janata, Marek

2016-01-01

Roč. 9, December (2016), s. 123-127 ISSN 2352-1791. [International Conference of Fusion Reactor Material (ICFRM-17) /17./. Aachen, 11.10.2015-16.10.2015] R&D Projects: GA ČR(CZ) GA14-12837S Institutional support: RVO:61389021 ; RVO:68378271 Keywords : ODS tungsten * yttrium oxide * porosity * phase stability * monoclinic phase * plasma focus * damage * erosion Subject RIV: JJ - Other Materials; JJ - Other Materials (FZU-D) http://dx.doi.org/10.1016/j.nme.2016.06.004

Tritium saturation in plasma-facing materials surfaces

International Nuclear Information System (INIS)

Longhurst, G.R.; Anderl, R.A.; Pawelko, R.J.; Causey, R.A.; Federici, G.; Haasz, A.A.

1998-01-01

Plasma-facing components in the international thermonuclear experimental reactor (ITER) will experience high heat loads and intense plasma fluxes of order 10 20 -10 23 particles/m 2 s. Experiments on Be and W, two of the materials considered for use in ITER, have revealed that a tritium saturation phenomenon can take place under these conditions in which damage to the surface results that enhances the return of implanted tritium to the plasma and inhibits uptake of tritium. This phenomenon is important because it implies that tritium inventories due to implantation in these plasma-facing materials will probably be lower than was previously estimated using classical recombination-limited release at the plasma surface. Similarly, permeation through these components to the coolant streams should be reduced. In this paper we discuss evidences for the existence of this phenomenon, describe techniques for modeling it, and present results of the application of such modeling to prior experiments. (orig.)

Tritium saturation in plasma-facing materials surfaces

Energy Technology Data Exchange (ETDEWEB)

Longhurst, G.R.; Anderl, R.A.; Pawelko, R.J. [Idaho Nat. Eng. and Environ. Lab., Idaho Falls, ID (United States); Causey, R.A. [Sandia National Labs., Livermore, CA (United States); Federici, G. [ITER Garching Joint Work Site, Garching (Germany); Haasz, A.A. [Toronto Univ., ON (Canada). Inst. for Aerospace Studies

1998-10-01

Plasma-facing components in the international thermonuclear experimental reactor (ITER) will experience high heat loads and intense plasma fluxes of order 10{sup 20}-10{sup 23} particles/m{sup 2}s. Experiments on Be and W, two of the materials considered for use in ITER, have revealed that a tritium saturation phenomenon can take place under these conditions in which damage to the surface results that enhances the return of implanted tritium to the plasma and inhibits uptake of tritium. This phenomenon is important because it implies that tritium inventories due to implantation in these plasma-facing materials will probably be lower than was previously estimated using classical recombination-limited release at the plasma surface. Similarly, permeation through these components to the coolant streams should be reduced. In this paper we discuss evidences for the existence of this phenomenon, describe techniques for modeling it, and present results of the application of such modeling to prior experiments. (orig.) 39 refs.

Microstructural stability of spark-plasma-sintered W f /W composite with zirconia interface coating under high-heat-flux hydrogen beam irradiation.

Czech Academy of Sciences Publication Activity Database

Avello de Lama, M.; Balden, M.; Greuner, H.; Höschen, T.; Matějíček, Jiří; You, J.H.

2017-01-01

Roč. 13, December (2017), s. 74-80 ISSN 2352-1791 R&D Projects: GA ČR GB14-36566G EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : tungsten-fibre/tungsten composites * plasma-facing components * spark plasma sintering Subject RIV: JI - Composite Materials OBOR OECD: Composites (including laminates, reinforced plastics, cermets, combined natural and synthetic fibre fabrics https://www.sciencedirect.com/science/article/pii/S2352179117300273

Tungsten oxide coatings deposited by plasma spray using powder and solution precursor for detection of nitrogen dioxide gas

Energy Technology Data Exchange (ETDEWEB)

Zhang, Chao, E-mail: zhangc@yzu.edu.cn [College of Mechanical Engineering, Yangzhou University, Yangzhou 225127 (China); Wang, Jie [College of Mechanical Engineering, Yangzhou University, Yangzhou 225127 (China); Geng, Xin [College of Mechanical Engineering, Yangzhou University, Yangzhou 225127 (China); College of Chemistry and Chemical Engineering, Yangzhou University, Yangzhou 225002 (China)

2016-05-25

Increasing attention has been paid on preparation methods for resistive-type gas sensors based on semiconductor metal oxides. In this work, tungsten oxide (WO{sub 3}) coatings were prepared on alumina substrates and used as gas sensitive layers. The coatings were deposited by atmospheric plasma spray using powder, solution precursor, or a combination of both. Tungsten oxide powder through a powder port and ammonium tungstate aqueous solution through a liquid port were injected into plasma stream respectively or together to deposit WO{sub 3} coatings. Phase structures in the coatings were characterized by X-ray diffraction analyzer. The field-emission scanning electron microscopy images confirmed that the coatings were in microstructure, nanostructure or micro-nanostructure. The sensing properties of the sensors based on the coatings exposed to 1 ppm nitrogen dioxide gas were characterized in a home-made instrument. Sensing properties of the coatings were compared and discussed. The influences of gas humidity and working temperature on the sensor responses were further studied. - Highlights: • Porous gas sensitive coatings were deposited by plasma spray using powder and solution precursor. • Crystallized WO{sub 3} were obtained through hybrid plasma spray plus a pre-conditioned step. • Plasma power had an important influence on coating microstructure. • The particle size of atmospheric plasma-sprayed microstructured coating was stable. • Solution precursor plasma-sprayed WO{sub 3} coatings had nanostructure and showed good responses to 1 ppm NO{sub 2}.

Helium effects on tungsten surface morphology and deuterium retention

International Nuclear Information System (INIS)

Ueda, Y.; Peng, H.Y.; Lee, H.T.; Ohno, N.; Kajita, S.; Yoshida, N.; Doerner, R.; De Temmerman, G.; Alimov, V.; Wright, G.

2013-01-01

Recent experimental results on tungsten surface morphology, especially nano-structure (fuzz), induced by helium plasma exposure at temperatures between 1000 K and 2000 K are reviewed. This structure was firstly reported in 2006. In this review, most of experimental results reported so far including characteristics and formation conditions of the nano-structure in both linear plasma devices and magnetic confinement devices, erosion and arcing by steady-state plasma exposure and ELM-like pulsed heat or pulsed plasma exposure by a laser and a plasma gun are summarized. In addition, He effects on D retention under simultaneous D/He irradiation on tungsten are presented

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

Science.gov (United States)

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.

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

Energy Technology Data Exchange (ETDEWEB)

Klimov, N.S., E-mail: klimov@triniti.ru [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye shosse 31, Moscow 115409 (Russian Federation); Putrik, A.B. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Linke, J. [Forschungszentrum Jülich GmbH, EURATOM Association, Jülich D-52425 (Germany); Pitts, R.A. [Karlsruhe Institute of Technology, P.O. Box 3640, Karlsruhe 76021 (Germany); Zhitlukhin, A.M. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Kuprianov, I.B. [Bochvar Institute, ul. Rogova, 5a, Moscow 123098 (Russian Federation); Spitsyn, A.V. [NRC «Kurchatov Institute», Akademika Kurchatova pl., 1, Moscow 123182 (Russian Federation); Ogorodnikova, O.V. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye shosse 31, Moscow 115409 (Russian Federation); Podkovyrov, V.L.; Muzichenko, A.D. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Ivanov, B.V.; Sergeecheva, Ya.V.; Lesina, I.G. [Bochvar Institute, ul. Rogova, 5a, Moscow 123098 (Russian Federation); Kovalenko, D.V.; Barsuk, V.A.; Danilina, N.A. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Bazylev, B.N. [Karlsruhe Institute of Technology, P.O. Box 3640, Karlsruhe 76021 (Germany); Giniyatulin, R.N. [Efremov Institute, Doroga na Metallostroy, 3 bld., Metallostroy, Saint-Petersburg 196641 (Russian Federation)

2015-08-15

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.

Microscopic Motion of Liquid Metal Plasma Facing Components In A Diverted Plasma

International Nuclear Information System (INIS)

Jaworski, M.A.; Gerhardt, S.P.; Morley, N.B.; Abrams, T.; Kaita, R.; Kallman, J.; Kugel, H.; Majeski, R.; Ruzic, D.N.

2010-01-01

Liquid metal plasma facing components (PFCs) have been identified as an alternative material for fusion plasma experiments. The use of a liquid conductor where significant magnetic fields are present is considered risky, with the possibility of macroscopic fluid motion and possible ejection into the plasma core. Analysis is carried out on thermoelectric magnetohydrodynamic (TEMHD) forces caused by temperature gradients in the liquid-container system itself in addition to scrape-off-layer currents interacting with the PFC from a diverted plasma. Capillary effects at the liquid-container interface will be examined which govern droplet ejection criteria. Stability of the interface is determined using linear stability methods. In addition to application to liquidmetal PFCs, thin film liquidmetal effects have application to current and future devices where off-normal events may liquefy portions of the first wall and other plasma facing components.

Gas-driven permeation of deuterium through tungsten and tungsten alloys

Energy Technology Data Exchange (ETDEWEB)

Buchenauer, Dean A., E-mail: dabuche@sandia.gov [Sandia National Laboratories, Energy Innovation Department, Livermore, CA 94550 (United States); Karnesky, Richard A. [Sandia National Laboratories, Energy Innovation Department, Livermore, CA 94550 (United States); Fang, Zhigang Zak; Ren, Chai [University of Utah, Department of Metallurgical Engineering, Salt Lake City, UT 84112 (United States); Oya, Yasuhisa [Shizuoka University, Graduate School of Science, Shizuoka (Japan); Otsuka, Teppei [Kyushu University, Department of Advanced Energy Engineering Science, Fukuoka (Japan); Yamauchi, Yuji [Hokkaido University, Third Division of Quantum Science and Engineering, Faculty of Engineering, Sapporo (Japan); Whaley, Josh A. [Sandia National Laboratories, Energy Innovation Department, Livermore, CA 94550 (United States)

2016-11-01

Highlights: • We have designed and performed initial studies on a high temperature gas-driven permeation cell capable of operating at temperatures up to 1150 °C and at pressures between 0.1–1 atm. • Permeation measurements on ITER grade tungsten compare well with past studies by Frauenfelder and Zahkarov in the temperature range from 500 to 1000 °C. • First permeation measurements on Ti dispersoid-strengthened ultra-fine grained tungsten show higher permeation at 500 °C, but very similar permeation with ITER tungsten at 1000 °C. Diffusion along grain boundaries may be playing a role for this type of material. - Abstract: To address the transport and trapping of hydrogen isotopes, several permeation experiments are being pursued at both Sandia National Laboratories (deuterium gas-driven permeation) and Idaho National Laboratories (tritium gas- and plasma-driven tritium permeation). These experiments are in part a collaboration between the US and Japan to study the performance of tungsten at divertor relevant temperatures (PHENIX). Here we report on the development of a high temperature (≤1150 °C) gas-driven permeation cell and initial measurements of deuterium permeation in several types of tungsten: high purity tungsten foil, ITER-grade tungsten (grains oriented through the membrane), and dispersoid-strengthened ultra-fine grain (UFG) tungsten being developed in the US. Experiments were performed at 500–1000 °C and 0.1–1.0 atm D{sub 2} pressure. Permeation through ITER-grade tungsten was similar to earlier W experiments by Frauenfelder (1968–69) and Zaharakov (1973). Data from the UFG alloy indicates marginally higher permeability (< 10×) at lower temperatures, but the permeability converges to that of the ITER tungsten at 1000 °C. The permeation cell uses only ceramic and graphite materials in the hot zone to reduce the possibility for oxidation of the sample membrane. Sealing pressure is applied externally, thereby allowing for elevation

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

Energy Technology Data Exchange (ETDEWEB)

Garrison, L. M., E-mail: garrisonlm@ornl.gov; Egle, B. J. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, Tennessee 37831 (United States); Fusion Technology Institute, University of Wisconsin-Madison, 1500 Engineering Drive, Madison, Wisconsin 53706 (United States); Zenobia, S. J.; Kulcinski, G. L.; Santarius, J. F. [Fusion Technology Institute, University of Wisconsin-Madison, 1500 Engineering Drive, Madison, Wisconsin 53706 (United States)

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 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{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.

2D surface temperature measurement of plasma facing components with modulated active pyrometry

International Nuclear Information System (INIS)

Amiel, S.; Loarer, T.; Pocheau, C.; Roche, H.; Gauthier, E.; Aumeunier, M.-H.; Courtois, X.; Jouve, M.; Balorin, C.; Moncada, V.; Le Niliot, C.; Rigollet, F.

2014-01-01

In nuclear fusion devices, such as Tore Supra, the plasma facing components (PFC) are in carbon. Such components are exposed to very high heat flux and the surface temperature measurement is mandatory for the safety of the device and also for efficient plasma scenario development. Besides this measurement is essential to evaluate these heat fluxes for a better knowledge of the physics of plasma-wall interaction, it is also required to monitor the fatigue of PFCs. Infrared system (IR) is used to manage to measure surface temperature in real time. For carbon PFCs, the emissivity is high and known (ε ∼ 0.8), therefore the contribution of the reflected flux from environment and collected by the IR cameras can be neglected. However, the future tokamaks such as WEST and ITER will be equipped with PFCs in metal (W and Be/W, respectively) with low and variable emissivities (ε ∼ 0.1–0.4). Consequently, the reflected flux will contribute significantly in the collected flux by IR camera. The modulated active pyrometry, using a bicolor camera, proposed in this paper allows a 2D surface temperature measurement independently of the reflected fluxes and the emissivity. Experimental results with Tungsten sample are reported and compared with simultaneous measurement performed with classical pyrometry (monochromatic and bichromatic) with and without reflective flux demonstrating the efficiency of this method for surface temperature measurement independently of the reflected flux and the emissivity

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

Science.gov (United States)

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.

Development of a plasma assisted ITER level controlled heat source and observation of novel micro/nanostructures produced upon exposure of tungsten targets

Energy Technology Data Exchange (ETDEWEB)

Aomoa, N.; Sarmah, Trinayan; Sah, Puspalata [CIMPLE-PSI Laboratory, Centre of Plasma Physics-Institute for Plasma Research, Sonapur 782 402 Assam (India); Chaudhuri, P.; Khirwarker, S.; Ghosh, J. [Institute for Plasma Research, Gandhinagar 382428 Gujarat (India); Satpati, B. [Saha Institute of Nuclear Physics, 1/AF Bidhannagar, Kolkata 700 064 (India); Kakati, M., E-mail: mayurkak@rediffmail.com [CIMPLE-PSI Laboratory, Centre of Plasma Physics-Institute for Plasma Research, Sonapur 782 402 Assam (India); De Temmerman, G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046 Saint Paul Lez Durance, Cedex (France)

2016-05-15

Highlights: • Developed a plasma assisted ITER level high heat flux device for material testing. • The beam deposits over 10 MW/m{sup 2} flux uniformly over a remote material target. • Hopper micro-crystals were growing while exposing Plansee tungsten in the device. • CIMPLE-PSI being developed for exact reproduction of Tokomak Divertor conditions. - Abstract: This paper reports on the development of a simple, low-cost, segmented plasma torch assisted high-heat flux device for material testing, which can simulate the extreme heat flux expected in future fusion devices. Calorimetric measurements confirmed uniform heat deposition by the well collimated argon plasma beam over a target surface with power fluxes in excess of 10 MW/m{sup 2} during high current, high gas flow rate operations. To understand the outcome of possible melting of first wall material in an ITER like machine, an Plansee tungsten target was exposed in this device, which witnessed growth of micrometer level Hopper crystals and their aggregation to vertical grains in central exposed region. Increase in viscosity of the metal during high under-cooling is believed to have lead to the skeletal patterns, observed for the first time for tungsten here. Transmission electron microscopy confirmed that re-solidified grains on the target actually had crystalline substructures in the nanometer level. This laboratory is in the process of developing an exact linear Tokamak Divertor simulator, where a magnetized hydrogen/helium collimated plasma jet will be produced at higher vacuum, for plasma material interaction studies with direct relevance to modern plasma fusion machines.

Demonstration of production of tungsten metal powder and its consolidation into shapes

International Nuclear Information System (INIS)

Majumdar, S.; Kishor, J.; Paul, B.; Kain, V.; Dey, G.K.

2016-01-01

Tungsten is a strategically important metal used as plasma facing component in fusion reactors, radiation shields in cancer therapy machines, ammunition in defence applications, high speed cutting tools etc. The primary resources or minerals occurring in India contain a very low value (0.25-0.5 wt. %) of tungsten. Mineral beneficiation processes involving crushing, grinding, primary and secondary gravity separation, floatation are essential to produce the ore-concentrate suitable for further processing up to the preparation of the intermediate ammonium para-tungstate (APT). APT was further converted to tungsten tri-oxide (WO_3). Hydrogen reduction of WO_3 producing high purity W metal powder was demonstrated in large scale batches. Densification of W powder was further studied using vacuum hot pressing at 1950°C, and high density W metal plates of 5 mm thickness and 60 mm diameter were produced. The products obtained at every stage were systematically characterized using X-Ray diffraction (XRD), scanning electron microscope (SEM), energy dispersive spectroscopy (EDS) and electron backscattered diffraction (EBSD) techniques. (author)

The heat removal capability of actively cooled plasma-facing components for the ITER divertor

Science.gov (United States)

Missirlian, M.; Richou, M.; Riccardi, B.; Gavila, P.; Loarer, T.; Constans, S.

2011-12-01

Non-destructive examination followed by high-heat-flux testing was performed for different small- and medium-scale mock-ups; this included the most recent developments related to actively cooled tungsten (W) or carbon fibre composite (CFC) armoured plasma-facing components. In particular, the heat-removal capability of these mock-ups manufactured by European companies with all the main features of the ITER divertor design was investigated both after manufacturing and after thermal cycling up to 20 MW m-2. Compliance with ITER requirements was explored in terms of bonding quality, heat flux performances and operational compatibility. The main results show an overall good heat-removal capability after the manufacturing process independent of the armour-to-heat sink bonding technology and promising behaviour with respect to thermal fatigue lifetime under heat flux up to 20 MW m-2 for the CFC-armoured tiles and 15 MW m-2 for the W-armoured tiles, respectively.

The heat removal capability of actively cooled plasma-facing components for the ITER divertor

International Nuclear Information System (INIS)

Missirlian, M; Richou, M; Loarer, T; Riccardi, B; Gavila, P; Constans, S

2011-01-01

Non-destructive examination followed by high-heat-flux testing was performed for different small- and medium-scale mock-ups; this included the most recent developments related to actively cooled tungsten (W) or carbon fibre composite (CFC) armoured plasma-facing components. In particular, the heat-removal capability of these mock-ups manufactured by European companies with all the main features of the ITER divertor design was investigated both after manufacturing and after thermal cycling up to 20 MW m - 2. Compliance with ITER requirements was explored in terms of bonding quality, heat flux performances and operational compatibility. The main results show an overall good heat-removal capability after the manufacturing process independent of the armour-to-heat sink bonding technology and promising behaviour with respect to thermal fatigue lifetime under heat flux up to 20 MW m - 2 for the CFC-armoured tiles and 15 MW m - 2 for the W-armoured tiles, respectively.

Pre-qualification of brazed plasma facing components of divertor target elements for ITER like tokamak application

International Nuclear Information System (INIS)

Singh, K.P.; Pandya, Santosh P.; Khirwadkar, S.S.; Patel, Alpesh; Patil, Y.; Buch, J.J.U.; Khan, M.S.; Tripathi, Sudhir; Pandya, Shwetang; Govindrajan, J.; Jaman, P.M.; Rathore, Devendra; Rangaraj, L.; Divakar, C.

2011-01-01

Qualification of tungsten (W) and graphite (C) based brazed plasma facing components (PFCs) is an important R and D area in fusion research. Pre-qualification tests for brazed joints between W-CuCrZr and C-CuCrZr using NDT (IR thermography and ultrasonic test) and thermal fatigue test are attempted. Mockups having good quality brazed joints of W and C based PFCs were identified using NDT. Subsequently, thermal fatigue test was performed on the identified mockups. All brazed tiles of W based PFC mockups could withstand thermal fatigue test, however, few tiles of C based PFC mockup were found detached. Thermal analyses of mockups are performed using finite element analysis (ANSYS) software to simulate the thermal hydraulic condition with 10 MW/m 2 uniform heat flux. Details about experimental and computational work are presented here.

Pre-qualification of brazed plasma facing components of divertor target elements for ITER like tokamak application

Energy Technology Data Exchange (ETDEWEB)

Singh, K.P., E-mail: kpsingh@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar, Gujarat (India); Pandya, Santosh P.; Khirwadkar, S.S.; Patel, Alpesh; Patil, Y.; Buch, J.J.U.; Khan, M.S.; Tripathi, Sudhir; Pandya, Shwetang; Govindrajan, J. [Institute for Plasma Research, Bhat, Gandhinagar, Gujarat (India); Jaman, P.M.; Rathore, Devendra; Rangaraj, L.; Divakar, C. [Materials Science Division, National Aerospace Laboratories, CSIR, Bangalore, Karnataka (India)

2011-10-15

Qualification of tungsten (W) and graphite (C) based brazed plasma facing components (PFCs) is an important R and D area in fusion research. Pre-qualification tests for brazed joints between W-CuCrZr and C-CuCrZr using NDT (IR thermography and ultrasonic test) and thermal fatigue test are attempted. Mockups having good quality brazed joints of W and C based PFCs were identified using NDT. Subsequently, thermal fatigue test was performed on the identified mockups. All brazed tiles of W based PFC mockups could withstand thermal fatigue test, however, few tiles of C based PFC mockup were found detached. Thermal analyses of mockups are performed using finite element analysis (ANSYS) software to simulate the thermal hydraulic condition with 10 MW/m{sup 2} uniform heat flux. Details about experimental and computational work are presented here.

Molecular-dynamics analysis of mobile helium cluster reactions near surfaces of plasma-exposed tungsten

Energy Technology Data Exchange (ETDEWEB)

Hu, Lin; Maroudas, Dimitrios, E-mail: maroudas@ecs.umass.edu [Department of Chemical Engineering, University of Massachusetts, Amherst, Massachusetts 01003-9303 (United States); Hammond, Karl D. [Department of Chemical Engineering, University of Missouri, Columbia, Missouri 65211 (United States); Wirth, Brian D. [Department of Nuclear Engineering, University of Tennessee, Knoxville, Tennessee 37996 (United States)

2015-10-28

We report the results of a systematic atomic-scale analysis of the reactions of small mobile helium clusters (He{sub n}, 4 ≤ n ≤ 7) near low-Miller-index tungsten (W) surfaces, aiming at a fundamental understanding of the near-surface dynamics of helium-carrying species in plasma-exposed tungsten. These small mobile helium clusters are attracted to the surface and migrate to the surface by Fickian diffusion and drift due to the thermodynamic driving force for surface segregation. As the clusters migrate toward the surface, trap mutation (TM) and cluster dissociation reactions are activated at rates higher than in the bulk. TM produces W adatoms and immobile complexes of helium clusters surrounding W vacancies located within the lattice planes at a short distance from the surface. These reactions are identified and characterized in detail based on the analysis of a large number of molecular-dynamics trajectories for each such mobile cluster near W(100), W(110), and W(111) surfaces. TM is found to be the dominant cluster reaction for all cluster and surface combinations, except for the He{sub 4} and He{sub 5} clusters near W(100) where cluster partial dissociation following TM dominates. We find that there exists a critical cluster size, n = 4 near W(100) and W(111) and n = 5 near W(110), beyond which the formation of multiple W adatoms and vacancies in the TM reactions is observed. The identified cluster reactions are responsible for important structural, morphological, and compositional features in the plasma-exposed tungsten, including surface adatom populations, near-surface immobile helium-vacancy complexes, and retained helium content, which are expected to influence the amount of hydrogen re-cycling and tritium retention in fusion tokamaks.

Bonding tungsten, W–Cu-alloy and copper with amorphous Fe–W alloy transition

Energy Technology Data Exchange (ETDEWEB)

Wang, Song, E-mail: wangsongrain@163.com [Laboratory of Special Ceramics and Powder Metallurgy, University of Science and Technology Beijing, Beijing 100083 (China); Laboratory of Advanced Materials, Tsinghua University, Beijing 100084 (China); Ling, Yunhan, E-mail: yhling@mail.tsinghua.edu.cn [Laboratory of Advanced Materials, Tsinghua University, Beijing 100084 (China); Zhao, Pei [Technical Institute of Physics and Chemistry, Chinese Academy of Sciences, Beijing 100190 (China); Graduate School of the Chinese Academy of Sciences, Beijing 100039 (China); Zang, Nanzhi [Laboratory of Advanced Materials, Tsinghua University, Beijing 100084 (China); Wang, Jianjun [Laboratory of Special Ceramics and Powder Metallurgy, University of Science and Technology Beijing, Beijing 100083 (China); Guo, Shibin [Technical Institute of Physics and Chemistry, Chinese Academy of Sciences, Beijing 100190 (China); Graduate School of the Chinese Academy of Sciences, Beijing 100039 (China); Zhang, Jun [Laboratory of Advanced Materials, Tsinghua University, Beijing 100084 (China); Xu, Guiying [Laboratory of Special Ceramics and Powder Metallurgy, University of Science and Technology Beijing, Beijing 100083 (China)

2013-05-15

W/Cu graded materials are the leading candidate materials used as the plasma facing components in a fusion reactor. However, tungsten and copper can hardly be jointed together due to their great differences in physical properties such as coefficient of thermal expansion and melting point, and the lack of solid solubility between them. To overcome those difficulties, a new amorphous Fe–W alloy transitional coating and vacuum hot pressing (VHP) method were proposed and introduced in this paper. The morphology, composition and structure of the amorphous Fe–W alloy coating and the sintering interface of the specimens were analyzed by scanning electron microscopy (SEM), energy dispersive spectrometer (EDS) and X-ray diffraction (XRD). The thermal shock resistance of the bonded composite was also tested. The results demonstrated that amorphous structure underwent change from amorphous to nano grains during joining process, and the joined W/Cu composite can endued plasma thermal shock resistance with energy density more than 5.33 MW/m{sup 2}. It provides a new feasible technical to join refractory tungsten to immiscible copper with amorphous Fe–W alloy coating.

Simulation of tungsten erosion and transport near the divertor plate during ELMs by a kinetic method

Energy Technology Data Exchange (ETDEWEB)

Sun, Zhenyue; Sang, Chaofeng; Hu, Wanpeng; Du, Hailong; Wang, Dezhen, E-mail: wangdez@dlut.edu.cn

2016-11-01

Highlights: • A kinetic method is used to simulate tungsten erosion and transport during ELMs. • The erosion of tungsten plate by different species (deuterium and carbon ions) is shown. • The charge states of sputtered tungsten particles are given statistically. - Abstract: Tungsten (W) is fore seen as one of the most important candidates of the plasma-facing materials (PFM) for future fusion devices, due to its beneficial properties. However, the high-Z characteristic makes it a potential contamination to the core plasma. Divertor is the main component that directly contacts the plasma, therefore, it is very important to understand the erosion of W divertor plate and the corresponding transport of the eroded wall impurity, especially during edge localized modes (ELMs). In this work, a one-dimension-in-space and three-dimensions-in-velocity particle-in-cell code (EPPIC1D) is used to simulate the erosion of W divertor plate, and the transport of eroded W impurity near the divertor plate is studied by a Monte Carlo code. Benefiting from the kinetic simulation, energy/particle flux to the target could be calculated accurately, and the erosion of W plate by different species is simulated during ELMs. The trajectories and distributions of eroded W impurity particles are demonstrated, which shows us a basic idea of how these impurity particles are generated and transported. It is found that C{sup 3+} plays a dominated role on the erosion of W divertor plate during ELMs even when its concentration is low. Both W atoms and ions distribute mainly near the divertor plate, indicating only a very small fraction of W impurity particles could escape from divertor region and penetrate into the core plasma.

Plasma facing device of thermonuclear device

International Nuclear Information System (INIS)

Sumita, Hideo; Ioki, Kimihiro.

1993-01-01

The present invention improves integrity of thermal structures of a plasma facing device. That is, in the plasma facing device, an armour block portion from a metal cooling pipe to a carbon material comprises a mixed material of the metal as the constituent material of the cooling pipe and ceramics. Then, the mixing ratio of the composition is changed continuously or stepwise to suppress peakings of remaining stresses upon production and thermal stresses upon exertion of thermal loads. Accordingly, thermal integrity of the structural materials can further be improved. In this case, a satisfactory characteristic can be obtained also by using ceramics instead of carbon for the mixed material, and the characteristic such as heat expansion coefficient is similar to that of the armour tile. (I.S.)

The formation of tungsten doped Al_2O_3/ZnO coatings on aluminum by plasma electrolytic oxidation and their application in photocatalysis

International Nuclear Information System (INIS)

Stojadinović, Stevan; Vasilić, Rastko; Radić, Nenad; Tadić, Nenad; Stefanov, Plamen; Grbić, Boško

2016-01-01

Highlights: • Tungsten doped Al_2O_3/ZnO coatings are formed by plasma electrolytic oxidation (PEO). • Coatings are mainly composed of alpha alumina, ZnO and metallic tungsten. • Photocatalytic activity of doped Al_2O_3/ZnO coatings is higher than of undoped ones. • The increase of photoluminescence corresponds to decrease of photocatalytic activity. • Tungsten acts as a charge trap to reduce the recombination rate of electron/hole pairs. - Abstract: Tungsten doped Al_2O_3/ZnO coatings are formed by plasma electrolytic oxidation of aluminum substrate in supporting electrolyte (0.1 M boric acid + 0.05 M borax + 2 g/L ZnO) with addition of different concentrations of Na_2WO_4·2H_2O. The morphology, crystal structure, chemical composition, and light absorption characteristics of formed surface coatings are investigated. The X-ray diffraction and X-ray photoelectron spectroscopy results indicate that formed surface coatings consist of alpha and gamma phase of Al_2O_3, ZnO, metallic tungsten and WO_3. Obtained results showed that incorporated tungsten does not have any influence on the absorption spectra of Al_2O_3/ZnO coatings, which showed invariable band edge at about 385 nm. The photocatalytic activity of undoped and tungsten doped Al_2O_3/ZnO coatings is estimated by the photodegradation of methyl orange. The photocatalytic activity of tungsten doped Al_2O_3/ZnO coatings is higher thanof undoped Al_2O_3/ZnO coatings; the best photocatalytic activity is ascribed to coatings formed in supporting electrolyte with addition of 0.3 g/L Na_2WO_4·2H_2O. Tungsten in Al_2O_3/ZnO coatings acts as a charge trap, thus reducing the recombination rate of photogenerated electron-hole pairs. The results of PL measurements are in agreement with photocatalytic activity. Declining PL intensity corresponds to increasing photocatalytic activity of the coatings, indicating slower recombination of electron-hole pairs.

Technical design of a solid tungsten divertor row for the ITER-like wall in the JET tokamak

Energy Technology Data Exchange (ETDEWEB)

Mertens, P.; Knaup, M.; Neubauer, O.; Sadakov, S.; Schweer, B.; Terra, A.; Samm, U. [Forschungszentrum Juelich, Association EURATOM-FZJ (DE). Inst. fuer Energieforschung IEF-4 (Plasmaphysik); Pintsuk, G. [Forschungszentrum Juelich, Association EURATOM-FZJ (DE). Inst. fuer Energieforschung IEF-2 (Werkstoffstruktur und Eigenschaften)

2009-07-01

ITER (originally International Thermonuclear Experimental Reactor) is now under construction in Cadarache, France. In order to investigate plasma scenarios compatible with an ITER relevant mix of materials, a new, complete inner wall will be installed in the JET tokamak vessel (Culham, UK) in 2010. The plasmafacing components in the main chamber will be made of beryllium whereas the exposed areas in the divertor shall be made of tungsten, mostly of tungsten coatings on a carbon-fibre composite substrate. A notable exception is the central row of tiles where the outer strike point is located. Fig. 1 illustrates it with a camera view during a suitable discharge which shows the emission of atomic hydrogen, hence the main interaction regions. Plasma-facing components at this position are exposed to very high particle fluxes which cause material sputtering, and to extremely high heat loads without active cooling, which is not available. It was accordingly decided to resort to solid tungsten in this particular case. An overview of the conceptual design was presented earlier. Manufacturing is just starting, so the technical design has been frozen to the largest extent as presented in the following. (orig.)

Temperature Dependent Surface Modification of Tungsten Exposed to High-Flux Low-Energy Helium Ion Irradiation

OpenAIRE

Damico, Antony Q; Tripathi, Jitendra K; Novakowski, Theodore J; Miloshevsky, Gennady; Hassanein, Ahmed

2016-01-01

Nuclear fusion is a great potential energy source that can provide a relatively safe and clean limitless supply of energy using hydrogen isotopes as fuel material. ITER (international thermonuclear experimental reactor) is the world first fusion reactor currently being built in France. Tungsten (W) is a prime candidate material as plasma facing component (PFC) due to its excellent mechanical properties, high melting point, and low erosion rate. However, W undergoes a severe surface morphology...

Simulation of residual thermostress in tungsten after repetitive ELM-like heat loads

Energy Technology Data Exchange (ETDEWEB)

Pestchanyi, S., E-mail: serguei.pestchanyi@kit.edu [Karlsruhe Institute of Technology, IHM (Germany); Garkusha, I. [Institute of Plasma Physics of the NSC KIPT, Kharkov (Ukraine); Landman, I. [Karlsruhe Institute of Technology, IHM (Germany)

2011-10-15

Brittle destruction of tungsten armour under action of edge localised modes of plasma instabilities (ELMs) in ITER is an important issue determining the lifetime of the divertor. Besides, cracking of the armour produces tungsten dust with characteristic size of 1-10 {mu}m flying from the armour surface with velocities up to 10 m/s. Influx of the tungsten dust into the ITER confinement decreases the temperature of the plasma, reduces the thermonuclear gain and even may run the confinement into disruption. This paper describes experiments in QSPA-Kh50 plasma gun and modeling, which has been performed for providing more insight into the physics of tungsten cracking under action of ELMs and for confirmation of the important result on stabilization of the crack development at the tungsten armour surface, predicted in our previous paper - the same authors, 2010. The threshold value of energy density deposition for start of tungsten cracking has been measured as 0.3 MJ/m{sup 2} after 5-10 shots. From analytical considerations three times smaller threshold value has been predicted with increasing number of shots.

Simulation of residual thermostress in tungsten after repetitive ELM-like heat loads

International Nuclear Information System (INIS)

Pestchanyi, S.; Garkusha, I.; Landman, I.

2011-01-01

Brittle destruction of tungsten armour under action of edge localised modes of plasma instabilities (ELMs) in ITER is an important issue determining the lifetime of the divertor. Besides, cracking of the armour produces tungsten dust with characteristic size of 1-10 μm flying from the armour surface with velocities up to 10 m/s. Influx of the tungsten dust into the ITER confinement decreases the temperature of the plasma, reduces the thermonuclear gain and even may run the confinement into disruption. This paper describes experiments in QSPA-Kh50 plasma gun and modeling, which has been performed for providing more insight into the physics of tungsten cracking under action of ELMs and for confirmation of the important result on stabilization of the crack development at the tungsten armour surface, predicted in our previous paper - the same authors, 2010. The threshold value of energy density deposition for start of tungsten cracking has been measured as 0.3 MJ/m 2 after 5-10 shots. From analytical considerations three times smaller threshold value has been predicted with increasing number of shots.

Steady state and transient thermal-hydraulic characterization of full-scale ITER divertor plasma facing components

International Nuclear Information System (INIS)

Tincani, A.; Malavasi, A.; Ricapito, I.; Riccardi, B.; Di Maio, P.A.; Vella, G.

2007-01-01

In the frame of the activities related to ITER divertor R and D, ENEA CR Brasimone was charged by EFDA (European Fusion Design Agreement) to investigate the thermal-hydraulic behaviour of the full-scale divertor plasma facing components, i.e. Inner Vertical Target, Dome Liner and Outer Vertical Target, both in steady state and during draining and drying transient. More in detail, for each PFC, the first phase of the work is the steady state hydraulic characterization which consists of: - measurements of pressure drops at different temperatures; - determination of the velocity distribution in the internal channels; - check the possible insurgence of cavitation. The subsequent phase of the thermal-hydraulic characterization foresees a testing campaign of draining and drying procedure by means of a suitable gas flow. The objective of this experimental procedure is to eliminate in the most efficient way the residual amount of water after gravity discharge. In order to accomplish this experimental campaign a significant modification of CEF1 loop has been designed and realized. This paper presents, first of all, the experimental set-up, the agreed test matrix and the achieved results for both steady state and transient tests. Moreover, the level of the implementation of a predictive hydraulic model, based on RELAP 5 code, as well as its results are described, discussed and compared with the experimental ones. (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

Alfonso, A., E-mail: aalz@dtu.dk [Section of Materials and Surface Engineering, Department of Mechanical Engineering, Technical University of Denmark, 2800 Lyngby (Denmark); Sino-Danish Center for Education and Research (China); Sino-Danish Center for Education and Research (Denmark); Juul Jensen, D. [Danish-Chinese Center for Nanometals, Section of Materials Science and Advanced Characterization, Department of Wind Energy, Technical University of Denmark, Risø Campus, 4000 Roskilde (Denmark); Sino-Danish Center for Education and Research (China); Sino-Danish Center for Education and Research (Denmark); Luo, G.-N. [Fusion Reactor Materials Science and Technology Division, Institute of Plasma Physics, Chinese Academy of Sciences, 230031 Hefei, Anhui (China); Sino-Danish Center for Education and Research (China); Sino-Danish Center for Education and Research (Denmark); Pantleon, W. [Section of Materials and Surface Engineering, Department of Mechanical Engineering, Technical University of Denmark, 2800 Lyngby (Denmark); Association EURATOM-DTU (Denmark); Sino-Danish Center for Education and Research (China); Sino-Danish Center for Education and Research (Denmark)

2014-12-15

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.

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Hydrogen release from 800 MeV proton-irradiated tungsten

Science.gov (United States)

Oliver, B. M.; Venhaus, T. J.; Causey, R. A.; Garner, F. A.; Maloy, S. A.

2002-12-01

Tungsten irradiated in spallation neutron sources, such as those proposed for the accelerator production of tritium (APT) project, will contain large quantities of generated helium and hydrogen gas. Tungsten used in proposed fusion reactors will also be exposed to neutrons, and the generated protium will be accompanied by deuterium and tritium diffusing in from the plasma-facing surface. The release kinetics of these gases during various off-normal scenarios involving loss of coolant and after heat-induced rises in temperature are of particular interest for both applications. To determine the release kinetics of hydrogen from tungsten, tungsten rods irradiated with 800 MeV protons in the Los Alamos Neutron Science Center (LANSCE) to high exposures as part of the APT project have been examined. Hydrogen evolution from the tungsten has been measured using a dedicated mass-spectrometer system by subjecting the specimens to an essentially linear temperature ramp from ˜300 to ˜1500 K. Release profiles are compared with predictions obtained using the Tritium Migration Analysis Program (TMAP4). The measurements show that for high proton doses, the majority of the hydrogen is released gradually, starting at about 900 K and reaching a maximum at about 1400 K, where it drops fairly rapidly. Comparisons with TMAP show quite reasonable agreement using a trap energy of 1.4 eV and a trap density of ˜7%. There is a small additional release fraction occurring at ˜550 K, which is believed to be associated with low-energy trapping at or near the surface, and, therefore, was not included in the bulk TMAP model.

Hydrogen release from 800 MeV proton-irradiated tungsten

International Nuclear Information System (INIS)

Oliver, B.M.; Venhaus, T.J.; Causey, R.A.; Garner, F.A.; Maloy, S.A.

2002-01-01

Tungsten irradiated in spallation neutron sources, such as those proposed for the accelerator production of tritium (APT) project, will contain large quantities of generated helium and hydrogen gas. Tungsten used in proposed fusion reactors will also be exposed to neutrons, and the generated protium will be accompanied by deuterium and tritium diffusing in from the plasma-facing surface. The release kinetics of these gases during various off-normal scenarios involving loss of coolant and after heat-induced rises in temperature are of particular interest for both applications. To determine the release kinetics of hydrogen from tungsten, tungsten rods irradiated with 800 MeV protons in the Los Alamos Neutron Science Center (LANSCE) to high exposures as part of the APT project have been examined. Hydrogen evolution from the tungsten has been measured using a dedicated mass-spectrometer system by subjecting the specimens to an essentially linear temperature ramp from ∼300 to ∼1500 K. Release profiles are compared with predictions obtained using the Tritium Migration Analysis Program (TMAP4). The measurements show that for high proton doses, the majority of the hydrogen is released gradually, starting at about 900 K and reaching a maximum at about 1400 K, where it drops fairly rapidly. Comparisons with TMAP show quite reasonable agreement using a trap energy of 1.4 eV and a trap density of ∼7%. There is a small additional release fraction occurring at ∼550 K, which is believed to be associated with low-energy trapping at or near the surface, and, therefore, was not included in the bulk TMAP model

Hydrogen release from 800 MeV proton-irradiated tungsten

Energy Technology Data Exchange (ETDEWEB)

Oliver, B.M. E-mail: brian.oliver@pnl.gov; Venhaus, T.J.; Causey, R.A.; Garner, F.A.; Maloy, S.A

2002-12-01

Tungsten irradiated in spallation neutron sources, such as those proposed for the accelerator production of tritium (APT) project, will contain large quantities of generated helium and hydrogen gas. Tungsten used in proposed fusion reactors will also be exposed to neutrons, and the generated protium will be accompanied by deuterium and tritium diffusing in from the plasma-facing surface. The release kinetics of these gases during various off-normal scenarios involving loss of coolant and after heat-induced rises in temperature are of particular interest for both applications. To determine the release kinetics of hydrogen from tungsten, tungsten rods irradiated with 800 MeV protons in the Los Alamos Neutron Science Center (LANSCE) to high exposures as part of the APT project have been examined. Hydrogen evolution from the tungsten has been measured using a dedicated mass-spectrometer system by subjecting the specimens to an essentially linear temperature ramp from {approx}300 to {approx}1500 K. Release profiles are compared with predictions obtained using the Tritium Migration Analysis Program (TMAP4). The measurements show that for high proton doses, the majority of the hydrogen is released gradually, starting at about 900 K and reaching a maximum at about 1400 K, where it drops fairly rapidly. Comparisons with TMAP show quite reasonable agreement using a trap energy of 1.4 eV and a trap density of {approx}7%. There is a small additional release fraction occurring at {approx}550 K, which is believed to be associated with low-energy trapping at or near the surface, and, therefore, was not included in the bulk TMAP model.

Blister/hole formation on tungsten surface due to low-energy and high-flux deuterium/helium plasma exposures

International Nuclear Information System (INIS)

Nishijima, D.; Iwakiri, H.; Yoshida, N.; Ye, M.Y.; Ohno, N.; Takamura, S.

2005-01-01

Deuterium/helium plasma exposures on tungsten surface bring serious damages such as blister and hole. Blistering occurs by cleaving along layered structure intrinsic to the press-roll manufacturing process. Mechanical polishing and helium pre-exposure on mirror-finished powder metallurgy tungsten drastically suppress blister formation. Small cracks made by a polishing would become paths to the surface for diffusing deuterium atoms in the substrate, resulting in no gas accumulation and no blister formation on the surface. Helium pre-exposure would make a helium-enriched layer near the surface, which becomes a kind of diffusion barrier for incident deuterium atoms. Blister formation and deuterium retention are suppressed on the surface with helium-enriched layer. (author)

Surface modifications on toughened, fine-grained, recrystallized tungsten with repetitive ELM-like pulsed plasma irradiation

Energy Technology Data Exchange (ETDEWEB)

Kikuchi, Y., E-mail: ykikuchi@eng.u-hyogo.ac.jp [Graduate School of Engineering, University of Hyogo, 2167 Shosha, Himeji, 671-2280 Hyogo (Japan); Sakuma, I.; Kitagawa, Y.; Asai, Y.; Onishi, K.; Fukumoto, N.; Nagata, M. [Graduate School of Engineering, University of Hyogo, 2167 Shosha, Himeji, 671-2280 Hyogo (Japan); Ueda, Y. [Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Kurishita, H. [Institute for Materials Research, Tohoku University, Ibaraki 311-1313 (Japan)

2015-08-15

Surface modifications of toughened, fine-grained, recrystallized tungsten (TFGR W) materials with 1.1 wt.% TiC and 3.3 wt.% TaC dispersoids due to repetitive ELM-like pulsed (∼0.15 ms) helium plasma irradiation have been investigated by using a magnetized coaxial plasma gun. No surface cracking at the center part of the TFGR W samples exposed to 20 plasma pulses of ∼0.3 MJ m{sup −2} was observed. The suppression of surface crack formation due to the increase of the grain boundary strength by addition of TiC and TaC dispersoids was confirmed in comparison with a pure W material. On the other hand, surface cracks and small pits appeared at the edge part of the TFGR W sample after the pulsed plasma irradiation. Erosion of the TiC and TaC dispersoids due to the pulsed plasma irradiation could cause the small pits on the surface, resulting in the surface crack formation.

Surface modifications on toughened, fine-grained, recrystallized tungsten with repetitive ELM-like pulsed plasma irradiation

Science.gov (United States)

Kikuchi, Y.; Sakuma, I.; Kitagawa, Y.; Asai, Y.; Onishi, K.; Fukumoto, N.; Nagata, M.; Ueda, Y.; Kurishita, H.

2015-08-01

Surface modifications of toughened, fine-grained, recrystallized tungsten (TFGR W) materials with 1.1 wt.% TiC and 3.3 wt.% TaC dispersoids due to repetitive ELM-like pulsed (∼0.15 ms) helium plasma irradiation have been investigated by using a magnetized coaxial plasma gun. No surface cracking at the center part of the TFGR W samples exposed to 20 plasma pulses of ∼0.3 MJ m-2 was observed. The suppression of surface crack formation due to the increase of the grain boundary strength by addition of TiC and TaC dispersoids was confirmed in comparison with a pure W material. On the other hand, surface cracks and small pits appeared at the edge part of the TFGR W sample after the pulsed plasma irradiation. Erosion of the TiC and TaC dispersoids due to the pulsed plasma irradiation could cause the small pits on the surface, resulting in the surface crack formation.

Experimental study of parallel multi-tungsten wire Z-pinch

International Nuclear Information System (INIS)

Huang Xianbin; China Academy of Engineering Physics, Mianyang; Lin Libin; Yang Libing; Deng Jianjun; Gu Yuanchao; Ye Shican; Yue Zhengpu; Zhou Shaotong; Li Fengping; Zhang Siqun

2005-01-01

The study of three parallel tungsten wire loads and five parallel tungsten wire loads implosion experiment on accelerator 'Yang' are reported. Tungsten wires (φ17 μm) with separation of 1 mm were used. The pinch was driven by a 350 kA peak current, 80 ns 10%-90% rise time. By means of pinhole camera and X-ray diagnostics technology, a non-uniform plasma column is formed among the wires and soft X-ray pulse are observed. the change of load current are analyzed, the development of sausage instability and kink instability, 'hot spot' effect and dispersion spot for plasma column are also discussed. (authors)

Final steps to an all tungsten divertor tokamak

International Nuclear Information System (INIS)

Neu, R.; Bobkov, V.; Dux, R.; Kallenbach, A.; Puetterich, Th.; Greuner, H.; Gruber, O.; Herrmann, A.; Hopf, Ch.; Krieger, K.; Maggi, C.F.; Maier, H.; Mayer, M.; Rohde, V.; Schmid, K.; Suttrop, W.

2007-01-01

Currently 85% of the plasma facing components of ASDEX Upgrade are tungsten coated. Carbon influx from W PFCs is still observed but a reduction of the C content is found in plasma discharges and a lower C fraction is measured in deposited layers in agreement with modelling. W sputtering from the low field side guard and ICRF limiters is mainly due to fast particles from NBI as well as from ions accelerated in the rectified sheath during ICRF operation. The increase of the W source area is reflected in increased W concentrations. For medium to high density discharges the techniques developed so far, namely central heating and ELM pace-making, allow keeping the W concentration in the range of 10 -5 . Boronisation strongly reduces the W influxes and similarly the W content especially during ICRF operation, but this reduction is only temporary and equilibrium is reached already after about 100 discharges

Impact and effects of simultaneous MeV-ion irradiation and helium plasma exposure to the formation of tungsten nano-tendrils

Science.gov (United States)

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.

ALPS - advanced limiter-divertor plasma-facing systems

International Nuclear Information System (INIS)

Allain, J. P.; Bastasz, R.; Brooks, J. N.; Evans, T.; Hassanein, A.; Luckhardt, S.; Maingi, R.; Mattas, R. F.; McCarthy, K.; Mioduszewski, P.; Mogahed, E.; Moir, R.; Molokov, S.; Morely, N.; Nygren, R.; Reed, C.; Rognlien, T.; Ruzic, D.; Sviatoslavsky, I.; Sze, D.; Tillack, M.; Ulrickson, M.; Wade, P. M.; Wong, C.; Wooley, R.

1999-01-01

The Advanced Limiter-divertor Plasma-facing Systems (ALPS) program was initiated in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is to demonstrate the advantages of advanced limiter/divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma. Most of the work to date has been applied to free surface liquids. A multi-disciplinary team from several institutions has been organized to address the key issues associated with these systems. The main performance goals for advanced limiters and diverters are a peak heat flux of >50 MW/m 2 ,elimination of a lifetime limit for erosion, and the ability to extract useful heat at high power conversion efficiency (approximately40%). The evaluation of various options is being conducted through a combination of laboratory experiments, modeling of key processes, and conceptual design studies. The current emphasis for the work is on the effects of free surface liquids on plasma edge performance

Study on the oxidation and reduction of tungsten surface for sub-50 nm patterning process

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Kyu; Nam, Seok Woo; Cho, Sung Il; Jhon, Myung S.; Min, Kyung Suk; Kim, Chan Kyu; Jung, Ho Bum; Yeom, Geun Young [Memory Division Semiconductor Business, Samsung Electronics, San No. 16 Banwol-Ri, Taean-Eup, Hwasung-City, Gyeonggi-Do 449-711, South Korea and Department of Advanced Materials Science and Engineering, Sungkyunkwan University, Suwon, Gyeonggi-do 440-746 (Korea, Republic of); Memory Division Semiconductor Business, Samsung Electronics, San No. 16 Banwol-Ri, Taean-Eup, Hwasung-City, Gyeonggi-Do 449-711 (Korea, Republic of); Department of Chemical Engineering and Data Storage Systems Center, Carnegie Mellon University, Pittsburgh, Pennsylvania 15213 (United States); Department of Advanced Materials Science and Engineering, Sungkyunkwan University, Suwon, Gyeonggi-do 440-746 (Korea, Republic of)

2012-11-15

The oxidation characteristics of tungsten line pattern during the carbon-based mask-layer removal process using oxygen plasmas have been investigated for sub-50 nm patterning processes, in addition to the reduction characteristics of the WO{sub x} layer formed on the tungsten line surface using hydrogen plasmas. The surface oxidation of tungsten lines during the mask layer removal process could be minimized by using low-temperature (300 K) plasma processing for the removal of the carbon-based material. Using this technique, the thickness of WO{sub x} on the tungsten line could be decreased to 25% compared to results from high-temperature processing. The WO{sub x} layer could also be completely removed at a low temperature of 300 K using a hydrogen plasma by supplying bias power to the tungsten substrate to provide a activation energy for the reduction. When this oxidation and reduction technique was applied to actual 40-nm-CD device processing, the complete removal of WO{sub x} formed on the sidewall of tungsten line could be observed.

Environmental fate of tungsten from military use

International Nuclear Information System (INIS)

Clausen, Jay L.; Korte, Nic

2009-01-01

This manuscript describes the distribution, fate and transport of tungsten used in training rounds at three small arms ranges at Camp Edwards on the Massachusetts Military Reservation (MMR), USA. Practice with tungsten/nylon rounds began in 2000 subsequent to a 1997 US Environmental Protection Agency ban on training with lead. Training with the tungsten rounds was halted in 2005 because of concerns regarding tungsten's environmental mobility and potential toxicity. This study, therefore, examines how tungsten partitions in the environment when fired on a small arms training range. Soil sampling revealed surface soil concentrations, highest at the berm face, up to 2080 mg/kg. Concentrations decreased rapidly with depth-at least by an order of magnitude by 25 cm. Nonetheless, tungsten concentrations remained above background to at least 150 cm. Pore-water samples from lysimeters installed in berm areas revealed a range of concentrations (< 1-400 mg/L) elevated with respect to background although there was no discernable trend with depth. Groundwater monitoring well samples collected approximately 30 m below ground surface showed tungsten (0.001-0.56 mg/L) attributable to range use

Suppression of tungsten accumulation during ELMy H-mode by lower hybrid wave heating in the EAST tokamak

Directory of Open Access Journals (Sweden)

L. Zhang

2017-08-01

Full Text Available EAST tokamak has been equipped with upper tungsten divertor since 2014. The tungsten accumulation has been often observed in NBI-heated H-mode discharges suggesting deleterious tungsten confinement in the plasma core. It causes not only H-L back transition but also plasma disruption in several discharges. Suppression of the tungsten accumulation is therefore the most important issue in EAST to achieve a long pulse H-mode discharge. In order to study the tungsten behavior in the long pulse discharge, tungsten spectra have been measured at 20–140Å. The tungsten density, nw, is evaluated from the intensity of tungsten unresolved transition array (W-UTA in a wavelength range of 45–70Å which is composed of several ionization stages of tungsten, e.g. W27+-W45+ at Te0∼2.5keV. It is found that the tungsten accumulation can be suppressed when the 4.6GHz LHW with PLHW∼0.8MW is superimposed on the NBI phase (PNBI= 1.9MW. During the superimposed phase the ELM frequency, fELM, increases from ∼30Hz to ∼60Hz and the tungsten density is halved compared to the NBI-heated discharge. The H-mode discharge can be thus steadily sustained for longer period. It is found that the nw is a large function of the ratio of LHW power to the total injection power, PLHW/(PLHW+PNBI, and the nw can be reduced, at least, in an order of magnitude smaller than that in NBI-heated discharges at PLHW/(PLHW+PNBI≥0.8. The result strongly suggests a possible way toward the steady H-mode discharge.

Tungsten impurity transport experiments in Alcator C-Mod to address high priority research and development for ITER

Energy Technology Data Exchange (ETDEWEB)

Loarte, A.; Polevoi, A. R.; Hosokawa, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Reinke, M. L. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Chilenski, M.; Howard, N.; Hubbard, A.; Hughes, J. W.; Rice, J. E.; Walk, J. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Köchl, F. [Technische Universität Wien, Atominstitut, Stadionallee 2, 1020 Vienna (Austria); Pütterich, T.; Dux, R. [Max-Planck-Institut für Plasmaphysik, Boltzmanstraße 2, D-85748 Garching (Germany); Zhogolev, V. E. [NRC “Kurchatov Institute,” Kurchatov Square 1, 123098 Moscow (Russian Federation)

2015-05-15

Experiments in Alcator C-Mod tokamak plasmas in the Enhanced D-alpha H-mode regime with ITER-like mid-radius plasma density peaking and Ion Cyclotron Resonant heating, in which tungsten is introduced by the laser blow-off technique, have demonstrated that accumulation of tungsten in the central region of the plasma does not take place in these conditions. The measurements obtained are consistent with anomalous transport dominating tungsten transport except in the central region of the plasma where tungsten transport is neoclassical, as previously observed in other devices with dominant neutral beam injection heating, such as JET and ASDEX Upgrade. In contrast to such results, however, the measured scale lengths for plasma temperature and density in the central region of these Alcator C-Mod plasmas, with density profiles relatively flat in the core region due to the lack of core fuelling, are favourable to prevent inter and intra sawtooth tungsten accumulation in this region under dominance of neoclassical transport. Simulations of ITER H-mode plasmas, including both anomalous (modelled by the Gyro-Landau-Fluid code GLF23) and neoclassical transport for main ions and tungsten and with density profiles of similar peaking to those obtained in Alcator C-Mod show that accumulation of tungsten in the central plasma region is also unlikely to occur in stationary ITER H-mode plasmas due to the low fuelling source by the neutral beam injection (injection energy ∼ 1 MeV), which is in good agreement with findings in the Alcator C-Mod experiments.

Plasma facing components design of KT-2 tokamak

International Nuclear Information System (INIS)

In, Sang Ryul; Yoon, Byung Joo; Song, Woo Soeb; Xu, Chao Yin

1997-04-01

The vacuum vessel of KT-2 tokamak is protected from high thermal loads by various kinds of plasma facing components (PFC): outer and inner divertors, neutral baffle, inboard limiter, poloidal limiter, movable limiter and passive plate, installed on the inner wall of the vessel. In this report the pre-engineering design of the plasma facing components, including design requirements and function, structures of PFC assemblies, configuration of cooling systems, calculations of some mechanical and hydraulic parameters, is presented. Pumping systems for the movable limiter and the divertor are also discussed briefly. (author). 49 figs

X-ray absorption spectroscopy of aluminum z-pinch plasma with tungsten backlighter planar wire array source

Energy Technology Data Exchange (ETDEWEB)

Osborne, G. C.; Kantsyrev, V. L.; Safronova, A. S.; Esaulov, A. A.; Weller, M. E.; Shrestha, I.; Shlyaptseva, V. V. [Physics Department, University of Nevada, Reno, Reno, Nevada 89557 (United States); Ouart, N. D. [Naval Research Laboratory, Washington, D.C. 20375 (United States)

2012-10-15

Absorption features from K-shell aluminum z-pinch plasmas have recently been studied on Zebra, the 1.7 MA pulse power generator at the Nevada Terawatt Facility. In particular, tungsten plasma has been used as a semi-backlighter source in the generation of aluminum K-shell absorption spectra by placing a single Al wire at or near the end of a single planar W array. All spectroscopic experimental results were recorded using a time-integrated, spatially resolved convex potassium hydrogen phthalate (KAP) crystal spectrometer. Other diagnostics used to study these plasmas included x-ray detectors, optical imaging, laser shadowgraphy, and time-gated and time-integrated x-ray pinhole imagers. Through comparisons with previous publications, Al K-shell absorption lines are shown to be from much lower electron temperature ({approx}10-40 eV) plasmas than emission spectra ({approx}350-500 eV).

Low flux and low energy helium ion implantation into tungsten using a dedicated plasma source

Energy Technology Data Exchange (ETDEWEB)

Pentecoste, Lucile [GREMI, CNRS/Université d’Orléans, 14 rue d’Issoudun, B.P. 6744, 45067 Orléans Cedex2 (France); Thomann, Anne-Lise, E-mail: anne-lise.thomann@univ-orleans.fr [GREMI, CNRS/Université d’Orléans, 14 rue d’Issoudun, B.P. 6744, 45067 Orléans Cedex2 (France); Melhem, Amer; Caillard, Amael; Cuynet, Stéphane; Lecas, Thomas; Brault, Pascal [GREMI, CNRS/Université d’Orléans, 14 rue d’Issoudun, B.P. 6744, 45067 Orléans Cedex2 (France); Desgardin, Pierre; Barthe, Marie-France [CNRS, UPR3079 CEMHTI, 1D avenue de la Recherche Scientifique, 45071 Orléans Cedex2 (France)

2016-09-15

The aim of this work is to investigate the first stages of defect formation in tungsten (W) due to the accumulation of helium (He) atoms inside the crystal lattice. To reach the required implantation conditions, i.e. low He ion fluxes (10{sup 11}–10{sup 14} ions.cm{sup 2}.s{sup −1}) and kinetic energies below the W atom displacement threshold (about 500 eV for He{sup +}), an ICP source has been designed and connected to a diffusion chamber. Implantation conditions have been characterized by means of complementary diagnostics modified for measurements in this very low density helium plasma. It was shown that lowest ion fluxes could only be reached for the discharge working in capacitive mode either in α or γ regime. Special attention was paid to control the energy gained by the ions by acceleration through the sheath at the direct current biased substrate. At very low helium pressure, in α regime, a broad ion energy distribution function was evidenced, whereas a peak centered on the potential difference between the plasma and the biased substrate was found at higher pressures in the γ mode. Polycrystalline tungsten samples were exposed to the helium plasma in both regimes of the discharge and characterized by positron annihilation spectroscopy in order to detect the formed vacancy defects. It was found that W vacancies are able to be formed just by helium accumulation and that the same final implanted state is reached, whatever the operating mode of the capacitive discharge.

Properties of tungsten coating deposited onto copper by high-speed atmospheric plasma spraying

Energy Technology Data Exchange (ETDEWEB)

Huang Jianjun, E-mail: huangjj@szu.edu.cn [Applied Low Temperature Plasma Laboratory, College of Physics Science and Technology, Shenzhen University, Shenzhen 518060 (China); Wang Fan; Liu Ying; Jiang Shishou; Wang Xisheng; Qi Bing; Gao Liang [Applied Low Temperature Plasma Laboratory, College of Physics Science and Technology, Shenzhen University, Shenzhen 518060 (China)

2011-07-01

Tungsten (W) coatings were fabricated on copper (Cu) by high-speed atmospheric plasma spray (HAPS) technique. The properties of the porosity, oxygen content, bonding strength and microhardness were measured. The results obtained indicated that the HAPS-W coating showed good properties particularly in terms of porosity and oxygen content. The porosity of the HAPS-W coating was 2.3% and the distribution of pore size diameter was mainly concentrated in the range of 0.01-1 {mu}m. The oxygen content of the coating measured by means of Nitrogen/Oxygen Determinator was about 0.10 wt.%. These initial results suggest that the HAPS-W coating has achieved the reported properties of the vacuum plasma spray (VPS) W coating. Compared with VPS, HAPS-W technique could provide a convenient and low cost way to obtain adequate W coatings for fusion applications.

Electron Impact Excitation and Dielectronic Recombination of Highly Charged Tungsten Ions

Directory of Open Access Journals (Sweden)

Zhongwen Wu

2015-11-01

Full Text Available Electron impact excitation (EIE and dielectronic recombination (DR of tungsten ions are basic atomic processes in nuclear fusion plasmas of the International Thermonuclear Experimental Reactor (ITER tokamak. Detailed investigation of such processes is essential for modeling and diagnosing future fusion experiments performed on the ITER. In the present work, we studied total and partial electron-impact excitation (EIE and DR cross-sections of highly charged tungsten ions by using the multiconfiguration Dirac–Fock method. The degrees of linear polarization of the subsequent X-ray emissions from unequally-populated magnetic sub-levels of these ions were estimated. It is found that the degrees of linear polarization of the same transition lines, but populated respectively by the EIE and DR processes, are very different, which makes diagnosis of the formation mechanism of X-ray emissions possible. In addition, with the help of the flexible atomic code on the basis of the relativistic configuration interaction method, DR rate coefficients of highly charged W37+ to W46+ ions are also studied, because of the importance in the ionization equilibrium of tungsten plasmas under running conditions of the ITER.

Flaw detection device for plasma facing wall in thermonuclear device

International Nuclear Information System (INIS)

Doi, Akira.

1996-01-01

The present invention concerns plasma facing walls of a thermonuclear device and provides a device for detecting a thickness of amour tiles accurately and efficiently with no manual operation. Namely, the position of the plasma facing surface of the amour tile is measured using a structure to which the amour tiles are to be disposed as a reference. Also in a case of disposing new armor tiles, the position of the plasma facing surface of the armor tiles is measured to thereby measure the wearing amount of the amour tiles based on the difference between the reference and the measured value. If a measuring means capable of measuring a plurality of amour tiles at once is used efficiency of the measurement and the detection can be enhanced. Several ten thousands of amour tiles are disposed to the plasma facing wall in a large scaled thermonuclear device, and a plenty of time was required for the detection. However, the present invention can improve the accuracy for the measurement and detection and provide time and labors-saving. (I.S.)

Atomic and molecular processes in fusion plasmas

International Nuclear Information System (INIS)

Kato, Daiji; Nakamura, Nobuyuki

2013-01-01

One of important issues concerning steady state sustainment of magnetically confined plasmas (MCPs) is distribution of impurity ions in the MCPs and radiation powers by the ions. Since tungsten divertors will be used in ITER, the primary element of heavy impurity ions would be tungsten. Tungsten cannot be fully ionized even in core plasmas of ITER. Line radiations by bound electrons of the tungsten ions following electron impact excitations decrease temperatures of the core plasmas. Thus, it is required to keep tungsten concentration in the core plasmas as small as possible. (J.P.N.)

The effects of tantalum addition on the microtexture and mechanical behaviour of tungsten for ITER applications

International Nuclear Information System (INIS)

Tejado, E.; Carvalho, P.A.; Munoz, A.; Dias, M.; Correia, J.B.

2015-01-01

Tungsten (W) and its alloys are very promising materials for producing plasma-facing components (PFCs) in the fusion power reactors of the near future, even as a structural part in them. However, whereas the properties of pure tungsten are suitable for a PFC, its structural applications are still limited due to its low toughness, ductile to brittle transition temperature and recrystallization behaviour. Therefore, many efforts have been made to improve its performance by alloying tungsten with other elements. Hence, in this investigation, the thermo-mechanical performance of two new tungsten-tantalum materials has been evaluated. Materials with W–5wt.%Ta and W–15wt.%Ta were processed by mechanical alloying (MA) and later consolidation by hot isostatic pressing (HIP), with distinct settings for each composition. Thus, it was possible to determine the relationship between the microstructure and the addition of Ta with the macroscopic mechanical properties. These were measured by means of hardness, flexural strength and fracture toughness, in the temperature range of 300–1473 K. The microstructure and the fracture surfaces features of the tested materials were analysed by Field Emission Scanning Electron Microscopy (FESEM).

The effects of tantalum addition on the microtexture and mechanical behaviour of tungsten for ITER applications

Energy Technology Data Exchange (ETDEWEB)

Tejado, E., E-mail: elena.tejado@mater.upm.es [Departamento de Ciencia de Materiales-CIME, ETSI Caminos, Canales y Puertos, Universidad Politécnica de Madrid, Madrid (Spain); Centro Nacional de Investigaciones Metalúrgicas (CSIC), Madrid (Spain); Carvalho, P.A. [Associação Euratom/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); ICEMS, Departamento de Bioengenharia, Instituto Superior Técnico, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); Munoz, A. [Departamento de Física, Universidad Carlos III, Leganés (Spain); Dias, M. [Associação Euratom/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); Correia, J.B. [Associação Euratom/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); LNEG, Laboratório Nacional de Energia e Geologia, Estrada do Paço do Lumiar, 1649-038 Lisboa (Portugal); and others

2015-12-15

Tungsten (W) and its alloys are very promising materials for producing plasma-facing components (PFCs) in the fusion power reactors of the near future, even as a structural part in them. However, whereas the properties of pure tungsten are suitable for a PFC, its structural applications are still limited due to its low toughness, ductile to brittle transition temperature and recrystallization behaviour. Therefore, many efforts have been made to improve its performance by alloying tungsten with other elements. Hence, in this investigation, the thermo-mechanical performance of two new tungsten-tantalum materials has been evaluated. Materials with W–5wt.%Ta and W–15wt.%Ta were processed by mechanical alloying (MA) and later consolidation by hot isostatic pressing (HIP), with distinct settings for each composition. Thus, it was possible to determine the relationship between the microstructure and the addition of Ta with the macroscopic mechanical properties. These were measured by means of hardness, flexural strength and fracture toughness, in the temperature range of 300–1473 K. The microstructure and the fracture surfaces features of the tested materials were analysed by Field Emission Scanning Electron Microscopy (FESEM).

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

DEFF Research Database (Denmark)

Alfonso Lopez, Angel; Juul Jensen, Dorte; Luo, G.-N.

2014-01-01

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...... 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...

Hydrogen Release From 800-MeV Proton-Irradiated Tungsten

International Nuclear Information System (INIS)

Oliver, Brian M.; Venhaus, Thomas J.; Causey, Rion A.; Garner, Francis A.; Maloy, Stuart A.

2002-01-01

Tungsten irradiated in spallation neutron sources such as those proposed for the Accelerator Production of Tritium (APT) project, or in proposed fusion reactors, will contain large quantities of generated helium and hydrogen gas. In the APT, spallation neutrons would be generated by the interaction of high energy (∼1 GeV) protons with solid tungsten rods or cylinders. In fusion reactors, tungsten used in a tokamak diverter will contain hydrogen, as well as deuterium and tritium diffusing in from the plasma-facing surface. The release kinetics of these gases during various off-normal scenarios involving loss of coolant and afterheat-induced rises in temperature is of particular interest for both applications. To determine the release kinetics of hydrogen from tungsten, tungsten rods irradiated with 800 MeV protons in the Los Alamos Neutron Science Center (LANCE) to high exposures as part of the APT project have been examined. Hydrogen evolution from the tungsten was measured using a dedicated mass spectrometer system by subjecting the specimens to an essentially linear temperature ramp from ∼323 K to ∼1473 K. Release profiles are compared with predictions obtained using the Tritium Migration Analysis Program (TMAP4). Input parameters for the modeling, consisting of diffusivity, recombination rate coefficient, and trapping, are discussed. The measurements show that for high proton doses, the majority of the hydrogen is released gradually, starting at about 900 K and reaching a maximum at about 1400 K, where it drops fairly rapidly. Comparisons with TMAP show reasonable agreement at high proton dose using a trap value of 1.4 eV and a trap density of 3%. There is also a small release fraction occurring at ∼600 K which predominates at lower proton doses, and which is relatively independent of dose. This lower-temperature release is predicted by TMAP if no traps are assumed, suggesting that this release may represent an adsorbed surface component

Deuterium desorption from tungsten using laser heating

Directory of Open Access Journals (Sweden)

J.H. Yu

2017-08-01

Full Text Available Retention and desorption of hydrogenic species need to be accurately modeled to predict the tritium inventory of next generation fusion devices, which is needed both for tritium fuel recovery and for tritium safety concerns. In this paper, experiments on thermal desorption of deuterium from intrinsic polycrystalline tungsten defects using laser heating are compared to TMAP-7 modeling. The samples during deuterium plasma exposure were at a temperature of 373K for this benchmark study with ion fluence of 0.7–1.0 ×1024Dm−2. Following plasma exposure, a fiber laser (λ= 1100nm heated the samples to peak surface temperatures ranging from ∼500 to 1400K with pulse widths from 10ms to 1s, and 1 to 10 pulses applied to each sample. The remaining deuterium retention was measured using temperature programmed desorption (TPD. Results show that > 95% of deuterium is desorbed when the peak surface temperature reached ∼950K for > 1s. TMAP-7 is used to predict deuterium desorption from tungsten for a range of surface temperatures and heating durations, and is compared to previous work on desorption from beryllium codeposits.

Collisional-Radiative Modeling of Tungsten at Temperatures of 1200–2400 eV

Directory of Open Access Journals (Sweden)

James Colgan

2015-04-01

Full Text Available We discuss new collisional-radiative modeling calculations of tungsten at moderate temperatures of 1200 to 2400 eV. Such plasma conditions are relevant to ongoing experimental work at ASDEX Upgrade and are expected to be relevant for ITER. Our calculations are made using the Los Alamos National Laboratory (LANL collisional-radiative modeling ATOMIC code. These calculations formed part of a submission to the recent NLTE-8 workshop that was held in November 2013. This series of workshops provides a forum for detailed comparison of plasma and spectral quantities from NLTE collisional-radiative modeling codes. We focus on the LANL ATOMIC calculations for tungsten that were submitted to the NLTE-8 workshop and discuss different models that were constructed to predict the tungsten emission. In particular, we discuss comparisons between semi-relativistic configuration-average and fully relativistic configuration-average calculations. We also present semi-relativistic calculations that include fine-structure detail, and discuss the difficult problem of ensuring completeness with respect to the number of configurations included in a CR calculation.

Experimental study of divertor plasma-facing components damage under a combination of pulsed and quasi-stationary heat loads relevant to expected transient events at ITER

International Nuclear Information System (INIS)

Klimov, N S; Podkovyrov, V L; Kovalenko, D V; Zhitlukhin, A M; Barsuk, V A; Mazul, I V; Giniyatulin, R N; Kuznetsov, V Ye; Riccardi, B; Loarte, A; Merola, M; Koidan, V S; Linke, J; Landman, I S; Pestchanyi, S E; Bazylev, B N

2011-01-01

This paper concerns the experimental study of damage of ITER divertor plasma-facing components (PFCs) under a combination of pulsed plasma heat loads (representative of controlled ITER type I edge-localized modes (ELMs)) and quasi-stationary heat loads (representative of the high heat flux (HHF) thermal fatigue expected during ITER normal operations and slow transient events). The PFC's tungsten armor damage under pulsed plasma exposure was driven by (i) the melt layer motion, which leads to bridges formation between neighboring tiles and (ii) the W brittle failure giving rise to a stable crack pattern on the exposed surface. The crack width reaches a saturation value that does not exceed some tens of micrometers after several hundreds of ELM-like pulses. HHF thermal fatigue tests have shown (i) a peeling-off of the re-solidified material due to its brittle failure and (ii) a significant widening (up to 10 times) of the cracks and the formation of additional cracks.

Elastic–plastic adhesive impacts of tungsten dust with metal surfaces in plasma environments

NARCIS (Netherlands)

Ratynskaia, S.; Tolias, P.; Shalpegin, A.; Vignitchouk, L.; de Angeli, M.; I. Bykov,; Bystrov, K.; Bardin, S.; Brochard, F.; Ripamonti, D.; N. den Harder,; De Temmerman, G.

2015-01-01

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

Thermal radiation characteristics and direct evidence of tungsten cooling on the way to nanostructure formation on its surface

Energy Technology Data Exchange (ETDEWEB)

Takamura, S., E-mail: takamura@aitech.ac.jp [Faculty of Engineering, Aichi Institute of Technology, Yakusa-cho, Toyota 470-0392 (Japan); Miyamoto, T. [Faculty of Engineering, Aichi Institute of Technology, Yakusa-cho, Toyota 470-0392 (Japan); Ohno, N. [Graduate School of Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan)

2013-07-15

The physical properties of tungsten with nanostructure on its surface are investigated focusing on the thermal radiation and cooling characteristics. First, direct evidence of substantial W surface cooling has been clearly shown with use of a very thin thermocouple inserted into W target, which solves an uncertainty associated with a radiation thermometer. Second, the above measurements of W surface temperature make it possible to estimate quantitatively the total emissivity from which we may evaluate the radiative power through the Stefan–Boltzmann equation, which is very important for mitigation evaluation of a serious plasma heat load to the plasma-facing component.

Thermal radiation characteristics and direct evidence of tungsten cooling on the way to nanostructure formation on its surface

International Nuclear Information System (INIS)

Takamura, S.; Miyamoto, T.; Ohno, N.

2013-01-01

The physical properties of tungsten with nanostructure on its surface are investigated focusing on the thermal radiation and cooling characteristics. First, direct evidence of substantial W surface cooling has been clearly shown with use of a very thin thermocouple inserted into W target, which solves an uncertainty associated with a radiation thermometer. Second, the above measurements of W surface temperature make it possible to estimate quantitatively the total emissivity from which we may evaluate the radiative power through the Stefan–Boltzmann equation, which is very important for mitigation evaluation of a serious plasma heat load to the plasma-facing component

Environmental fate of tungsten from military use

Energy Technology Data Exchange (ETDEWEB)

Clausen, Jay L. [Research and Development Center, Cold Regions Research and Engineering Laboratory, 72 Lyme Road, Hanover, New Hampshire, 03755 (United States)], E-mail: Jay.L.Clausen@erdc.usace.army.mil; Korte, Nic [1946 Clover Ct., Grand Junction, Colorado, 81506 (United States)

2009-04-01

This manuscript describes the distribution, fate and transport of tungsten used in training rounds at three small arms ranges at Camp Edwards on the Massachusetts Military Reservation (MMR), USA. Practice with tungsten/nylon rounds began in 2000 subsequent to a 1997 US Environmental Protection Agency ban on training with lead. Training with the tungsten rounds was halted in 2005 because of concerns regarding tungsten's environmental mobility and potential toxicity. This study, therefore, examines how tungsten partitions in the environment when fired on a small arms training range. Soil sampling revealed surface soil concentrations, highest at the berm face, up to 2080 mg/kg. Concentrations decreased rapidly with depth-at least by an order of magnitude by 25 cm. Nonetheless, tungsten concentrations remained above background to at least 150 cm. Pore-water samples from lysimeters installed in berm areas revealed a range of concentrations (< 1-400 mg/L) elevated with respect to background although there was no discernable trend with depth. Groundwater monitoring well samples collected approximately 30 m below ground surface showed tungsten (0.001-0.56 mg/L) attributable to range use.

Electro-deposition metallic tungsten coatings in a Na{sub 2}WO{sub 4}-WO{sub 3} melt on copper based alloy substrate

Energy Technology Data Exchange (ETDEWEB)

Liu, Y.H., E-mail: dreamerhong77@126.com [School of Materials Science and Engineering, University of Science and Technology Beijing, 30 Xueyuan Road, Haidian District, Beijing 100083 (China); Zhang, Y.C.; Liu, Q.Z.; Li, X.L.; Jiang, F. [School of Materials Science and Engineering, University of Science and Technology Beijing, 30 Xueyuan Road, Haidian District, Beijing 100083 (China)

2012-11-15

Highlights: Black-Right-Pointing-Pointer The tungsten coating (>1 mm) was obtained by electro-deposition method in molten salt. Black-Right-Pointing-Pointer Different thickness tungsten coatings were obtained by using different durations. Black-Right-Pointing-Pointer Good performance of coating was obtained when pulse parameters were modulated. - Abstract: The tungsten coating was prepared by electro-deposition technique on copper alloy substrate in a Na{sub 2}WO{sub 4}-WO{sub 3} melt. The coating's surface and cross-section morphologies as well as its impurities were investigated by XPS, SEM and line analysis. Various plating durations were investigated in order to obtain an optimal coating's thickness. The results demonstrated that the electro-deposited coating was compact, voidless, crackless and free from impurities. The tungsten coating's maximum Vickers hardness was measured to be 520 HV. The tungsten coating's minimum oxygen content was determined to be 0.018 wt%. Its maximum thickness was measured to be 1043.67 {mu}m when the duration of electrolysis was set to 100 h. The result of this study has demonstrated the feasibility of having thicker tungsten coatings on copper alloy substrates. These electrodeposited tungsten coatings can be potentially implemented as reliable armour for the medium heat flux plasma facing component (PFC).

Effect of disruptions on plasma-facing components

International Nuclear Information System (INIS)

Gilligan, J.G.; Bourham, M.A.; Tucker, E.C.

1995-01-01

Erosion of plasma-facing components during disruptions is a limiting factor in the design of large tokamaks like ITER. During a disruption, much of the stored thermal energy of the plasma will be dumped onto divertor plates, resulting in local heat fluxes, which may exceed 100 GW/m 2 over a period of about 0.1--1.0 msec. Melted and/or vaporized material is produced which is redistributed in the divertor region. Simulation of disruption damage is summarized from code results and from experimental exposure of materials to high heat-flux plasmas in plasma guns. In the US several codes have been used to predict both melt/vaporization and heat transfer on surfaces as well as energy and momentum transport in the vapor/plasma shield produced at the surface

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

Science.gov (United States)

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} in the bcc crystal system and the {111} 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.

Simulation of cracks in tungsten under ITER specific heat loads

International Nuclear Information System (INIS)

Peschany, S.

2006-01-01

The problem of high tritium retention in co-deposited carbon layers on the walls of ITER vacuum chamber motivates investigation of materials for the divertor armour others than carbon fibre composite (CFC). Tungsten is most probable material for CFC replacement as the divertor armour because of high vaporisation temperature and heat conductivity. In the modern ITER design tungsten is a reference material for the divertor cover, except for the separatrix strike point armoured with CFC. As divertor armour, tungsten should withstand severe heat loads at off-normal ITER events like disruptions, ELMs and vertical displacement events. Experiments on tungsten heating with plasma streams and e-beams have shown an intense crack formation at the surface of irradiated sample [ V.I. Tereshin, A.N. Bandura, O.V. Byrka et al. Repetitive plasma loads typical for ITER type-I ELMs: Simulation at QSPA Kh-50.PLASMA 2005. ed. By Sadowski M.J., AIP Conference Proceedings, American Institute of Physics, 2006, V 812, p. 128-135., J. Linke. Private communications.]. The reason for tungsten cracking under severe heat loads is thermo stress. It appears as due to temperature gradient in solid tungsten as in resolidified layer after cooling down. Both thermo stresses are of the same value, but the gradiental stress is compressive and the stress in the resolidified layer is tensile. The last one is most dangerous for crack formation and it was investigated in this work. The thermo stress in tungsten that develops during cooling from the melting temperature down to room temperature is ∼ 8-16 GPa. Tensile strength of tungsten is much lower, < 1 GPa at room temperature, and at high temperatures it drops at least for one order of magnitude. As a consequence, various cracks of different characteristic scales appear at the heated surface of the resolidified layer. For simulation of the cracks in tungsten the numeric code PEGASUS-3D [Pestchanyi and I. Landman. Improvement of the CFC structure to

The formation of tungsten doped Al{sub 2}O{sub 3}/ZnO coatings on aluminum by plasma electrolytic oxidation and their application in photocatalysis

Energy Technology Data Exchange (ETDEWEB)

Stojadinović, Stevan, E-mail: sstevan@ff.bg.ac.rs [University of Belgrade, Faculty of Physics, Studentski trg 12-16, 11000 Belgrade (Serbia); Vasilić, Rastko [University of Belgrade, Faculty of Physics, Studentski trg 12-16, 11000 Belgrade (Serbia); Radić, Nenad [University of Belgrade, Institute of Chemistry, Technology and Metallurgy, Department of Catalysis and Chemical Engineering, Njegoševa 12, 11000 Belgrade (Serbia); Tadić, Nenad [University of Belgrade, Faculty of Physics, Studentski trg 12-16, 11000 Belgrade (Serbia); Stefanov, Plamen [Institute of General and Inorganic Chemistry, Bulgarian Academy of Sciences, Acad. G. Bonchev Str., Block 11, 1113 Sofia (Bulgaria); Grbić, Boško [University of Belgrade, Institute of Chemistry, Technology and Metallurgy, Department of Catalysis and Chemical Engineering, Njegoševa 12, 11000 Belgrade (Serbia)

2016-07-30

Highlights: • Tungsten doped Al{sub 2}O{sub 3}/ZnO coatings are formed by plasma electrolytic oxidation (PEO). • Coatings are mainly composed of alpha alumina, ZnO and metallic tungsten. • Photocatalytic activity of doped Al{sub 2}O{sub 3}/ZnO coatings is higher than of undoped ones. • The increase of photoluminescence corresponds to decrease of photocatalytic activity. • Tungsten acts as a charge trap to reduce the recombination rate of electron/hole pairs. - Abstract: Tungsten doped Al{sub 2}O{sub 3}/ZnO coatings are formed by plasma electrolytic oxidation of aluminum substrate in supporting electrolyte (0.1 M boric acid + 0.05 M borax + 2 g/L ZnO) with addition of different concentrations of Na{sub 2}WO{sub 4}·2H{sub 2}O. The morphology, crystal structure, chemical composition, and light absorption characteristics of formed surface coatings are investigated. The X-ray diffraction and X-ray photoelectron spectroscopy results indicate that formed surface coatings consist of alpha and gamma phase of Al{sub 2}O{sub 3}, ZnO, metallic tungsten and WO{sub 3}. Obtained results showed that incorporated tungsten does not have any influence on the absorption spectra of Al{sub 2}O{sub 3}/ZnO coatings, which showed invariable band edge at about 385 nm. The photocatalytic activity of undoped and tungsten doped Al{sub 2}O{sub 3}/ZnO coatings is estimated by the photodegradation of methyl orange. The photocatalytic activity of tungsten doped Al{sub 2}O{sub 3}/ZnO coatings is higher thanof undoped Al{sub 2}O{sub 3}/ZnO coatings; the best photocatalytic activity is ascribed to coatings formed in supporting electrolyte with addition of 0.3 g/L Na{sub 2}WO{sub 4}·2H{sub 2}O. Tungsten in Al{sub 2}O{sub 3}/ZnO coatings acts as a charge trap, thus reducing the recombination rate of photogenerated electron-hole pairs. The results of PL measurements are in agreement with photocatalytic activity. Declining PL intensity corresponds to increasing photocatalytic activity of the

Nitrogen implantation in tungsten and migration in the fusion experiment ASDEX upgrade

International Nuclear Information System (INIS)

Meisl, Gerd Korbinian

2015-01-01

The implantation of nitrogen ions into tungsten was studied in laboratory experiments to understand the interaction of nitrogen containing fusion plasmas with tungsten walls. The resulting model of W-N interaction was tested by experiments in the tokamak ASDEX Upgrade. Using the measurements from these experiments as boundary condition, nitrogen transport and re-distribution in the plasma were modeled by self-consistent WallDYN-DIVIMP simulations.

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

Science.gov (United States)

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.

Thermal loads on tokamak plasma-facing components during normal operation and disruptions

International Nuclear Information System (INIS)

McGrath, R.T.

1990-01-01

Power loadings experienced by tokamak plasma-facing components during normal operation and during off-normal events are discussed. A model for power and particle flow in the tokamak boundary layer is presented and model predictions are compared to infrared measurements of component heating. The inclusion of the full three-dimensional geometry of the components and of the magnetic flux surface is very important in the modeling. Experimental measurements show that misalignment of component armour tile surfaces by only a millimeter can lead to significant localized heating. An application to the design of plasma-facing components for future machines is presented. Finally, thermal loads expected during tokamak disruptions are discussed. The primary problems are surface melting and vaporization due to localized intense heating during the disruption thermal quench and volumetric heating of the component armour and structure due to localised impact of runaway electrons. (author)

Surface studies of tungsten erosion and deposition in JT-60U

International Nuclear Information System (INIS)

Ueda, Y.; Fukumoto, M.; Nishikawa, M.; Tanabe, T.; Miya, N.; Arai, T.; Masaki, K.; Ishimoto, Y.; Tsuzuki, K.; Asakura, N.

2007-01-01

In order to study tungsten erosion and migration in JT-60U, 13 W tiles have been installed in the outer divertor region and tungsten deposition on graphite tiles was measured. Dense local tungsten deposition was observed on a CFC tile toroidally adjacent to the W tiles, which resulted from prompt ionization and short range migration of tungsten along field lines. Tungsten deposition with relatively high surface density was found on an inner divertor tile around standard inner strike positions and on an outer wing tile of a dome. On the outer wing tile, tungsten deposition was relatively high compared with carbon deposition. In addition, roughly uniform tungsten depth distribution near the upper edge of the inner divertor tile was observed. This could be due to lift-up of strike point positions in selected 25 shots and tungsten flow in the SOL plasma

On tungsten technologies and qualification for DEMO

International Nuclear Information System (INIS)

Laan, J. van der; Hegeman, H.; Wouters, O.; Luzginova, N.; Jonker, B.; Van der Marck, S.; Opschoor, J.; Wang, J.; Dowling, G.; Stuivenga, M.; Carton, E.

2009-01-01

Tungsten alloys are considered prime candidates for the in-vessel components directly facing the plasma. For example, in the HEMJ helium cooled divertor design tiles may be operated at temperatures up to 1700 deg. C, supported by a structure partially consisting of tungsten at temperatures from 600 to 1000 deg. C, and connected to a HT steel structure. The tungsten armoured primary wall is operated at 500-900 deg. C. Irradiation doses will be few tens dpa at minimum, but FPR requirements for plants availability will stretch these targets. Recently injection moulding technology was developed for pure tungsten and representative parts were manufactured for ITER monobloc divertors and DEMO HEMJ thimbles. The major advantages for this technology are the efficient use of material feedstock/resources and the intrinsic possibility to produce near-finished product, avoiding machining processes that are costly and may introduce surface defects deteriorating the component in service performance. It is well suited for mass-manufacturing of components as well known in e.g. lighting industries. To further qualify this material technology various specimen types were produced with processing parameters identical to the components, and tested successfully, showing the high potential for implementation in (fusion) devices. Furthermore, the engineering approach can clearly be tailored away from conventional design and manufacturing technologies based on bulk materials. The technology is suitable for shaping of new W-alloys and W-ODS variants as well. Basically this technology allows a particular qualification trajectory. There is no need to produce large batches of material during the material development and optimization stage. For the verification of irradiation behaviour in the specific neutron spectra, there is a further attractive feature to use e.g. isotope tailored powders to adjust to available irradiation facilities like MTR's. In addition the ingrowth of transmutation

Ultrasonic techniques for quality assessment of ITER Divertor plasma facing component

International Nuclear Information System (INIS)

Martinez-Ona, Rafael; Garcia, Monica; Medrano, Mercedes

2009-01-01

The divertor is one of the most challenging components of ITER machine. Its plasma facing components contain thousands of joints that should be assessed to demonstrate their integrity during the required lifetime. Ultrasonic (US) techniques have been developed to study the capability of defect detection and to control the quality and degradation of these interfaces after the manufacturing process. Three types of joints made of carbon fibre composite to copper alloy, tungsten to copper alloy, and copper-to-copper alloy with two types of configurations have been studied. More than 100 samples representing these configurations and containing implanted flaws of different sizes have been examined. US techniques developed are detailed and results of validation samples examination before and after high heat flux (HHF) tests are presented. The results show that for W monoblocks the US technique is able to detect, locate and size the degradations in the two sample joints; for CFC monoblocks, the US technique is also able to detect, locate and size the calibrated defects in the two joints before the HHF, however after the HHF test the technique is not able to reliably detect defects in the CFC/Cu joint; finally, for the W flat tiles the US technique is able to detect, locate and size the calibrated defects in the two joints before HHF test, nevertheless defect location and sizing are more difficult after the HHF test.

Vacuum Plasma Spraying W-coated Reduced Activation Structural Steels for Fusion Plasma Facing Components

Energy Technology Data Exchange (ETDEWEB)

Noh, Sanghoon; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

Tungsten (W) and its alloys are considered as candidate materials for plasma facing materials of the first wall and diverter components in fusion reactor systems because of high sputtering resistance and low tritium retention in a fusion environment. Therefore, it is considered that the joining between W and reduced activation structural steels, and its evaluation, are critical issues for the development of fusion reactors. However, the joining between these materials is a very challenging process because of significant differences in their physical properties, particularly the mismatch of coefficients of thermal expansion (CTE). For instance, the CTE of pure W is known to be about 4.3Χ10{sup -6}K{sup -1}; however, that of martensitic steels reaches over three times, about 12-14Χ10{sup -6}K{sup -1} at room temperature even up to 373K. Nevertheless, several joining techniques have been developed for joining between W and structural steels, such as a vapor deposition method, brazing and diffusion bonding. Meanwhile, vacuum plasma spraying (VPS) is supposed to be one of the prospective methods to fabricate a sufficient W layer on the steel substrates because of the coating of a large area with a relatively high fabricating rate. In this study, the VPS method of W powders on reduced activation steels was employed, and its microstructure and hardness distribution were investigated. ODS ferritic steels and F82H steel were coated by VPS-W, and the microstructure and hardness distribution were investigated. A microstructure analysis revealed that pure W was successfully coated on steel substrates by the VPS process without an intermediate layer, in spite of a mismatch of the CTE between dissimilar materials. After neutron irradiation, irradiation hardening significantly occurred in the VPSW. However, the hardening of VPS-W was lesser than that of bulk W irradiated HFIR at 773K. Substrate materials, ODS ferritic steels, and F82H steel, did not show irradiation hardening

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

Science.gov (United States)

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.

Plasma thermal performance of a dual-process PVD/PS tungsten coating on carbon-based panels for nuclear fusion application

International Nuclear Information System (INIS)

Kim, Hyunmyung; Lee, Ho Jung; Kim, Sung Hwan; Jang, Changheui

2016-01-01

Highlights: • Plasma thermal performance of a dual-process PVD/PS W coating was evaluated. • Steady-state heat fluxes of 1–3 MW/m 2 were applied to the W coated specimens. • Less micro-pores and grain growth were observed for the dual-process coating. • Loss of coating thickness was observed for the simple PS W coating. • Dual-process PVD/PS W coating was resistant to erosion due to the surface PVD layer. - Abstract: Various tungsten (W) coating techniques have been used for the application of plasma facing material in nuclear fusion devices, which resulted in limited success. In this study, a dual-process W coating structure was developed on a graphite substrate to improve the thermal performance of the coating structure. The dual-process coating structure consisted of a thin (∼7 μm) multilayer W/Mo physical vapor deposition (PVD) coating layer deposited on top of the relatively thick (∼160 μm) plasma spray (PS) W coating on a graphite substrate panel. Then the coated sample was exposed to plasma heat flux of 1–3 MW/m 2 for 300 s. With addition of a thin surface PVD coating layer, the microstructure change in underlying PS W coating was substantially reduced compared to the simple PS W coating structure. The thickness of overall coating structure was maintained for the dual-process PVD/PS coated samples after the thermal loading tests, while a significant reduction in thickness due to surface erosion was observed for the simple PS W coated samples. The improvement in surface erosion resistance in the dual-process coating structure was discussed in view of the characteristics of PVD and PS coating layers.

Plasma thermal performance of a dual-process PVD/PS tungsten coating on carbon-based panels for nuclear fusion application

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyunmyung; Lee, Ho Jung; Kim, Sung Hwan; Jang, Changheui, E-mail: chjang@kaist.ac.kr

2016-11-01

Highlights: • Plasma thermal performance of a dual-process PVD/PS W coating was evaluated. • Steady-state heat fluxes of 1–3 MW/m{sup 2} were applied to the W coated specimens. • Less micro-pores and grain growth were observed for the dual-process coating. • Loss of coating thickness was observed for the simple PS W coating. • Dual-process PVD/PS W coating was resistant to erosion due to the surface PVD layer. - Abstract: Various tungsten (W) coating techniques have been used for the application of plasma facing material in nuclear fusion devices, which resulted in limited success. In this study, a dual-process W coating structure was developed on a graphite substrate to improve the thermal performance of the coating structure. The dual-process coating structure consisted of a thin (∼7 μm) multilayer W/Mo physical vapor deposition (PVD) coating layer deposited on top of the relatively thick (∼160 μm) plasma spray (PS) W coating on a graphite substrate panel. Then the coated sample was exposed to plasma heat flux of 1–3 MW/m{sup 2} for 300 s. With addition of a thin surface PVD coating layer, the microstructure change in underlying PS W coating was substantially reduced compared to the simple PS W coating structure. The thickness of overall coating structure was maintained for the dual-process PVD/PS coated samples after the thermal loading tests, while a significant reduction in thickness due to surface erosion was observed for the simple PS W coated samples. The improvement in surface erosion resistance in the dual-process coating structure was discussed in view of the characteristics of PVD and PS coating layers.

Current status of nanostructured tungsten-based materials development

International Nuclear Information System (INIS)

Kurishita, H; Matsuo, S; Arakawa, H; Hatakeyama, M; Shikama, T; Sakamoto, T; Kobayashi, S; Nakai, K; Okano, H; Watanabe, H; Yoshida, N; Torikai, Y; Hatano, Y; Takida, T; Kato, M; Ikegaya, A; Ueda, Y

2014-01-01

Nanostructured tungsten (W)-based materials offer many advantages for use as plasma facing materials and components exposed to heavy thermal loads combined with irradiation with high-energy neutron and low-energy ion. This paper first presents the recent progress in nanostructured toughened, fine grained, recrystallized W materials. Thermal desorption spectrometry apparatus equipped with an ion gun has been installed in the radiation controlled area in our Center at Tohoku University to systematically investigate the effects of displacement damage due to high-energy neutron irradiation on hydrogen isotope retention in connection with the nano- or micro-structures in W-based materials. In this paper, the effects of high-energy heavy ion irradiation on deuterium retention in W with different microstructures are described as a preliminary work with the prospective view of neutron irradiation effects. (paper)

Automatic kinetic Monte-Carlo modeling for impurity atom diffusion in grain boundary structure of tungsten material

Directory of Open Access Journals (Sweden)

Atsushi M. Ito

2017-08-01

Full Text Available The diffusion process of hydrogen and helium in plasma-facing material depends on the grain boundary structures. Whether a grain boundary accelerates or limits the diffusion speed of these impurity atoms is not well understood. In the present work, we proposed the automatic modeling of a kinetic Monte-Carlo (KMC simulation to treat an asymmetric grain boundary structure that corresponds to target samples used in fusion material experiments for retention and permeation. In this method, local minimum energy sites and migration paths for impurity atoms in the grain boundary structure are automatically found using localized molecular dynamics. The grain boundary structure was generated with the Voronoi diagram. Consequently, we demonstrate that the KMC simulation for the diffusion process of impurity atoms in the generated grain boundary structure of tungsten material can be performed.

Analytical method for thermal stress analysis of plasma facing materials

Science.gov (United States)

You, J. H.; Bolt, H.

2001-10-01

The thermo-mechanical response of plasma facing materials (PFMs) to heat loads from the fusion plasma is one of the crucial issues in fusion technology. In this work, a fully analytical description of the thermal stress distribution in armour tiles of plasma facing components is presented which is expected to occur under typical high heat flux (HHF) loads. The method of stress superposition is applied considering the temperature gradient and thermal expansion mismatch. Several combinations of PFMs and heat sink metals are analysed and compared. In the framework of the present theoretical model, plastic flow and the effect of residual stress can be quantitatively assessed. Possible failure features are discussed.

Analytical method for thermal stress analysis of plasma facing materials

International Nuclear Information System (INIS)

You, J.H.; Bolt, H.

2001-01-01

The thermo-mechanical response of plasma facing materials (PFMs) to heat loads from the fusion plasma is one of the crucial issues in fusion technology. In this work, a fully analytical description of the thermal stress distribution in armour tiles of plasma facing components is presented which is expected to occur under typical high heat flux (HHF) loads. The method of stress superposition is applied considering the temperature gradient and thermal expansion mismatch. Several combinations of PFMs and heat sink metals are analysed and compared. In the framework of the present theoretical model, plastic flow and the effect of residual stress can be quantitatively assessed. Possible failure features are discussed

Status of the ITER full-tungsten divertor shaping and heat load distribution analysis

International Nuclear Information System (INIS)

Carpentier-Chouchana, S; Hirai, T; Escourbiac, F; Durocher, A; Fedosov, A; Ferrand, L; Kocan, M; Kukushkin, A S; Jokinen, T; Komarov, V; Lehnen, M; Merola, M; Mitteau, R; Pitts, R A; Sugihara, M; Firdaouss, M; Stangeby, P C

2014-01-01

In September 2011, the ITER Organization (IO) proposed to begin operation with a full-tungsten (W) armoured divertor, with the objective of taking a decision on the final target material (carbon fibre composite or W) by the end of 2013. This period of 2 years would enable the development of a full-W divertor design compatible with nuclear operations, the investigation of further several physics R and D aspects associated with the use of W targets and the completion of technology qualification. Beginning with a brief overview of the reference heat load specifications which have been defined for the full-W engineering activity, this paper will report on the current status of the ITER divertor shaping and will summarize the results of related three-dimensional heat load distribution analysis performed as part of the design validation. (paper)

Evaluation of energy and particle impact on the plasma facing components in DEMO

International Nuclear Information System (INIS)

Igitkhanov, Yuri; Bazylev, Boris

2012-01-01

removal capability. From the plasma side it is particularly demanding to keep the bulk plasma contamination during the reactor long operational discharges below the fatal level. The possible damage of the FW materials due to the plasma sputtering erosion is estimated. The minimum thickness of the tungsten amour about 3 mm for W/EUROFER sandwich structure will keep the maximum EUROFER temperature below the critical limit for EUROFER steel under steady-state operation and ITER like cooling conditions.

Evaluation of energy and particle impact on the plasma facing components in DEMO

Energy Technology Data Exchange (ETDEWEB)

Igitkhanov, Yuri, E-mail: juri.gitkhanov@ihm.fzk.de [Karlsruhe Institute of Technology, IHM, Karlsruhe (Germany); Bazylev, Boris [Karlsruhe Institute of Technology, IHM, Karlsruhe (Germany)

2012-08-15

-state operation heat transfer into the coolant must remain below the critical heat flux (CHF) to avoid the possible severe degradation of the coolant heat removal capability. From the plasma side it is particularly demanding to keep the bulk plasma contamination during the reactor long operational discharges below the fatal level. The possible damage of the FW materials due to the plasma sputtering erosion is estimated. The minimum thickness of the tungsten amour about 3 mm for W/EUROFER sandwich structure will keep the maximum EUROFER temperature below the critical limit for EUROFER steel under steady-state operation and ITER like cooling conditions.

The Role of Spraying Parameters and Inert Gas Shrouding in Hybrid Water-Argon Plasma Spraying of Tungsten and Copper for Nuclear Fusion Applications

Czech Academy of Sciences Publication Activity Database

Matějíček, Jiří; Kavka, Tetyana; Bertolissi, Gabriele; Ctibor, Pavel; Vilémová, Monika; Mušálek, Radek; Nevrlá, Barbara

2013-01-01

Roč. 22, č. 5 (2013), s. 744-755 ISSN 1059-9630 R&D Projects: GA MPO FR-TI2/702; GA TA ČR TA01010300 Institutional support: RVO:61389021 Keywords : plasma spraying * tungsten * copper * inert gas shrouding * water-argon plasma torch * gas shroud * hybrid plasma torch * influence of spray parameters * nuclear fusion * oxidation Subject RIV: JG - Metallurgy Impact factor: 1.491, year: 2013 http://link.springer.com/content/pdf/10.1007%2Fs11666-013-9895-x.pdf

Material testing facilities and programs for plasma-facing component testing

Science.gov (United States)

Linsmeier, Ch.; Unterberg, B.; Coenen, J. W.; Doerner, R. P.; Greuner, H.; Kreter, A.; Linke, J.; Maier, H.

2017-09-01

Component development for operation in a large-scale fusion device requires thorough testing and qualification for the intended operational conditions. In particular environments are necessary which are comparable to the real operation conditions, allowing at the same time for in situ/in vacuo diagnostics and flexible operation, even beyond design limits during the testing. Various electron and neutral particle devices provide the capabilities for high heat load tests, suited for material samples and components from lab-scale dimensions up to full-size parts, containing toxic materials like beryllium, and being activated by neutron irradiation. To simulate the conditions specific to a fusion plasma both at the first wall and in the divertor of fusion devices, linear plasma devices allow for a test of erosion and hydrogen isotope recycling behavior under well-defined and controlled conditions. Finally, the complex conditions in a fusion device (including the effects caused by magnetic fields) are exploited for component and material tests by exposing test mock-ups or material samples to a fusion plasma by manipulator systems. They allow for easy exchange of test pieces in a tokamak or stellarator device, without opening the vessel. Such a chain of test devices and qualification procedures is required for the development of plasma-facing components which then can be successfully operated in future fusion power devices. The various available as well as newly planned devices and test stands, together with their specific capabilities, are presented in this manuscript. Results from experimental programs on test facilities illustrate their significance for the qualification of plasma-facing materials and components. An extended set of references provides access to the current status of material and component testing capabilities in the international fusion programs.

Vacuum hot-pressed beryllium and TiC dispersion strengthened tungsten alloy developments for ITER and future fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Liu, Xiang, E-mail: xliu@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan (China); Chen, Jiming; Lian, Youyun; Wu, Jihong; Xu, Zengyu; Zhang, Nianman; Wang, Quanming; Duan, Xuro [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan (China); Wang, Zhanhong; Zhong, Jinming [Northwest Rare Metal Material Research Institute, CNMC, Ningxia Orient Group Co. Ltd.,No.119 Yejin Road, Shizuishan City, Ningxia,753000 (China)

2013-11-15

Beryllium and tungsten have been selected as the plasma facing materials of the ITER first wall (FW) and divertor chamber, respectively. China, as a participant in ITER, will share the manufacturing tasks of ITER first-wall mockups with the European Union and Russia. Therefore ITER-grade beryllium has been developed in China and a kind of vacuum hot-pressed (VHP) beryllium, CN-G01, was characterized for both physical, and thermo-mechanical properties and high heat flux performance, which indicated an equivalent performance to U.S. grade S-65C beryllium, a reference grade beryllium of ITER. Consequently CN-G01 beryllium has been accepted as the armor material of ITER-FW blankets. In addition, a modification of tungsten by TiC dispersion strengthening was investigated and a W–TiC alloy with TiC content of 0.1 wt.% has been developed. Both surface hardness and recrystallization measurements indicate its re-crystallization temperature approximately at 1773 K. Deuterium retention and thermal desorption behaviors of pure tungsten and the TiC alloy were also measured by deuterium ion irradiation of 1.7 keV energy to the fluence of 0.5–5 × 10{sup 18} D/cm{sup 2}; a main desorption peak at around 573 K was found and no significant difference was observed between pure tungsten and the tungsten alloy. Further characterization of the tungsten alloy is in progress.

Soft X-ray radiation parameters of nested tungsten wire array

International Nuclear Information System (INIS)

Ning Jiamin; Jiang Shilun; Xu Rongkun; Xu Zeping; Li Zhenghong; Yang Jianlun

2011-01-01

Implosions with nested tungsten wire array were performed at the Angara-5-1 facility in Russian Research Centre. The experimental results of nested tungsten wire array are compared with those of single array. Radiation parameters of nested array are discussed based on four different dynamic models. When the implosions of outer and inner wire arrays are synchronized,the relatively uniform distribution of inner layer plasma will improve the uniformity of outer layer plasma. As compared with single array, nested array has an increase of 32% in X-ray radiation power. (authors)

A study of tungsten spectra using large helical device and compact electron beam ion trap in NIFS

Science.gov (United States)

Morita, S.; Dong, C. F.; Goto, M.; Kato, D.; Murakami, I.; Sakaue, H. A.; Hasuo, M.; Koike, F.; Nakamura, N.; Oishi, T.; Sasaki, A.; Wang, E. H.

2013-07-01

Tungsten spectra have been observed from Large Helical Device (LHD) and Compact electron Beam Ion Trap (CoBIT) in wavelength ranges of visible to EUV. The EUV spectra with unresolved transition array (UTA), e.g., 6g-4f, 5g-4f, 5f-4d and 5p-4d transitions for W+24-+33, measured from LHD plasmas are compared with those measured from CoBIT with monoenergetic electron beam (≤2keV). The tungsten spectra from LHD are well analyzed based on the knowledge from CoBIT tungsten spectra. The C-R model code has been developed to explain the UTA spectra in details. Radial profiles of EUV spectra from highly ionized tungsten ions have been measured and analyzed by impurity transport simulation code with ADPAK atomic database code to examine the ionization balance determined by ionization and recombination rate coefficients. As the first trial, analysis of the tungsten density in LHD plasmas is attempted from radial profile of Zn-like WXLV (W44+) 4p-4s transition at 60.9Å based on the emission rate coefficient calculated with HULLAC code. As a result, a total tungsten ion density of 3.5×1010cm-3 at the plasma center is reasonably obtained. In order to observe the spectra from tungsten ions in lower-ionized charge stages, which can give useful information on the tungsten influx in fusion plasmas, the ablation cloud of the impurity pellet is directly measured with visible spectroscopy. A lot of spectra from neutral and singly ionized tungsten are observed and some of them are identified. A magnetic forbidden line from highly ionized tungsten ions has been examined and Cd-like WXXVII (W26+) at 3893.7Å is identified as the ground-term fine-structure transition of 4f23H5-3H4. The possibility of α particle diagnostic in D-T burning plasmas using the magnetic forbidden line is discussed.